ML17346A887

Forwards Response to NRC 850226 Request for Addl Info Re Reactor Plant Surveillance Matl Program

ML17346A887
Person / Time
Site:	Turkey Point
Issue date:	03/06/1985
From:	Williams J FLORIDA POWER & LIGHT CO.
To:	Varga S Office of Nuclear Reactor Regulation
References
L-85-97, NUDOCS 8503120503
Download: ML17346A887 (12)

Text

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'Operat)ng, Reactors'r'anchi 1>

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+ Llg FLORIDAPOWER & LIGHTCOMPANY March 6, 1985 L-85-97.

Office of Nuclear Reactor Regulation Attention:

Mr. S.

A. Varga, Chief Operating Reactors Branch III1 Division of Operating Reactors U.

S. Nuclear Regulatory Commission Washington, D.

C.

20555

Dear Mr. Varga:

Re: Turkey Point Units 3 and 4

Docket Nos. 50-250 and 50-251

Response

to Request for Additional Information on the Reactor Plant Surveillance Material Pro ram By letter dated February 26, 1985, you requested additional information concerning the above-referenced subject.'he attachment to this letter contains that additional information.

We appreciate your prompt review of this subject and are available should you have any further questions.

Very truly yours, J.

W. Williams, Jr.

Group Vice President Nuclear Energy JWW/SAV/js Attachment cc: Harold F. Reis, Esquire File 933.1 PNS>>LI-85-092-1 BS031QOS03 8S03pg tT DOCK QS000260 POR PEOPLE... SERVING PEOPLE

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ATTACHMENT Re: Turkey Point Units 3 and 4

Docket Nos.

50-250 and 50-251

Response

to Request for Additional Information on the Reactor Plant Surve'illanc'e Mateirial Program guestion 1

Are the materials and designs for the core, thermal shield, core barrel and vessel the same for both units (PTP 3 and 4)?

Response

Yes.

The materials and designs are the same.

The references are:

The Updated Final Safety Analysis Report Volume 2.

(Common for both units)

Materials reference is Table 4.2-1.

Design reference is contained in Sections 3 and 4.

guestion 2

Is the figur e in Attachment 3 (of the proposed Tech Spec change) representative of both units?

Response

Yes.

It is representative of both units.

Reference:

SwRI report of project numbers 02-5131 and 02-5380.

This document is the capsule "S" 'surveillance report, May 1979 for both units 3 and 4.

guestion 3

What are the materials (base and weld) in the beltline of each vessel?

Indicate the material specification, heat numbers (flux and weld wire), amount of Cu, Ni, P, Initial RTNDT for each material.

How were initial RTNDT, Cu, Ni and P determined.

Response

Table 1 shows the values.

Cu, Ni analyses are based on the statistcal mean of 51 and 41 individual analyses as reported to the NRC in FPL letter L-84-31, dated February 10, 1984.

P concentrations are based on WCAPs 7656 and 7660,

,(Table Al), the surveillance programs for both units.

Initial RTNDT f<< the weld material SA1101 is based on dropweight and charpy data as reported in FPL letter L-84-31.

RT<D> (initial) values were never generated for the forging materials using botli dropweights and charpy data.

Forging charpy curves and data are shown in WCAPs 7656 and 7660.

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guestion 4

How was it determined that SA1101 was the limiting material.

What is the projected EOL RT

>~ values from Table 1. (This is very conservative for forgings since 50 Ft. lb. initial RTNDT was used with the 2 sigma-term. guestion 5 Describe the excore dosimetry program and how it will be used to determine the peak neutron fluence for each vessel. Has benchmarking been completed.

Response

Dosimetry (activation threshold foils, fission foils and HEDL SSTRs) have been installed in the reactor cavity area of both units since the last refueling outages.

The dosimentry will verify analytical transport models.

We plan to continue to monitor Unit 3.

Benchmarking will be completed after Unit 3 cycle 9 and Unit 4 cycle 10 dosimetry has been evaluated.

guestion 6

Are the neutron spectrum and neutron flux the same for each vessel at the peak fluence locations?

Are the spectrum and flux the same for each capsule T, V, and X in each vessel?

How will the dosimetry from one vessel be used to estimate the neutron fluence and spectrum for each vessel?

Response

The variations in neutron spectra

(> 1.0026 Mev) profiles are negligible for both units at the peak fuence locations.

This is based on FPL DOT 4.3 transport code predictions.

Fuel management and cycle lenths are very similar.

Neutron spectrum at capsule T (270') is slightly softer than spectrum at capsule X (50').

The higher flux is received at capsule T.

Vessel fluence will be estimated by using measured surveillance capsule data and predicted lead factors from each vessel.

Unfolded measured spectra from in core and excore dosimetry will be used to refine predicted vessel neutron spectra.

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guestion 7

What is the end of life peak neutron fluence at the 1/4T location?

How was this value and the EOL fluence at the surface determined.

Response

The predicted EOL 1/4T fluence is 6.8xl0 n/cm2.

The bases, for these predictions are reactor vessel fluence calculations using FPL PDg7 code for cycle 1 through 9.

The PD(7 power distribution data was transformed'nto neutron source data by SORREL for input to DOT 4.3.

As built dimensions and materials properties were incorporated into the DOT 4.3 calculational models.

FPL will document calculational

methods, computer codes, benchmarks, and vessel neutron fluence results.

guestion 8

!<hat is the projected neutron fluence to be received by each capsule at the time of its withdrawal7 Does the withdrawal schedule comply with ASTN E185-82.

If it does not, state why the requested schedule is more reasonable.

Response

The projected fluences are as follows:

Capsule g Unit Removal Date Capsule Fluence E185-82 requirement 3V 4V 3X 12 years 24 years 33 years 7.9xl018n/cm2

1. 6xl019n/cm2 1.1xl019n/cm2 6 8x 1018 n/cm2 1.25x10 n/cm

) 1.25x1019 n/cm2

< 2.5 x 19 n/cm The 3V and 4V capsules meet the r equi rements of the specification.

The 3X capsule fluence is lower than the specification requires but the specification allows this value to be modified based on previous testi ng or it may never be tested since this is the EOL capsule.

This schedule is more appropriate because it contains all critical material and will supply much more meaningful results than the existing schedule.

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TABLE 1 Unit Component Heat Number Cu Initial EOL RTNPT RTNPT intermediate SA 508 CL2 shell 123P461VA-1 3

lower shell SA 508CL2 123S266VA-1 0.058 0.70 0.01

-25'F*

73'F

0. 079

0. 68

0. 01

-41'F*

72'F weld 'wire weld flux PAGE 71249 Linde 80 8445 0.26 0.60 0.011 10'F 257'F intermediate SA 508 Cl2 shell 123P481VA-1 lower shell 122S180VA-1 4

weld wire PAGE 71249

0. 054

0. 71

0. 01 25'F*

123'F 0.056 0.70 0.01 O'F*

98'F weld flux

• 50 ft.lb. Temp Linde 80 8445

0. 26

0. 60

0. 014 10'F 257'F

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