ML20216J604

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Evaluation of New Surveillance Data on Chemistry Factor for Weld Wire Heat 61782,Point Beach Unit 1
ML20216J604
Person / Time
Site: Point Beach NextEra Energy icon.png
Issue date: 09/23/1999
From: Hanneman R, Pfefferle J, Spry T
WISCONSIN ELECTRIC POWER CO.
To:
Shared Package
ML20216J602 List:
References
98-0156-00-A, NUDOCS 9910050300
Download: ML20216J604 (10)
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Text

' NPL 99-0555 Nuclear Power Business Unit Page1of to Attachment 2 CALCULATION DOCUMENT FORM Calculation / Addendum Number: Title of Calculation /Addeng:

98-0156-00-A Evaluation of New Surveillance Data on Cnemistry Factor for Weld Wire Heat 61782, Point Beach Unit 1

@ OriginalCalculation/ Addendum @ Supersedes Calculation / Addendum O Revised Calculation / Addendum Revision # Framatome Technologies Calculation No. 32-5002184-00 en A=sociated Documents:

@ QA Scope ,

O Non-QA Scope Superse M By Calculation / Addendum #

This Calculation has been reviewed in accordance with NP 7.2.47 The review was accomplished by one or a combination of the following (as checked):

A review of a representative sample of repetitive A detailed review of the original calculations. calculation.

A review of the calculation against a similar A review by an alternate, simplified, or

/ calculation previously performed. approximate method of calculation.

Page Inventory:

Page 1 - 4 Form PBF-1608 Attachments: None Page 5 - 10  : Pages through Prepared By: Date:

James Pfefferle . 9/2t/sq99 Print Name /

[ S/gliature Reviewed dy: Date:

1/22/99 Thomas Spry 4 '

Print Name Signature -

I /

Approved By: f Date:

R. K. Hanneman e h A i7 Print Name Signature ' '

rt10050300 N i.p A . ADOCK O PDR N

Reference:

NP ?.2 4

r tWL @9-0555 a

Nuclear Power Business Unit Attachment 2 Calculation / Addendum: 98-0156-00-A CALCULATION DOCUMENT FORM Page 2 0f 10 Preparer: J.R. Pfefferle j' Date:

9/22/1999 ( _

l . . "

Cciculation Checklist (Optional for Non-QA Scope)

I' Attribute Description N/A Author Reviewer N

l 1. Purpose

a. Is the purpose clearly stated indicating issue to Myes no be resolved or information to be determined?
2. Methodology and Acceptance Criteria
a. Has the method / approach been described? T
b. Have appropriate acceptance criteria and their sources been identified?

l 3. Assumptions I ,

a. Are the assumptions provided with sufficient [ dyes no rationale to permit verification?
b. Have assumptions associated with pending N/A U Uyes Uno l

plant or procedure changes tnat require verification been identified?

g

c. Have the requirements to revise governing N/A yes no calculations or verify pending assumptions been documented in r, modification or an EWR7 A
5. References
a. Have all the appropriate references, including revisions and/or dates, been identified?

T Oyes Ono

b. Are all references readily available in the PBNP Z Byes Ono Records System, as public documents, or attached?
4. Inputs
a. Have the applicable inputs and sources been T identified?
b. Is the source for each input identified and listed in the References and/or Assumptions?

T Syes Ono

6. Calculation
a. Have formulae and inputs been provided consistent with the source document, including 7 Myes no engineering units?
7. Computer-Aided Design Calcalations (NP 7.2.4 l Attachment A)
a. Has the computer program been validated per N/A yes no

, the requirments of Attachment A? N//l

h. HSve the program version and revision been N/A yes no identified on the computer run and in the calculation?

' PDF-1608 Revision 4 06/30/99

Reference:

NP 7.2.4

m WPL @9-0555

~

~

Nuclear Pow:r Business Unit Cdculation/ Addendum: 98- 1 $

CALCULATION DOCUMENT FORM Page3of to i Preparer: J.R. PfefferleM Date:

9/22/1999 (

, . w Attribute Description N/A Author Reviewer

c. Is the input to the computer program adequately N/A O Oyes Ono documented? N/A  ;
d. If spreadsheet or other simple computer aided N/A Oyes Ono l

~' ' '

tools are used in the calculation, have the formulae been documented in the calculation? MM

e. Have the attributes been documented in the N/A _ O Oyes Ono calculation for any input or output data files supporting the calculation, including file name, N//l date stamp, time stamp (hour and minute only),

and file size?

8. Summary of Results and Conclusions
a. Do the summary of results and conclusians T clearly state the calculation results and respond to the purpose?
b. Do the conclusions address the acceptability / T unacceptability of the results?
c. Has a CR been initiated to identify any N/A Uyes Uno ,

unsatisfactory conditions? N/A

d. Have all engineering judgments been provided [ylyes Ono j with sufficient rationale?
9. Administrative
a. Have calculation format and content as noted in T NP 7.2.4 been followed? ,

i

b. Have all required attachments been included in @ Myes Ono the document and numbered appropriately? ,
c. Has the calculation been prepared neatly and Syes Ono legibly with sufficient contrast to allow satisfactory record copies to be produced?
d. Are the calculation number, preparer's initials, T hyes Ono preparation date, and page number provided on each page?
e. Have revisions been clearly identified by N/A U Uyes [no revision bars or other appropriate means (for revised calculations only)?. M
f. If the calculation is a revision, has a listing of N/A yes Ono successor documents, including impact and any required actions initiated, been incorporated as an attachment?

PDF-1608

- Revision 4 06/30/99~

Reference:

NP 7.2.4

g, NPL 99-0555 Attachment 2

~

Nuclear Power Business Unit _. C lculation/ Addendum: 98-0156-00-A

~

.~~

CALCULATION DOCUMENT FORM Page 4 of 10 Preparer:

JA PfefferleM Date: 9/22/1999

," Attribute Description N/A Author Reviewer

g. If the calculation supersedes a previous [ dyes no calculation, is this noted on the cover sheet?
h. Has the calculatica been appropriately @ gyes Ono identiSed as QA or Non-QA scope?
i. Has the review method been clearly identified on the cover page? -
j. Is all information required by PBF-1620 entered D on the form?
k. If calculation creates a potential DBD open item, has a form PBF-1611 been prepared and Mo Mb g in accordance with NP 7.7.37 OI cJ COMMENTS AND RESOLUTION "/"'

Reviewer Comments: Resolution:

l l

PBF-1608 Revision 4 06/30/99

Reference:

NP 7.2.4 9

I M Attachment 2 PATE

[ 5 CF_R l ,

j CALCULATICN SHEET l, CALCULATIUN/ ADDENDUM NO. 98-0156-00-A TITLE Evaluation of New Surveillance Data on Chemistry MADE BY J.R. Pfefferler DATE 9/22/99 Factor for Weld Wire Heat 61782, Point Beach Unit 1 REV'D. BY T.D. Sorvl l DATE 9/22/99

\ f Introduction &

Purpose:

Based on guidance from the NRC Staff (Ref.1) and on the Pressurized Thermal Shock j (PTS) Rule,10 CFR 50.61 (Ref. 2), reactor vessel surveillance data must be considered in evaluating reactor vessel integrity. The best-estimate copper and nickel chemical compositions are used in the evaluation of the surveillance data. The process of evaluating surveillance data includes a credibility assessment and the calculation of the chemistry factor based on the surveillance data. The NRC Staff provided guidance on performing evaluations of surveillance data in a public meeting between the Staff, Nuclear Energy Institute (NEI), and industry representatives on November 12,1997. A summary of this meeting is documented in a meeting summary dated November 19,1997 (Ref.1).

This calculation prov: des she evaluation of weld wire heat 61782 surveillance data for assessing the integrity of the Point Beach Unit 1 reactor vessel. The NRC Staff's guidelines are used to account for differences in material best estimate chemistry and reactor operating temperature in computing the chemistry factor. The impact of this chemistry factor on the Point Beach Unit 1 limiting beltline material will be determined.

Table of Contents:

Introduction & Purpose 5 References 5 Methods & Acceptance Criteria 6 Assumptions 6 inputs 6 Calculations 7 Results and Conclusions 10

References:

1. Mernorandum from Keith R. Wichman to Edmund J. Sullivan, " Meeting Summary for November 12,1997, Meeting with Owners Group Representatives and NEl Regarding Review of Respases to Generic Letter 92-01, Revision 1, Supplement 1 Responses,"

dated November 19,1997.

2. 10 CFR 50.61, " Fracture toughness requirements for protection against pressurized thermal shock events."
3. NRC Regulatory Guide 1.9i.:, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials," May 1988.
4. BAW-2325, Revision 1," Response to Request for Additional Information (RAl)

Regarding Reactor Pressure Vessel Integrity," January 1999.

5. WCAP-14684, "R. E. Ginna Heatup and Cooldown Limit Curves for Normal Operation,"

June,1996.

6. WEPCO Calculation No. 98-0156, Revision 0, " Evaluation of P-T Limit and LTOP l Applicability Date and Pressurized Thermal Shock," 12/7/98. ]
7. WEPCO letter VPNPD-93-186, NRC-93-115, November 1,1993.

I 1

l l

p NPL 99-0555 l

Attachment 2 PA7E 6 CF R

l. CALCULATISN SHEET

}

p, CALCULATI5N/ ADDENDUM N3. 98-0156-00-A TITLE Evaluation of New Surveillance Data on Che.T. Ju v MADE BY 3.R. Pfefferi DATE 9/22/99 Factor for Weld Wire Heat 61782, Point Beach Unit 1 REV'D. BY T.D. Sorv / \ DATE 9/22/99 V

Methods & Acceptance Criteria:

The methodology and acceptance criteria of this calculation are consistent with Reg ilatory Guide 1.99, Rev. 2 (Ref. 3) and the NRC's guidance from the November 12,1997 meeting.

It follows the steps listed below:

1. Credibility of the available surveillance data is evaluated by determining the magnitude of difference between the measured ARTuor and the predicted ARTuor for each surveillance capsule. Acceptance esiteria: if the difference between the measured and predicted ARTuor is less than or equal to 28'F, the surveillance data is deemed credible.
2. The chemistry factor for the surveillance data relat!ve to the Point Beach Unit 1 reactor vessel is then determined as the slope of a best-fit line through the surveillance data, adjusted to account for differences in chemical composition (i.e., copper and nickel contents) and irradiation environment (i.e., irradiation temperature) between the i capsules ard the vessel being assessed (i.e., R.E. Ginna surveillance capsules versus Point Beach Unit 1 weld metal).

Assumptions:

1. It is assumed that multiple units of the same nuclear steam supply system (NSSS) design (e.g., Westinghouse) are closer in terms of irradiation environment than units of different NSSS design (e.g., Westinghouse versus Babcock & Wilcox). This is consister:t with the NRC Staff guidance of November 12,1997,
2. Based on Table 5 of Reference 7, 538 F is a conservative cold leg temperature to use for Point Beach Unit 1.

1 Inputs:

The lower shell longitudinal weld in the Point Beach Unit 1 reactor vessel was fabricated using the weld wire heat 61782 (Ref. 6). The best estimate copper and nickel chemical I compositions for this wire heat are as follows (Ref. 4): l Cu = 0.23 wt% Ni = 0.52 wt%

Weld wire heat 61782 surveillance data are not available from the Point Beach Unit 1 plant- l j specific surveillance program, but are available from other sources. The available weld wire l' heat 61782 surveillance data are presented in Table 1.

L I

4

NPL 99-0555 Attachment 2 PA2E 7 CF_1(L

. CALCULATICN SHEET

  • l

. CALCtlLATION/ ADDENDUM NO. 98-0156-00-A TITLE Evaluation of New Surveillance Data on Chemistry MADE BY J.R. Pfefferi DATE 9/22/99 Fa'2or for Weld Wire Heat 61782, Point Beach Unit 1,,, *EV'D. BY T.D. Sorv!

DATE _9122/99 Table 1. Available Surveillance Data for Wald Wire Heat 61782 (Ref. 4 & 5)

Irradiation Flue e Measured Capsule ID Cu Ni Temperature (x10 ARTun, l'F)

(including source) wt% wt%

( F) n/cm') (TANH)

B&WOG: Capsule DB1-LG1 SA-1135: ONS 2 Nozzle Belt 0.27 0.59 556- 1.03 141 Dropout Material i Ginna: Capsule V, SA-1036:

0.24 0.!;2 545 0.5028 146 Plant Specific RVSP Material Ginna: Capsule R, SA-1036:

0.24 0.52- 545 1.105 167 Plant Specific RVSP Material Ginna: Capsule T, SA-1036:

Plant Specific RVSP Material 0.24 0.52 545 1.864 169 Ginna: Capsule S, SA-1036:

0.24 0.52 545 3.746 223 Plant Specific RVSP Material The B&WOG Capsule DB1-LG1 was irradiated in the Davis-Besse reactor vessel which is of

' Babcock & Wilcox NSSS design. The weld metal surveillance data for R.E. Ginna was  !

irradiated as part of its plant-specific surveillance program.

Calculations:

Evaluation and Use of Wald Wire Heat 61782 Surveillance Data Examination of the available surveillance data for weld wire 61782 reveals that the magnitude of the temperature adjustment with respect to Point Beach Unit 1 (cold leg l temperature (CL) = 538'F, based on Ref. 7) is lower for the R.E. Ginna (CL = 545'F) data than it would be for Davis-Besse (CL = 556 F). In addition, Point Beach Unit 1 and R.E.

Ginna are both Westinghouse 2-loop NSSS designs. Therefore, based on the guidance l provided by the NRC Staff on November 12,1997, only the R.E. Ginna weld metal 61782 I surveillance data is used in assessing the integrity of the Point Beach Unit 1 vessel.

Credibility Assessment Because the R.E. Ginna weld metal 61782 surveillance data is from one (1) source, a best-fit line is first determined using a least squares fit function based on Regulatory Guide 1.99 Rev. 2 to relate the measured (unadjusted) ARTuor to the fluence factor (determined from the capsule fluence). Table 2 presents the determination of the slope uf the best-fit line for the R.E. Ginna weld metal 61782 surveillance data.

r: WM~W2Whis i Attachment 2 PATE 8 C F_1Q_

. CALCULATION SHEET l I' .

CALCULATIEN/ ADDENDUM N3. 98-0156-00-A TITLE Evaluation of New Surveillance Data on Chemistry MADE BY J.R. Pfefferi DATE 9/22/99 Factor for Weld Wire Heat 61782. Point Beach Unit 1 REV'D. BY DATE 9/22/99 T.D. Sorv( ~ l

! Table'2. Slope of Best-Fit Line for R.E. Ginna Weld Metal 61782 Surveillance Data I Fluer e * **" d Capsule ID Fluence a tor x (Fluence (including source) A" j Factor (A) Measured Factor) 2, TA h, I ARTuor Ginna: Capsule V, SA-1036:

Plant Specific RVSP Material 0.5028 0.8081 146 118.0 0.6531 Ginna: Capsule R, SA-1036:

Plant Specific RVSP Material

. 05 1.028 167 171.7 1.057 j

Ginna: Capsule T, SA-1036:

1.864 1.171 169 197.8 1.370 Plant Specific RVSP Material Ginna: Capsule S, SA-1036:

Plant Specific RVSP Material -

3.746 1.342 223- 299.2 1.801

)

i SUM: 786.7 4.880 l l

Slope of Best-Fit Line = 786.7 + 4.880 = 161.2 Note (Af: The dimensionless Fluence Factor (FF) calculated in accordance with Regulatory l Guide 1.99 Rev. 2, Equation 2: FF = f m.2a-cao ma n , where f is the fluence in units l of E19 n/cm .2 For example, for Capsule V above, FF = 0.5028 m.2e-cao wo o.so2:3 j 0.8081.

Slope of Best-Fit Line = 161.2*F Table 3 presents the credibility assessment for weld wire heat 61782, comparing measured to predicted shifts using the best-fit line, using only the R.E. Ginna plant-specific weld ,

metal surveillance data.

Table 3. Credibility Assessment of R.E. Ginna Weld Metal 61782 Surveillance Data Meas. Predicted Measured -

' Pe est. A uor, AR "

Capsule Designation Fluence Factor p. p Predicted

, Best (TANH) Line ('F) (A) ARTuor M Ginna: Capsule V Plant i Specific RVSP Material -

Ginna: Capsule R Flant 1.028 161.2 167 165.7 1.3 Specific RVSP Meterial Ginna: Capsule T Plant -19.7 1.171 161.2 169 188.7 Specific RVSP Material Ginna: Capsule S Plant 216.3 6.7 1.342 161.2 223 Specific RVSP Material Note (A): Predicted ARTwor from best fit line = (Slopeu ,,m)

  • Fluence Factor

F.

NFL 99-0555 Attachment 2 PAGE 9 CF_1[L l . CALCULATISN SHEET

' }

CALCULATION / ADDENDUM ND. 98-0156-00-A

} TITLE Evaluation of New Surveillance Data on Chemistry MADE BY J.R. Pfeffer DATE 9/22/99 ]

Factor for Weld Wire Heat 61782, Point Beach Unit 1 REV'D. BY T.D. Sors DATE 9/22/99 1 V

These data are credible, since the scatter is less than t28'F for all surveillance capsule data points.

Determination of Chemistry Factor The surveillance data chemistry factor is determined from a best-fit line through the surveillance data adjusted to account for differences in chemical composition (i.e., copper and nickel contents) and irradiation environment (i.e., irradiation temperature) between the capsules and the vessel being assessed (i.e., R.E. Ginna versus Point Beach Unit 1).

The operating cold leg temperature (Tn nt) for the Point Beach Unit 1 reactor vessel is 538'F, and the R.E. Ginna surveillance capsules have an irradiation temperature (Tc.,,oi,) of 545'F, which by inspection is greater than Tn ,,. Therefore, based on the guidance provided by the NRC Staff on November 12,1997, for the capsules with Tc.,,o greater than 538'F (i.e., Tn nt), the ARTuor, .. o,,o must be adjusted by increasing the measured ARTuor by 1.0*F for each degree difference in irradiation temperature to yield the temperature adjusted ARTuor (i.e., ARTuor,1,mp.4ajo,,,o).

To account for the difference in chemical compositions between the surveillance data and the weld wire heat best estimate, the surveillance data are normalized to the best estimate of the vessel being assessed. To obtain the " temperature and ratio" adjusted ARTuor, the surveillance data are adjusted as follows:

Temperature-Ratio Adjusted ARTuor = [CFr,u,,v,. ics, / CFr i,,,, son,,cn,m]

  • ARTuor,1,m,.4 3o.,,o The assessment of the surveillance data for weld wire heat 61782 with respect to Point Beach Unit 1 using the R.E. Ginna plant-specific weld metal surveillance data is presented in Table 4.

Table 4. Surveillance Data Ass 6ssment for Weld Wire Heat 61782, Applicable to Point Beach Unit 1 Cu Ni

    • I"8 Ad sted P RT Capsule Designation 1,

Ad us d wt% wt% (F A nor, ggy,,,,(op) ct F ,

(T Ginna: Capsule V Plant 149.2 0.24 0.52 161.4 545 146 153 Specific RVSP Material Ginna: Capsule R Plant 0.24 0.52 161.4 545 167 174 169.7 Specific RVSP Material Ginna: Capsule T Plant 171.6 0.24 0.52 161.4 545 169 176 Specific RVSP Material l

Ginna: Capsule S Plant 274.3 0.24 0.52 161.4 545 223 230 Specific RVSP Material _

Point Beach Unit 1:

0.23 0.52 157.4 538 Vessel Average

e NPL 99-0555 Attachment 2 PAIE 10 OF_1Q_

j' CALCULATION SHEET

. , CALCULATION / ADDENDUM N3. 98-0156-00-A TITI.E Evaluation of New Surveillance Data on chemistry MADE BY J.R. Pfefferi DATE 9/22/99

- Factor for Weld Wire Heat 61782. Point Beach Unit 1 REV'D. BY T.D. Sorv \ _DATE 9/22/09 The best-fit line is determined relating the "tempersture and ratio" adjusted ARTuor to the fluence factor (determined from the capsule fluence). As shown in Table 5, the slope of this best fit line is the chemistry factor calculated from surveillance data, CF3 ,om.  ;

Table 5. Temperature-Ratio Adjusted Best-Fit Line, Point Beach Unit 1 Relative to R.E.

~ Ginna

"*" P '

Capsule ID Fluence actor x T- (Fluence g0 Ad ed (including source) * "

n/cm') ARTuor, ('F)

ART Ginna: Capsule V, SA-1036:

0.5028 0.8081 149.2 120.6 0.6531 Plant Lpecific RVSP Material Ginna: Capsule R, SA-1036:

1.105 1.028 169.7 174.4 1.057 Plant Specific RVSP Material Ginna: Capsule T, SA-1036:

1.864 1.171 171.6 200.9 1.370 Plant Specific RVSP Material Gin la: Capsule S, SA 1036:

3.746 1.342 224.3 301.0 1.801 Plant Specific RVSP Material SUM: 796.9 4.880 Slope of Best-Fit Line = 796.9 + 4.880 = 163.3 Note (A1: The dimensionless Fluence Factor (FF) calculated in accordance with Regulatory Guide 1.99 Rev. 2, Equation 2: FF = f m.2s-oao8 8, where f is the fluence in units of E19 n/cm 2. For example, for Capsule V above, FF = 0.5028 5 28 0A0 'on o.s am ,

0.8081.

CF... o,. = 163.3*F I

Results and

Conclusions:

l In accordance with the NRC guidelines provided at the November 12, 1997 meeting, the surveillance data for weld wire heat 61782 was evaluated with respect to the Point Beach Unit 1 ,

reactor vessel. On,y the R.E. Ginna plant-specific weld metal surveillanca data was used in the assessment because it requires the least amount of adjustments to the measured data. The plant-specific weld metal surveillance data from R.E. Ginna was determined to be credible with respect to the five criteria specified.in both Regulatory Guide 1.99, Revision 2 and 10 CFR

. 50.61. Using the R.E. Ginna plant-specific weld metal data and making the required adjustments to the data, the surveillance data chemistry factor was calculated to be 163.3*F.

Using this chemistry factor for weld wire heat 61782, it was determined by inspection that the lirniting beltline material determined in Calculation 98-0156 Revision 0 for Point Beach Unit 1 would not be affected.

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