ML18227D307
| ML18227D307 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 11/30/1976 |
| From: | Robert E. Uhrig Florida Power & Light Co |
| To: | Lear G Office of Nuclear Reactor Regulation |
| References | |
| Download: ML18227D307 (7) | |
Text
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F,ILE NUMOEA DATE'OF DOCUMENT 11-30-76
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furnishing supplemental info re'steam gen-erator inspection results.......
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FL'ORIDA POWER & LIGHTCOMPANY
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'V Office of Nuclear Reac Attn:
George Lear, Chief Operating Reactors B
ch 53 Division of Operating Reactors U. S. Nuclear Regulatory Commission Washington, D. C.
20555 November 30, 1976 L-76-409
Dear Mr. Lear:
Re:
Turkey Point Unit 4 Docket, No. 50-251 Steam Generator Inspection Results Su lemental Information DES Nts~
QAC9KQ This letter supplements the information forwarded to you on November 24, 1976 (L-76-403) concerning the subject, inspec-tions.
In order to gain some further measure of assurance that row 2 tubes are not susceptible to the cracking phenomenon, cracking calculations have been done to compare the equiva-lent strain for a row 1 tube which has not experienced cracking in service to the ectuivalent strain for the most potentially affected row 2 tube at the location coinciding with the center of the flow slot.
These calculations are explained and summarized in the response to question 5 on pages 14-20 in the enclosure to VEPCO/NRC letter Serial No.
260B/092276, dated November 15, 1976 (Docket No. 50-281).
Additional calculations have since been completed for a Series 44 steam generator row 2 tube.
These additional calculations assume complete closure of tne adjacent flow slot resulting in a 2.75 inch displacement of the legs of the U-bend.
The equivalent, strain calculated for this case was less than the equivalent strain of 0.135 in/in. for the uncracked row 1 tube used for comparison in the VEPCO sub-mittal.
Since the tubing for the Surry and Turkey Point steam generators is identical, this is a valid comparison.
.Table 1 provides flow slot closures for tube support plate 1
"(bottom) and 6 (top),
and calculated closure rate for steam
-" generator 4B.
Examination of steam generator inspection data to date has shown that hour glassing is more advanced in steam generator 4B, as compared to'team generators 4A'and 4C.
Therefore, the calculated closure rate for'team 'generator'B should exceed the.closure rates for steam generators 4A-and 4C.
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TABLE NO.
1 STEAM GENERATOR 4B FLOW SLOT NUMBER
~GAP (inches)~~~h-h Pl Tube Support Plate a)
April-June 1976 b)
November 1976 2-3/8 2-5/16 2
2-1/4 1-7/8 1
1-1/4 2-1/16 1-15/16 1-15/16 0'6 Tube Support Plate November 1976 1-7/8*
2 2-1/4 (lhnimum Hourglassing)
R1 = Rate of Closure for i'21 TSP Growth of Worst Slot Considering Worst Case (0'4 Slot)
Hours S/G Operated above 350 F
.938 3.3 x 10 'n 1
'~2828.2 2828.2 oper hr
.238 in.
oper month R6 = Rate of Closure for 86 TSP
= thClosure
.'6 TSP
~G.=I TS 1
(2.3/4 -
1 7/8)
(.239 in
)
.110 in 2 3/
(
~on~h) oper month
- Row 1, Row 2, and Row 3 Tubes, adjacent to this flow slot have been plugged as a
result of tube sample removal.
1")
To:
George Lear, NRC Re:
Turkey Point Unit 4
Steam Generator Ins ection Results November 30, 1976 Page Using a closure rate of.11 inches per operating month above 350'F (Table 1),
542 days of operation above 350'F would be required to reach a gap width of 0.0".
Using the same closure
- rate, 406 days of operation would be required to reach the 0.5 inch gap used as acceptance criteria on page 9 of VEPCO letter 260B/092276, dated November 15, 1976.
VEPCO's use of a 0.5 inch gap as acceptance criteria was based on the absence of cracking in row 2 tube samples removed from lo-'ations adjacent to the flow slot with a measured gap. of 0.5 inches.
These calculations use flow slot No.
2 (46 TSP) as the datum point, since this flow slot is the flow slot with the minimum gap whose adjacent row 2 tubes are not plugged'Rows 1, 2, 3 tubes adjacent to No.
1 flow slot have been plugged as the result of tube sample removal).
Turkey Point Unit.
4 steam generators will be reinspected during its next scheduled refueling outage, or within 140 equivalent operating days with reactor coolant system temperature greater than 350'F, whichever occurs sooner.
Unit 4 is now scheduled to commence its next refueling during April 1977.
The basis for this reinspection schedule is provided in the preceding paragraphs.
Very truly yours, Robert. E. Uhrig Vice President.
REU/GDN/hlc Attachment cc:
Norman C.'oseley, Region II Robert Lowenstein, Esq.