ML17216A690

Forwards Response to 860807 Request for Addl Info Re 860715 Application to Amend License NPF-16,revising pressure-temp Limits for RCS

ML17216A690
Person / Time
Site:	Saint Lucie
Issue date:	09/04/1986
From:	Woody C FLORIDA POWER & LIGHT CO.
To:	Tourigny E Office of Nuclear Reactor Regulation
References
L-86-354, NUDOCS 8609100021
Download: ML17216A690 (9)

Text

0 REGULATORY INFORNATION DISTRIBUTION SYSTEN (RIDS>

ACCESSION NBR: 8609100021 DOC. DATE: 86/09'/04 NOTARIZED: NO DOCKET ¹ FACIL: 50-389't. Lucie Plant> Unit 2> Florida Power 5 Light Co. 0 500038'P AUTH. NAME AUTHOR AFFILIATION WOODYz C. O. Florida Power 8r Light Co.

RECIP. NAME RECIPIENT AFFILIATION TOURIQNY> E. Q. PNR Prospect Directorate 8

SUBJECT:

Forwards response to 860807 request for addi info re 860715 application to amend License NPF-16'evising pressure-temp limits f.or RCS.

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FLORIDA POWER & LIGHT COMPANY

$g 4 1986 L-86-354 Office of Nuclear Reactor Regulation Attention: Mr. E. G. Tourigny, Project Manager PWR Project Directorate II8 Division of PWR Licensing - B U.S. Nuclear Regulatory Commission Washington, D.C. 20S55

Dear Mr. Tourigny:

Re: St. Lucie Unit 2 Docket No. 50-389 Pressure - Tem erature Limits By letter L-86-281, dated July IS, l 986, Florida Power 6 Light Company submitted proposed Technical Specifications to incorporate revised pressure-temperature limits for the reactor coolant system. On August 7, l986 you issued a request for additional information pursuant to NRC staff review of our proposal. The information you requested is attached.

Please contact us if you have any questions about this submittal.

Very truly yours, C. O. Woody Group Vic resident Nuclear nergy COW/MAS/cvb Attachment cc: Dr. J. Nelson Grace, USNRC, Region II Harold F. Reis, Esquire, Newman. 6 Holtzinger Sb0910002i ab0904 PDR ADQCK 05000389

PDR MAS2/035/ I PEOPLE... SERVING PEOPLE

ATTACHMENT Re: St. Lucie Unit 2 Docket No. 50-389 Pressure-Tem erature Limits Question I For an EOL fluence of either 3.64 x IO l9 n/cm (capsule report submitted November 5, l985, L-85-423) or 4e79 x IO l9 n/cm 2 (PTS submittal dated January 23, l986, L-86-25) the transition temperature shift is greater than 200 F (Figure B3/4.4-l, page B3/4 4-IO). You committed to meet Appendix H of IO CFR 50 Part l00, which references ASTM EI85. Table I of ASTM EI85-82 calls for 5 surveillance capsules with the following withdrawal sequence.

~Ca sule Schedule in FPY I st I.5 2nd 3 3rd 6 4th I5 5th EOL Table 4 4-5 of the Technical Specifications does not reflect this withdrawal schedule (page 3/4 4-34). Please provide a justification for your withdrawal schedule or commit to the above schedule.

~Res onse I Based on the response to Question 2 below, either the current Figure B3/4.4-I or the revised Figure B3/4.4-I (when it becomes available) will be compared with the appropriate version of ASTM EI85 to derive a surveillance capsule withdrawal sequence. If the current sequence needs to be revised, a proposed Technical Specification Change will be submitted by December I, I 986.

MAS3/039/ I

Question 2 Based on the information obtained from the first capsule withdrawn from the reactor pressure vessel, is the curve in Figure B3/4.4-I still applicable?

~Res onse 2 Based on information from the first surveillance capsule removed from the St. Lucie Unit 2 reactor pressure vessel, we have determined that the current Figure 83/4.4-I is conservative. However, we also intend to use the capsule data to revise the Figure. If sufficient benefit can be derived from using the revised Figure 83/4.4-I, a proposed Technical Specification change will be submitted by December I, l 986.

MAS3/039/2

Question 3 Please provide the data used to calculate the P-T limits.

What is (are):

a. Limiting (controlling) material and its Cu and Ni content?

Res onse 3a The limiting material in the reactor vessel through End of Life (EOL) is the intermediate shell plate M605-2. Its copper and nickle content is O. I3 and 0.62 w/o, respectively.

b. Fluence used at 5 EFPY and EOL?

Res onse 3b EFPY ID I/4 t 3/4 t 5 0.9I x IO n/cm 0.49 x IO n/cm O. I I x IO n/cm 32 5.8 x IO n/cm2 3. I x IO n/cm2 0.70 x IO n/cm This data was calculated based on a detailed octant symmetric DOT model of the core-to-vessel configuration. The core power distribution was modeled in a detailed manner using pinwise power distribution data. The core power distribution was provided by Florida Power 6 Light Company and was chosen to bound future fuel management strategies with respect to the fluence accumulated in the reactor vessel. Ordinary out-in and (low leakage) in-out fuel management strategies will be bounded.

Both the 5 EFPY and 32 EFPY fluence values were based on the same bounding fuel management strategy. The fluence values were quoted for the peak fluences at a given vessel radius.

The DOT model did not include a representation of the concrete cavity wall but the 3/4T fluence values wer'e adjusted to account for the backscatter due to the wall and cavity structures. The values are based upon 2700 Mwt power operation and on 8.625" vessel thickness.

MAS3/039/ I

c. ID, OD, and thickness of reactor pressure vessel?

~R* 3*

Inside Diameter (ID) I 72.44" Outside Diameter (OD) I 89.69" Wall Thickness (t) 8.625"

d. Initial RTNDT for flange, girth weld, longitudinal weld, and base metal.

Res onse 3d Initial RT~DT For: lni t i al RTNDT (oF)

Flange Girth Weld -IO'7O'O'.

Intermediate Shell Plate M605-I Intermediate Shell P late M605-2 Intermediate Shell Plate M605-3 Lower Shell Plate M4 I I 6- I Lower Shell Plate M4 I I 6-2 IO'IO'O'O'O'8O'oo Lower Shell Plate M4 I I 6-3 Intermediate Shell Long. Weld I 0 I- I 24A Intermediate Shell Long. Weld IOI-I24B Intermediate Shell Long. Weld I 0 I- I 24C Lower Shell Long. Weld IOI- I42A Lower Shell Long. Weld IOI-I42B

-so'so'so'so'.

Lower Shell Long. Weld I 0 I -I 42C Standard deviation of initial RTNDT (6;) for girth weld, longitudinal weld, and base metal?

Res onse 3e All of the initial RTNDT values from above are single reported values. As such, the standard deviations are equal to zero.

MA53/039/2

f. Standard deviation of Q RTNDT 0+ for girth weld, longitudinal weld, and base metal?

Res onse 3f The shift in RTNDT was calculated using the proposed Revision 02 to Regulatory Guide l.99. Accordingly, the standard deviation for the welds are 28 F, and for the base metal is l7 F.

g. Temperature measurement error?

Res onse3 The temperature measurement error was calculated to be 8.0 F.

h. Design pressure and operating pressure?

Res onse 3h The design pressure of the reactor coolant system is 2500 psia.

The operating pressure of the reactor coolant system is 2250 psia.

i. Hydrostatic head?

Res onse 3i The hydrostatic head is calculated to be l9.I psia from the pressurizer to RV bel t line.'

A53/039/3

j. Pressure measurement error?

Res onse 3'he pressure measurement error is calculated to be as follows:

23.9 psia for digital display P-I I03, P-I l05 for pressures less than or equal to 750 psia.

64.5 psia for displays P-I I07, P-I I08 for pressures greater than 750 psia.

k. What are the calculated valves of the adjusted reference temperature for the limiting material?

Res onse 3k The predicted adjusted RTNDT for plate M605-2 is I I7 F for the I/4t location and 84 F for the 3/4t location at 5 EFPY. The predicted adjusted RTNDT for plate M605-2 is l63 F for the I/4t location and l26 F for the 3/4t location at 32 EFPY.

MA53/039/4