ML20154B810
| ML20154B810 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 09/23/1998 |
| From: | Pfefferle J, Spry T, Walther V WISCONSIN ELECTRIC POWER CO. |
| To: | |
| Shared Package | |
| ML20154B808 | List: |
| References | |
| 98-0132, 98-0132-R00, 98-132, 98-132-R, NUDOCS 9810060018 | |
| Download: ML20154B810 (25) | |
Text
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NUCLEAR POWER BUSINESS UNIT Page 1 of 25 CALCULATION DOCUMENT FORM i
Calculation Number-Title of Calculation:
l 98-0132, Revision 0 Evaluation of 23.6 EFPY P-T Limit and LTOP Applicability Date and Pressurized l
Thermal Shock at 32 EFPY j
O OriginalCalculation g Supersedes Calculation # N-92-066, Rev. 0
- N-92-067, Rev. O O Revised Calculation. Revisian #
g QA Scope Governina Calculationt N/A g
Titic; N/A Associated Modification or Procedure:
N/A Illica N/A 1
This Calculation has been reviewed in accordance with NP 7.2.4. The review was accomplished by one or a combination of the following (as checked)-
_l A detailed review of the original A review of a representative sample of repetitive
_ calculations.
calculation.
A review of the calculation against a similar A review by an alternate, simplified, or j
_ calculation previously performed.
approximate method of calculation.
Page Inventory:
Page I - 4 Form PDF-1608 Attachments:
Page 5 - 17 Calculation Framatome Technologies, Inc. letter FTI-98-2563," Reevaluation of Weld Wire Heat 61782," August 26,1998:
Pages 18 through 25 Prepared By:
Date:
Thomas D. Snry f-Signature 1
Reviewed By:
Date:
f[7T James R. Pfefferie Sigr%ture
///
i
//
Dals:
Approved By:
Victoria A. Walther Signature 9810060018 980928 II POR ADOCK 05000266 Reference (s): NP 7.2.4 P
PDR g i
NUCLEAR POWER BUSINESS UNIT Calculation *:93-0132, Rev. O CALCULATION DOCUMENT FORM Page 2 or25 Preparer: T.D. Spry TDS Date: 9/15/98 Calcu'latinn. Checklist (Optional for Non-QA Scope)
Item Attribute Description N/A Author Reviewer No.
1.
Purpose a.
Is the purpose clearly stated indicating Qyes Ono issue to be resolved or information to be determined?
2.
Methodology and Acceptance Criteria a.
Has the method / approach been
[gyes Ono y
described?
W Q[yes Ono b.
Have appropriate acceptance criteria and their sources been identified?
3.
Assumptions a.
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7 b.
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c.
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References
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b.
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y 4.
Inputs a.
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ges Ono been identified?
a 6.
Calculation a.
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gyes Ono consistent with the source document, including engineering units?
o 7.
Computer-Aided Design Calculations (NP l
7.2.4 Attachment A) a.
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Oyes Ono per the requirements of Attachment A?
/V/A-b.
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Oyes Ono been identified on the computer run and in y/4 the calculation?
I FBF-1608 Revision 2 01/07/98 Reference (s): NP 7.2.4 i
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NUCLEAR POWER BUSINESS UNIT Calculation =:98-0132. Rev. O CALCULATION DOCUMENT FORM Page 3 of 25 Preparer: T.D. Spry TOS Date: 9/15/98 Item Attribute Description N/A Author Reviewer No.
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' 8.
Summary of Results and Conclusions 39 a.
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CR 98-2340 d.
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l PDF 1608 f.'
Revision 2 01/07/98 Reference (s): NP 7.2.4
(
1 l
NUCLEAR POWER BUSINESS UNIT Calculation #:98-0132, Rev. O CALCULATION DOCUMENT FORM Page 4 of 25 Preparer: T.D. Spry 7Dk Date: 9/15/98 Item Attribute Description N/A Author Reviewer No.
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w COMMENTS AND RESOLUTION 1
Reviewer Comments:
Resolution:
ovk WS^
dwNT' WW No S
9/shr I
l PDF-1608 Revision 2 01/07m Reference (s): NP 7.2.4
_m CALCULATION SHEET CALC. NO.98-0132. Rsv _0 115 TITLE Evalp_a_tle.r.tpL23.6 EFPY P-T Limit and LTOP Applicability Date and MADE BY T.D. Syrv DATE 9/15/98 PrestpAzed Thermal ShoqA_all2 EFPY REV'D. BY J.R. Pfefferie DATE 9/15/98 l
l
Purpose:
This calculation provides an evaluation of Unit 1 and Unit 2 vessel beltline material adjusted reference temperatures (ARTS) and their inputs, currently used as input to P-T limits (in Calculation #N-94-058, Revision 2, with a stated applicability date of 23.6 EFPY) and LTOP setpoints (in Calculation No. 98-0046, Revision 0, calculated for the fluence projected as of January 1, 2001). The latest available best-estimato chemistry values and all applicable surveillance data for all vessel beltline materials will be utilized. A revised applicability date in terms of effective full power years (EFPY) is provided for the most i
limiting vessel, Unit 2.
This calculation also provides a pressurized thermal shock evaluation for Unit 1 and 2 at 32 EFPY in accordance with 10 CFR 50.61, using the latest available best-estimate chemistry values and all applicable surveillance data for all vessel 4
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beltline materials.
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Table of Contents:
0 j
Purpose 5
References 6
Methods & Acceptance Criteria 6
Assumptions 7
Inputs -
7 Calculations 1.
(see table below) 7 Table Table Description Page Number Number Table 1.
Point Beach Unit 1 RPV Bettline 23.6 EFPY 8
Fluence Values Table 2.
Point Beach Unit 2 RPV Beltline 23.6 EFPY 9
Fluence Values Table 3.
Point Beach Unit 1 RPV 1/41 Beltline Material 10 Adjusted Reference Temper.stures at 23.6 EFPY Table 4.
Point Beach Unit 2 RPV 1/4t Beltline Material 11 Adjusted Reference Temperatures at 23.6 EFPY Table 5.
Point Beach Unit 1 RPV 3/4t Beltline Material 12 Adjusted Reference Temperatures at 23.6 EFPY Table 6.
Point Beach Unit 2 RPV 3/4t Beltline Material 13 Adjusted Reference Temperatures at 23.6 EFPY l
Table 7.
Point Beach Unit 1 RPV Beltline Material RTris 14 Values at 32 EFPY Table 8.
Point Beach Unit 2 RPV Beltline Material RTrTs 15 Values at 32 EFPY ll.
Fluence Projection to January 2001 16 l
Results and Conclusions 16 Attachments - FTl Letter FTI-98-2563, August 26,1998 17
~-
= _. _
~.. -. - _ -.
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CALCULATION SHEET CALC. NO.98-0132. Rtv. O "1T)5 TITLE Evaluation of 23.6 EFPY P-T Limit and LTOP Applicability Date and MADE BY T.D. Spry DATE 9/15/98 Pressurized Thermal Shock at 32 EFPY REV'D. BY J.R. Pfefferle DATE 9/15/98
References:
1.
Instruction Manual 132-Inch 1.D. Reactor Pressure Vessel, Babcock & Wilcox, September 1969.
2.
Instruction Manual, Reactor Vessel, Point Beach Nuclear Plant No. 2, Combustion Engineering, CE Book #4869, October 1970.
i 3.
NRC Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials," May 1988.
4.
10 CFR 50.61, " Fracture toughness requirements for protection against pressurized thermal shock events."
5.
WCAP 12794, " Reactor Cavity Neutron Measurement Program for Point Beach Unit 1," Rev. 3, December 1995.
6.
WCAP-12795, "Heactor Cavity Neutron Measurement Program for Point Beach Unit 2," Rev. 3, August 1995.
7.
CEOG Report "Best Estimate Copper and Nickel Values in CE Fabricated Reactor Vessel Welds," CE NPSD-1039, Revision 2, Final Report, June 1997.
8.
BAW-2325, " Response to Request for Additional information (RAJ Regarding Reactor Pressure Vessel Integrity," May 1998.
9.
Framatome Technologies, Inc. Letter FTI-98-2563, " Reevaluation of Weld Wire Heat 61782," August 26,1998 (see attachment).
- 10. ASTM E 29-93a, " Standard Practice for Using Significant Digits in Test Data to Determine Conformance with Specifications."
- 11. WEPCO Letter NPL 98-0159, " Monthly Operating Reports," March 6,1998.
- 12. Point Beach Nuclear Plant Unit Nos.1 and 2 Final Safety Analysis Report.
- 13. WEPCO Calculation No. N-94-058, Revision 2, " Reactor Coolant System Calculations
- Effective Through January 2001," 1/23/96.
- 14. WEPCO Calculation No. 98-0046, Revision 0, " Determination of PBNP LTOP Setpoint Using ACME Code Case N-514 (applicable through appx Jan. 2001)," 4/27/98.
Methods & Acceptance Criteria:
The methodology of this calculation follows the steps listed below:
Using References 5 and 6, determine tne projected fluence for the limiting reactor
. vessel materials at the reactor vessel inside surface (clad-base metal interface) at 23.6
' EFPY. See Tables 1 and 2 for a description of the methodology used.
l II. Determine the corresponding reactor vessel fluence and fluence factors at the one-fourth thickness (1/4T) and three-fourths thickness (3/4T) location from the clad-base metal interface. See Tables 1 and 2 for a description of the methodology used.
l 111. Determine the chemistry factor, initial properties, and margin term for the PBNP reactor vessel beltline materials using the latest available best-estimate chemistry values and i
all applicable surveillance data for all vessel beltline materials.
i.
IV. Determine the projected adjusted reference temperature at the 1/4T and 3/4T location for the reactor vessel beltline materials at 23.6 EFPY. See Tables 3 through 6 for a description of the methodology used.
CALCULATION SHEET CALC. NO.98-0132. R5v. 0 Y
TITLE fyJLu_alon of 23.6 EFPY P-T Limit and LTOP Applicability Date and MADE BY T.D. Sorv DATE 9/15/96 Pressurized Thermal Shock at 32 EFPY REV'D. BY J.R. Pfefferle DATE 9/15/98 V. Compare the projected ARTS at the 1/4T and 3/4T location for the reactor vessel beltline materials to the values used for the calculation of current licensing basis P-T limits. If the projected ART values at a given thickness are less than the values used for the calculation of current P-T limits, current P-T limits are conservative.
If the projected ART values at a given thickness are greater than the values used for the calculation of current licensing basis P-T limits, determine the applicability date using the latest available best-el stimate chemistry values and all applicable surveillance data for the most limiting vessel beltline material. See Tables 4 and 6 for a description of the methodology used.
VI. Compare newly determined applicability dates against the EFPY projected to occur through January 1, 2001. If the newly determined applicability date is beyond the projected value on January 1, 2001, current licensing basis P-T limits, LTOP pressure setpoints, and Tenable temperature are acceptable through that date.
PTS Evaluation Vll. Using References 5 and 6, determine projected fluence for the limiting reactor vessel materials at the reactor vessel inside surface (clad-base metal interf ace) at 32 EFPY.
Vill. Determine the corresponding reactor vessel fluence factors at the clad base metal interface. See Tables 7 and 8 for a description of the methodology used.
IX.
Using the chemistry factor, initial properties, and margin term for the PBNP reactor vessel beltline materials identified in Step til (above), determine the projected RTris i
value at the reactor vessel inside surface at 32 EFPY. See Tables 7 and 8 for a description of the methodology used.
X.
Compare the projected RTeis value to the PTS Screening Criteria of 10 CFR 50.61.
If the projected RTris value is less than the screening criteria, no submittal to NRC i
pursuant to 10 CFR 50.61 is required.
Assumptions:
- 1. A capacity factor of 80% is assumed for future plant operation. This is a standard industry value for reactor vessel evaluations and will be verified against future plant performance.
- 2. Changes in core design will not increase vessel inner-surface fluence beyond the values provided at specific vessel altitude and azimuth locations in WCAPs 12794, j
Rev. 3 and 12795, Rev. 3.
i
- 3. Reactor vessel thickness is based on the distance from the clad-base metal interface to the outside diameter.
1 Inputs:
Specified in Tables 1 through 8.
Calculations:
1.
Fluence, Adjusted Reference Temperature (ART), and Pressurized Thermal Shock (PTS) Evaluations (Tables 1 through 8)
N Date: 9/15/98 Page 8 of 25 Calculation Number: 98-0132. Rey, O Preparer: T.D. Spry Tcble 1.
Print Bench Unit 1 RPV Beltline 23.6 EFPY Fluencs Vtlues Based on WCAP-12794," Reactor Cavity Neutron Measurement Program For Wisconsin Electric Power Company Point Beach Unit 1,* Rev. 3, December 1995. Note that the estimated fluence at a specific point in time is not linearly interpolated between zero and the estimated fluence at 32 EFPY, due to changes in core design at certain points in the operating history of the urut. As intermedtate input to further calculations, these values are not rounded in accordance with ASTM E29.
Yesset Manufacturer:
Babcock & Wilcox Plate and Weld Thickness (without claddina)'
6 5' without clad (D)
Component Description Heat or 32 EFPY 23.6 EFPY 23.6 EFPY 23.6 EFPY 23.6 EFPY 23.6 EFPY Heat / Lot inside Surface Inside Surface 1/4T 1/4T 3/4T 3/4T Fluence (E19 Fluence Fluence Fluence Fluence Fluencej)E19 n/cm n/cm )(A)
(E19 n/cm')
Factor (C)
(E19 n/cm')
Factor (C) 2 (B)
(B)
Nozzle Bett Forging 122P237 0.547 0.4166 0.2821 0.6545 0.1293 0.4702 Intermediate Shen Plate A9811-1 2.91 2.22 1.503 1.113 0.689 0.8955 Lower Shelt Plate C1423-1 2.46 1.971 1.335 1 08 06117 0.8623 Nozzle Belt to Intermed.
8T1762 (SA-0547 0.4166 0.2821 0 6545 0.1293 0.4702 Shell Circ Weld (100%)
1426)
Intermediate Shell Long 1P0815 (SA-1.87 1.425 0.9648 0 99 N/A N/A Seam (ID 27%)
812)
Intermediate Shell Long 1P0661 (SA-1.87 1.425 N/A N/A 0.4423 0.7731 Seam (OD 73%)
775)
Intermed. to Lower Sheu 71249 (SA-2.46 1.971 1.335 1.08 0.6117 0.8623 Circ. Weld (100%)
1101)
Lower Shell Long Seam 61782 (SA-1.66 1.309 0.8863 0.9662 0.4063 0.7502 (100%)
847)
Footnotes:
2 (A)
Interpolation of neutron exposure (in units of E19 n/cm, E>1 MeV) to a particular value of effective fuit power years (EFPY) is performed based on WCAP-12794, Revision
- 3. For example, for the nozzle belt forging, beat no.122P237, 0.4166 E19 n/cm' fluence = 0.339 + f 0 547 - 0 339
) x (23.6 EFPY-18 6 EFPY)
=
From an inside surface fluence value (not including cladding), fluence is attenuated to a desired thickness using equation (3) of Regulatory Guide 1.99, Revision 2. f = f.,,,
x e* *", where fuis expressed in units of E19 n/cm', E > 1MeV, and x is the desired depth in errhes into the vessel wall. For exampie, for the nozzle belt forging, heat no 122P237, at 23 6 EFPY, at a depth of % of the 6.5* vessel wall (1.625"), f = 0.4166 x e m2s = 0.2821 E19 n/cm'.
(C)
The dimensionless fluence factor is calculated using the fluence factor formula from equation (2) of Regulatory Guide 1.99, Revision 2: ff = fM 28 - tie soo 9, where f is the fluence in units of E19 n/cm. For example, the 23.6 EFPY 1/4T fluence factor for nozzle belt forging, heat no.122P237, ff = 0.2821 *** -* *** ***2" = 0.6545.
2 (D)
Instruction Manual,132-inch i D. Reactor Pressure Vessel, Babcock & Wilcox. September 1969.
Calculation Number
- 98-0132 RevJ Preparer: T.D. Sory 'TD5 Date: 9/15/98
.Page 9 ofE Tcble 2.
Point Beach Unit 2 RPV Beltline 23.6 EFPY Fluence Values Based on WCAP-12795," Reactor Cavity Neutron Measurement Pr'ogram For Wisconsin Electric Power Company Point Beach Unit 2,* Rev. 3, August 1995. Note that the estimated fluence at a specific point in time is not linearty interpolated between zero and the estimated fluence at 32 EFPY, due to changes in core design at certain points in the operating history of the unit. As intermedste input to further calculations, these values are not rounded in accordance with ASTM E29.
Vessel Manufacturer-Babcock & Wilcox and Combust on Engineent:a Plate and Weld Theckness (without claddina)-
6 5*, without clad (D)
Component Description Heat or 32 EFPY 23.6 EFPY 23.6 EFPY 23.6 EFPY 23.6 EFPY 23.6 EFPY Heat / Lot inside inside il4T il4T 3/4T 3/4T Surface Surface Fluence Fluence Fluence Fluence FluenceJE19 Fluence (E19 (E19 nlcm')
Factor (C)
(E19 nlcm')
Factor (C) nicm I nicm*) (A)
(B)
(B)
Nozzle Beit Forging 123V352 0.548 0.4244 0.2873 0 6592 0 1317 0.4742 Intermediate Shelt Forging 123V500 3.01 2.322 i.572 1.125 0.7207 0.9081 Lower Shell Forging 122W195 2.52 2.015 1.364 1.086 0 6254 0 8685-Nozzle Belt to Intermed.
21935 0.548 0.4244 0.2873 0.6592 0.1317 04742 Shell Circ Weld (100%)
Intermed. to Lower Shell 72442 (SA-2.49 2.003 1.356 1.085 0 6217 0 8668 Cire. Weld (100%)
1484)
Footnotes (A)
Interpolation of neutron exposure (in units of E19 nlcm. E>1 MeV) to a particular value of effective full power years (EFPY) rs performed based on WCAP-12795. Revison
- 3. For example, for the nozzle belt forging, heat no.123V352, fluence = 0 345 + i 0 548 - 0 345
) x (23 6 EFPY-18.2 EFPY) = 0 4244E19 nlcm'
(B)
From an inside surface fluence value (not including cladding), fluence is attenuated to a desired thickness using equation (3) of Regulatory Gu.de 1.99, Revision 2: f = f,s x e* **", where f.,is expressed in units of E19 n/cm'. E > 1MeV, and x is the desired depth in inches into the vessel wall. For example, for the nozzle belt forging. heat a
no.123V352, at 23.6 EFPY, at a depth of % of the 6.5" vessel wall (1.625*), f = 0 4244 x e4 av s2" = 0 2873 E19 n!cm,
(C)
The dimensionless fluence factor is calculated using the fluence factor formula from equation (2) of Regulatory Guide 1.99, Revision 2-ff = f w as-e io n.e o, where f is the fluence in units of E19 n/cm', For example, the 23.6 EFPY 1/4T fluence factor for nozzle belt forging, heat no.123V352, ff = 0 2873 p as-o seios o msra = 0 6592.
(D) instruction Manual, Reactor Vessel, Point Beach Nuclear Plant No. 2 Combustion Engineering, CE Book #4869, October 1970.
Calculation Number: 98-0132, Rev. O Preparer: T.D. Spry TM3 Date: 9/15/98 Page 10 of 25 Tcbb 3.
Point Be:ch Unit 1 RPV 1/4T Beltline M teri:1 Adju;ted R;ferenca Tcmperctures at 23.6 EFPY Unless otherwise noted, all ART input dats obtained frorn BAW-2325, " Response to Request for Additional Information (RAI) Regardeng Reactor Pressure Vessel integnty," May 1998, including the most recent best-estimate chemistry values for welds, applying current B&WOG mean-of-the-sources approach. All be!!hne materiats are included for companson.
Vessel Manufacturer-Babcock & Wilcox Plate and Weld Thickness (without cladding)-
6 5*, without clad (F)
Component Description Heat or initial RTuor
%Cu
%Ni CF CF 114T 23.6 ARTuor oi 03 Margin ART ("F)
Heat / Lot
(*F)
Method EFPY Fluence
(*F)
(*F)
(E)
Factor (A)
Nozzle Belt Forgino 122P237
+50 0 11 0 82 77 Table 06545 50 4 0
17 34 134 intemnate Shell Plate A9811-1
+1 0 20 0 06 88 Table 1113 97.94 26 9 17 63 64 163 79.3 Surv. Data 88.26 85 56 42 146 (B)
Lower Shell Plate C1423-1
+1 0 12 0 07 55 3 Table 1 08 59 74 26 9 17 63 64 124 35.8 Surv. Data 38 67 8.5 56 42 96 (B)
Nozzle Bett to Intermed Shell 8T1762 (SA-
-5 0.19 0 57 152.4 Table 06545 99.75 19 7 28 68 47 163 Circ Weld (100%)
1426)
Intermediate Shell Long Seam IP0815 (SA-
-5 0.17 0.52 138.2 Table 0.99 136 82 19.7 28 68 47 200 (ID 27%)
812)
Int rmediate Shell Long Seam 1P0661 (SA-
-5 0.17 0 64 157.6 Table N/A N/A 19 7 N/A N/A N/A (OD 73%)
775)
Int rmed To Lower Shell Circ.
71249 (SA-
+10 0.23 0.59 167.6 Table (C) 1.08 181.01 0
28 56 247 (G)
Weld (100%)
1101)
Lower Shell Long Seam 61782 (SA-
-5 0.23 0.52 157.4 Table 0.9662 152.08 19 7 28 68 47 216 (100%)
847) 161.1 Surv. Data 155 65 14 48 34 199 (D)
Footnotes:
(A)
See Table 1.
(B)
Credible Surveillance Data, see BAW-2325 for evaluation..
(C)
Non-credible surveillance data; see BAW-2325 for evaluation. Table CF conservative because difference between ratio-adjusted measured ARTuor and predicted ARTuor based on Table CF is less than 2a (56*F).
(D)
Credible Surveillance Data; see Framatome Technologies, Inc. letter FTI-98-2563, August 26,1998, " Reevaluation of Weld Wire Heat 61782,* utilizing latest time-weighted temperature data for Point Beach Unit 1, which supersedes BAW-2325.
(E)
Adjusted reference temperaturej) ART) calculated per Regulatory Guide 1.99, Rev. 2. ART = Initial RTwor + ARTuor + Margrn, where ARTuor =
Factor, and Margin = 2(o,' + o, "', with a defined as the standard deviation of the initial RTuor, and oa defined as the standard deviation of ARTuor. For example, for s
nozzle belt forging, heat no.122P237 ART = 50 +(77 x 0.6545) + 34 = 134*F. Calculated ART values are rounded to the nearest *F in accordance with the rounding-off method of ASTM Practice E29.
(F)
Instruction Manual,132-inch 1 D Reactor Pressure Vessef, Pabcock & Wilcox, September 1969.
(G)
By inspection, for these limiting material properties, the amicability date of 23.6 EFPY for P-T limits in Calculation N-94-058 Revision 2 based on a 1/4T ART of 258 4*F is conservative for Unit 1, since the newly calculated value of 247*F is lower.
Calculation Number-98-0132. Rev. O Preparer: T.D. Spry TOS Date: 9/15/98 Page 11 of 25 Tcble 4.
Paint Be:ch Unit 2 RPV 1/4T Beltline Materiil Adjunted R;ferenca Temper;tures ct 23.6 EFPY Unless otherwise noted, all ART input data obtained from BAW-2325, " Response to Request for Additional Information (RAl) Regarding Reactor Pressure Vessel integnty" May 1998, including the most recent best-estimate chemistry values for welds, applying current B&WOG mean-of-the-sources approach All belthne materials are included for comparison.
Vessel Manufacturer:
Babcock & Wilcox and Combustion Engineenna Plate and Weld Thickness (without claddingP 6 5", without clad (F)
Component Description Heat or initial RTwor
%Cu
%Ni CF CF 1/4T 23.6 ARTuor.
o, o,
Margin ART (*F)
Heat / Lot
(*F)
Method EFPY Fluence
(*F)
(*F)
(E)
Factor (A)
Nozzle Belt Forgina 123V352
+40 0 11 0 73 76 Table 06592 50 1 0
17 34 124 fntermediate Shell Forgina 123V500
+40 0 09 0 70 58 Table (B) 1 125 65 25 0
17 34 139 Lower Shell Foraino 122W195
+40 0 05 0.72 31 Table 1 086 33 67 0
17 34 108 42.8 Surv. Data 46 48 8.5 17 103 (C)
Nozzle Belt to Intermed. Shell 21935
-56 0.18 0.70 170 Table (H) 0.6592 112.1 17 28 65 51 122 Cire Weld (100%)
Int:rmed. To Lower Shell Cire.
72442 (SA-
-5 0.26 0 60 180 Table (D) 1.085 195.3 19.7 28 68 47 259 (G)
Weld (100%
1484)
Footnotes:
(A)
See Table 2.
(B)
Non-credible surveillance data; see BAW-2325 for evaluation. Table CF conservative because difference between measured ARTuor and predicted ARTwor based on
' Table CF is less than 2o (34' ).
(C)
Credible surveillance data; see BAW-2325 for evaluahon.
(D)
Non-credible surveillance data; Table CF value based on best-estimate chemistry is higher than best fit calculated using surveillance data, and therefore consercative (E)
Adjusted reference temperature j) ART) calculated per Regulatory Guide 1.99, Rev. 2. ART = Initial RTwor + ARTwor + Margin, where ART or = Chemi N
Factor, and Margin = 2(o,' + o * *. uth o, defined as the standard deviation of the initial RT or, and ca def:ned as the standard deviation of ARTuor. For example, for h
nozzle belt forging. heat no.123V352, ART = 40 +(76 x 0 6592) + 34 = 124*F. Calculated ART values are rounded to the nearest *F in accordance with the rounding-off method of ASTM Practice E29 (F) instruction Manual, Reactor Vessel, Point Beach Nuclear Plant No. 2. Combustion Engineering, CE Book #4869. October 1970.
(G)
For these limiting matenal properties, based on Reg. Guide 1.99 Rev. 2 and Table 2, Footnote (A), the apphcability date for P-T hmits in Calculation N-94-058, Revision 2, based on a 1/4T ART of 258.4*F would be:
258 4*F = Initial RTuor + ARTwor + Margin = -5 + (180 x Applicable EFPY 1/4T Fluence Factor) + 68 47 Applicable EFPY 1/4T (1.625") Fluence Factor = 1.0829444 = (1/4T f) 8 * **8 "3 1/4T f = 1.3475 = f.wx e* **,'
f s = 1.9902 = 1.69 + 2 49-169 x (Applicable EFPY - 18.2) 32 - 18.2 Applicable EFPY = 23.4 (H)
Table CF value based on best-estimate chemistry data from CEOG Report *Best Estimate Copper and Nickel Values in CE Fabricated Reactor Vessel Welds " CE NPSD-1039, Revision 2, Final Report, June 1997.
Calculation Number: 98-0132. Rev. O Preparer: T.D. Sprv ~Tb5 Date: 9/15/98 Page 12 of 25 Tcbla 5.
P: int Be:ch Unit i RPV 3/4T Beltlina M;teri:1 Adju;ted R;ferenca Temperatures ct 23.6 EFPY Unless otherwise noted, all ART input data obtained from BAW-2325, " Response to Request for Additional Information (RAl) Regarding Reactor Pressure Vessel Integnty," May 1998, including the most recent bestestimate chemistry values for welds, applying current B&WOG mean-of-the-sources approach. All belthne matenals are includad for comparison.
V sset Manufacturer-Babcock & Wilcox Plate and Weld Thickness (without claddinq)-
6 5*, without clad (F)
Margin ART (*F)
Component Description Heat or Initial RTuor
%Cu
%Ni CF CF 3/4T 23.6 ARTuor a,
ca Heat / Lot
(*F)
Method EFPY Fluence
(*F)
(*F)
(E)
Factor (A)
Nozzle Belt Forging 122P237
+50 0 11 0 82 77 Table 04702 36 21 0
17 34 120 Inttrmediate Shell Plate A9811-1
+1 0 20 0 06 88 Table 08955 78 8 26 9 17 63 64 143 79.3 Sury. Data 71.01 85 56 42 128 (B)
Lower Shell Plate C1423-1
+1 0 12 0 07 55 3 Table 08623 47 69 26 9 17 63 64 112 35 8 Surv. Data 30 87 85 56 42 88 (B)
Nozzle Belt to intermed. Shell 8T1762 (SA-
-5 0.19 0.57 152.4 Table 0.4702 71.66 19.7 28 68 47 135 Cire Weld (100%)
1426)
Intirmediate Shell Long Seam 1P0815 (SA-
-5 0.17 0 52 138 2 Table N/A N/A 19.7 N/A N/A N/A (ID 27%)
812)
Intirmediate Shell Long Seam 1P0661 (SA-
-5 0.17 0 64 157 6 Table 0.7731 121.84 19.7 28 68 47 185 (OD 73%)
775)
Intirmed To Lower Shell Circ.
71249 (SA-
+10 0.23 0 59 167 6 Table (C) 08623 144.52 0
28 56 211 (G)
Weld (100%)
1101)
Lower Shell Long Seam 61782 (SA-
-5 0.23 0.52 157.4 Table 07502 118 08 19 7 28 68 47 182 (100%)
847) 161.1 Sury. Data 120 86 14 48.34 164 (D)
Footnotes:
(A)
See Table 1.
(B)
Credible Surveillance Data; see BAW-2325 for evaluation.
(C)
Non-credible suiveillance data; see BAW-2325 for evaluation. Table CF conservative because difference between ratro-adjusted measured ARTuor and predected ARTuor based on Table CF is less tnan 20 (56*F).
(D)
Credible Surveillance Data, see Framatome Technologies, Inc, letter FTI-98-2563, August 26,1998,
- Reevaluation of Weld Wire Heat 61782.* utilizing latest tirne-weighted temperature data for Point Beach Unit 1, which supersedes BAW-2325.
Adjusted reference temperature j) ART) calculated per Regulatory Guide 1.99, Rev. 2. ART = Initial RTuor + ARTuor + Margin, where ARTu (E)
Factor, and Margin = 2(o," + oa ", with o, defined as the standard deviation of the Initial RTuor, and oa defined as the standard deviation of ARTuor. For example, for nozzle belt forging heat no.122P237, ART = 50 +(77 x 0.4702) + 34 = 120*F. Calculated ART values are rounded to the nearest *F in accordance with the rounding.otf method of ASTM Practice E29.
(F) instruction Manual,132-inch i D. Reactor Pressure Vessel. Babcock & Wilcox, September 1969.
(G)
By inspection, for these limiting material propertes, the apphcabihty date of 23 6 EFPY for P-T hmits in Calculation N-94-058 Revision 2 based on a 3/4T ART of 219 5' F is conservative for Unit 1, since the newly calculated value of 211*F is lower.
Calculation Number-98-0132. Rev. O Preparer: T.D. Spry TD)
Date: 9/15/98 Page 13 of 25 Tcbla6.
Point Be:ch Unit 2 RPV 3/4T Beltline Materi:1 Adju;ted R:.ferenc:a Tcmper;tures ct 23.6 EFPY Unless otherwise noted, all ART input data obtained from BAW-2325,
- Response to Request for Additional;nformation (RAl) Regarding Reactor Pressure Vessel integnty," May 1998, including the most recent best-estimate chemistry values for welds, applying current B&WOG mean-of-the-sources approach All beltline materials are included for companson.
Vesse! Manufacturer:
Babcock & Wilcox and Combustion Enoineenno Plate and Weld Thschness (without cladding)'
6 5", without clad (F)
Component Description Heat or initial RTar
- /.Cu
%Ni CF CF 3/4T 23.6 ARTer Margin ART ("F) as om Heat / Lot
(*F)
Method EFPY Fluence
(*F)
(*F)
(E)
Factor (A)
Nozzle Belt Forging 123V352
+40 0 11 0 73 76 Table 04742 36 04 0
17 34 110 intermediate Sheff Forging 123V500
+40 0 09 0 70 58 Table (B) 09081 52 67
-0
- 17 34 127 Lower Shell Foroino 122W195
+40 0 05 0 72 31 Table 08685 26 92 0
17 34 101 42.8 Surv. Data 37.17 85 17 94 (C)
No221e Belt to Intermed. Shelt 21935
-56 0.18 0.70 170 Table (H) 0.4742 80 61 17 28 65.51 90 Cire Weld f100%)
Intermed. To Lower Shell Cire.
72442 (SA-
-5 0.26 0 60 180 Table (D) 08668 156 02 19.7 28 68 47 219 (G)
Weld (100%)
1484)
Footnotes:
(A)
See Table 2.
(B)
Non-credible surveillance data; see BAW-2325 for evaluation. Table CF conservative because difference between measured ARTerand predicted ARTer based on Table CF is less than 2a (34*F).
(C)
Credible surveillance data; see BAW-2325 for evaluation.
(D)
Non-credible survestlance data; Table CF value based on best-estimate chemistry is higher than best fit calculated using surveillance data, and therefore conservative (E)
Adjusted reference temperature (ART) calculated per Regulatory Guide 1.99, Rev. 2. ART = Init al RTer + ARTer + Margin, where ARTer = Chemistry Factor x Fluence Factor, and Margin = 2(ci + ca ) ', with cidefined as the standard deviation of the Initial RTer, and a, defined as the standard deviation of ARTer. For example, for nozzle belt forging, heat no.123V352, ART = 40 +(76 x 0 4742) + 34 = 110"F. Calculated ART values are rounded to the nearest *F in accordance with the rounding-off method of ASTM Practice E29.
(F)
Instruction Manual, Reactor Vessel, Point Beach Nuclear Plant No. 2. Combustion Engineering, CE Book #4869, October 1970.
(G)
By inspection, for these kmiting material properties, the applicability date of 23 6 EFPY for P-T limits in Calculation N-94-058 Revision 2 based on a 3/4T ART of 219 5'F is conservative for Unit 2, since the newly calculated value of 219"F is lower.
(H)
Table CF value based or! best-estimate chemistry data from CEOG Report "Best Estimate Copper and Nickel Values in CE Fabricated Reactor Vessel Welds." CE NPSD-1039. Revision 2, Final Report, June 1997.
Cticulation Number: 98-0132. Rev. O Preparer: T D. Spry M Date: 9/15/98 Page 14 of 25 Tcble 7.
P; int BerCh Unit 1 RPV Beltiins M:teri 1 RT Vclues ct 32 EFPY PTS Unless otherwise noted, all RTers input data obtained from BAW-2325, " Response to Request for Additional information (RAI) Regarding Reactor Pressure Vessel Integnty," May 1998, including the most recent best+ stimate chemistry values for welds, applying current B&WOG mean-of-the-sources approach. All bekline materials are inctoded for companson.
Vvsset Manufacturer:
Babcock & Wilcox Plate and Weld Thickness (without cladding)-
6 5*, without clad (G)
Margin RTers PTS Component Heat or RTuow
%Cu
%Ni CF CF 32 EFPY Inner 32 EFPY ARTPts ou oa Description HeatiLot
(*F)
Method Surface inner Surface
(*F)
(*F)
(*F) (F)
Screening Fluence (E19 Fluence Criteria 2
n/cm )(A)
Factor (B)
Nozzle Bett 122P237
+50 0 11 0.82 77 Table 0.547 0.8313 64.01 0
17 34 148 270 Fo@no int;rmediate Shell A9811-1
+1 0.20 0.06 88 Table 2.91 1.2834 112.94 26.9 17 63 64 178 270 Plate 101.77 26.9 85 56 42 159 270 79.3 Surv. Data (C)
Lower Shell Plate C1423-1
+1 0 12 0 07 55 3 Table 2 46 1 2422 68 69 26 9 17 63 64 133 270 44 47 26.9 8.5 56 42 102 270 35.8 Surv. Data (C)
Nozzle Belt to 8T1762 (SA-
-5 0.19 0.57 152.4 Table 0.547 0 8313 126 69 19.7 28 68.47 190 300 intermed. Shell 1426)
Cire Weld (100%)
intumediate Shell 1P0815 (SA-
-5 0.17 0.52 138.2 Table 1.87 1.1715 161.9 19.7 28 68.47 225 270 Long Seam (ID 812) 27%)
Intxrmediate Shell 1P0661 (SA-
-5 0.17 0.64 157.6 Table N/A N/A N/A 19.7 N/A N/A N/A N/A Long Seam (OD 775)
_7_3%)
Intermed. To Lower 71249 (SA-
+10 0 23 0 59 167.6 Table (D) 2.46 1.2422 208.19 0
28 56 274 300 Shell Circ. Weld 1101)
(100%)
l Lower Shell Long 61782 (SA-
-5 0.23 0.52 157.4 Table 1.66 1.1397 179.39 19.7 28 68 47 243 270 Seam (100%)
847) 161.1 Surv. Data 183 61 19 7 14 48.34 227 270 l
(E)
Footnotes:
(A)
Based on WCAP-12794, ' Reactor Cavity Neutron Measurement Program For Wisconsin Electric Power Company Point Beach Unit 1," Rev. 3, December 1995. As intermediate input to further calculations, these values are not rounded in accordance with ASTM E29 (B)
The dimensionless fluence factor ts calculated using the fluence factor fo Jia from equation (2) of Regulatory Guide 1.99, Revision 2: ff = fto 2s - e se eoo o, where f is the fluence in units of E19 n/cm. For exampie, the 32 EFPY inner surface fluence factor for nozzle belt forging, heat no.122P237, ff = 0 547 p 2e -a mas e sen = 0 8313.
2 (C)
Credible surveillance data; see BAW-2325 for evaluation.
(D)
Non-credible surveit:ance data; see BAW-2325 for evaluation. Table CF conservative because difference between ratio-adjusted measured ARTaor and predicted ARTuor based on Table CF is less than 2a (56*F).
(F)
Credible Surveillance Data; see Framatome Technologies, Inc. letter FTI-98-2563, August 26,1998 " Reevaluation of Weld Wire Heat 61782,* utshzing latest time-weighted temperature data for Point Beach Unit 1, which supersedes BAW-2325.
(F)
RTpyscalculated per 10 CFR 50 61. RTers = RTuorius + ART,,or + Margin, where RTuo, = unirradiated initial RTuor, ART ot = Chemistry Factor x Fluence Factor, and N
Margin = 2(a ' + c #)*
- with o defined as the standard deviation of the RTwortu>. and c. defined as the standard deviation of ARTuor. For example, for nozzle belt forging, u
u a
heat no.122P237. RTers = 50 +(77 x 0.8313) + 34 = 148*F. Calculated RTpysvalues are rounded to the nearest *F in accordance with the rounding-off method of ASTM Practice E29.
(G)
Instruction Manual,132-inch 1 D. Reactor Pressure Vessel, Babcock & Wilcox, September 1969.
Calculation Number: 98-0132. Rev. O Preparer: T.D. Spry TD)
Date: 9/15/98 Page 15 of 25 Tcble 8.
Paint Be:Ch Unit 2 RPV Beltline Materill RT s Vclues ct 32 EFPY PT Unless otherwise noted, all RTers input date obtained from BAW-2325, " Response to Request for Additional Information (RAI) Regarding Reactor Pressure Vessel Integrity," May 1998, including the most recent best-eshmate chemistry values for welds, applying current B&WOG mean-of-the-sources approach. All belthne matenals are included for companson.
V:;sset Manufacturer Babcock & Wilcox and Combusbon Engineenna
_ Piate and Weld Thickness (without cladding)'
6 5", without clad (G)
Component Heat or RTworcui
%Cu
%Ni CF CF 32 EFPY Inner 32 EFPY ARTpr.
ou n
Margin RTers PTS a
Description Heat / Lot
(*F)
Method Surface inner Surface
(*F)
("F)
(*F) (F)
Screening Fluence (E19 Fluence Criteria n!cm') (A)
Factor (B)
Nozzle Bett 123V352
+40 0.11 0.73 76 Table 0.548 0 8318 63 22 0
17 34 137 270 Forging intermediate Shell 123V500
+40 0 09 0.70 58 Table (C) 3 01 1 292 74 94 0
17 34 149 270 Forging Lower Shell 122W195
+40 0 05 0.72 31 Table 2.52 1.248 38 69 0
17 34 113 270 Forcino 42.8 Surv. Data 53 41 85 17 110 270 (D)
Nozzle Belt to 21935
-56 0.18 0.70 170 Table (H) 0.548 0 8318 141.41 17 28 65 51 151 300 Intermed. Shell Circ Weld (100%)
Intermed. To Lower 72442 (SA-
-5 0 26 0 60 180 Table (Q 2.49 1.245 224.1 19.7 28 68 47 288 300 Shell Circ. Weld 1484)
(100%)
Footnotes:
(A)
Based on WCAP-12795, "Reador Cavity Neutron Measurement Program For Wisconsin Elecinc Power Company Point Beach Unit 2," Rev. 3, August 1995. As intermediate input to further calculations, these values are not rounded in accordance with ASTM E29.
(B)
The dimensionless fluence factor is calculated using the fluence factor formula from equation (2) of Regulatory Guide 1.99, Revision 2 ff = fM 28 - o te ias l, where f is the r
fluence in units of E19 n/cm'. For example, the 32 EFPY inner surface fluence factor for nozzle belt forging, heat no.123V352, ff = 0 548 "-aie s o ual = 0 8318 (C)
Non-credible surveillance data; see BAW-2325 for evaluation. Table CF conservative because difference between measured ART m and predicted ART or based on N
N Table CF is less than 2a (34*F).
(D)
Credible surveillance data; see BAW-2325 for evaluation.
(O Non-credible surveillance data, Table CF value based on best-estimate chemistry is higher than best fit calculated using surveillance data, and therefore conservative.
(F)
RTerscalculated per 10 CFR 50 61. RTers = RTwortui + ARTuor + Margin, where RTworcu, = unirradiated initial RTaor, ARTuor = Chemistry Factor x Fluence Factor, and Margin = 2(o * + o #)
- with o defined as the standard deviahon of the RTworcos, and oa defined as the standard deviation of ARTuor. For example, for nozzle bett forging.
u a
u heat no.123V352, RTers = 40 +(76 x 0.8318) + 34 = 137*F. Calculated RTers values are rounded to the nearest "F in accorda_nce wittt the rounding-off method of ASTM Practice E29.
(G) instruction Manual, Reactor Vessel, Point B mch Nuclear Plant No. 2, Combustion Engineenng, CE Book #4869, October 1970.
(H)
Table CF value based on best-eshmate cic ' rtnt data from CEOG Report "Best Estimate Copper and Nickel Values in CE Fabncated Reactor Vessel Welds." CE NPSD-1039, Revision 2, Final Report, June 1997,
PAGE 16 OF_2.jl CALCULATION SHEET CALC. NO.98-0132. Rrv. O
%S TITLE Evaluation of 23.6 EFPY P-T Limit and LTOP Apolicability Datq_.pn_d_
MADE BY T.D. Sprv DATE 9/15/98 Pressurized Thermal Shock at 32 EFPY REV'D. BY J.R. Pfefferle DATE 9/15/98 Calculetions (cont'd):
11.
Fluence Projection to January 1, 2001 (Calender Date Used as input to LTOP Setpoints in Calculation No. 98-0046, Revision 0)
A)
Maximum rated reactor thermal output:
1518.5 MWTh (Ref.12)
The total thermal oumut'for each unit as of March 1,1998, is (Ref.11):
Unit 1: Total thermal output = 273,540,535 MW hours - thermal Unit 2: Total thermal output = 266,886,275 MW hours - thermal Converting to EFPY:
Unit 1:
273,540,535 MWTh x eff. full nower x 1 day x 1 vear 20.5 EFPY
=
1518.5 MWTh 24 hrs.
365.25 days Unit 2:
20.0 EFPY 266,886,275 MWTh x eff. full oower x 1 day x 1 vear
=
1518.5 MWTh 24 hrs.
365.25 days B)
EFPY until 1/1/2001 Calender years from 3/1/1998 to 1/1/2001 is 2.83 years; an 80%
capacity factor is assumed (Assumption 1):
2.83 years x 0.80 = 2.3 EFPY C)
Total EFPY through 1/1/2001:
Unit 1:20.5 EFPY + 2.3 EFPY = 22.8 EFPY Unit 2:20.0 EFPY + 2.3 EFPY = 22.3 EFPY Results and
Conclusions:
This calculation demonstrates that the current Technical Specification P-T limits are conservative through the current licensing basis applicability date of 23.6 EFPY for Unit 1.
For Unit 2, as a resu!t of small, conservative changes from the inputs used in Calculation N-94-058 Revision 2, based on the latest understanding of initial RTuor and oi or limiting f
beltline weld SA-1484, the applicability date of the current licensing basis P-T limits is reduced slightly, to 23.4 EFPY. Administrative requirements to restrict operation of the l
PAGE 17 OF_jL5_
CALCULATION SHEET l*
l CALC. NO.98-0132. _Rlv 0 7tf l
TITLE Evaluation of 23.6 EFPY P-T Limit and LTOP Appli_cability Date and MADE BY T.D. Sorv DATE 9/15/98 Pressurized Thermal Shock atj)2 EFPY REV'D. BY J.R. Pfefferie_ DATE - 9/15/98 Unit 2 reactor vessel to no more than 23.4 EFPY using the current licensing basis P-T limits must be established. This calculation demonstrates that this value of EFPY will not be exceeded before the January 1, 2001 calender applicability date used in Calculation No. 98-0046, Revision 0, for LTOP pressure setpoints and enable temperature.
This calculation demonstrates that the PTS Screening Criteria are not exceeded through 32 EFPY.~
The LTOP. pressure setpoint and enable temperature established in Calculation No. 98-0046, Rev. O are unchanged through January 1, 2001, because that calculation used the same material property inputs and fluence estimation bases as this calculation.
I 1
i i
l l
l I
C.Mc
% -cist Lee c Page IS of 2i aus A t+4c k ud Tb5 F R AM ATO M E TECHHOLOGIEs August 26,1998 FTI-98-2563 Mr. J. R. Pfefferle Wisconsin Electric Power Company 231 W. Michigan Street P. O. Box 2046 Milwaukee, WI 53201
Subject:
Reevaluation ofWeld Wire Heat 61782
Attachment:
FTI Calculation 32-5002184-00
Dear Mr. Pfefferle:
The attached calculation summarizes the reevaluation of weld wire heat 61782 for Point Beach I for an inadiation temperature of 538 F.
If you should have any questions regarding this calculation summary, please feel free to call me at 804/832-3293 or Matt Devan at 804/832-3160.
Sincerely, D. L. Howell Project Manager B&W Owners Group Management DLH/mcl Attachment l
l 3315 Old Forest Road, P.O. Box 10935, Lynchburg, VA 24506-0935 Telephone: 804-832-3000 Fax: 804-832-3663 i
Internet: http /www.framatech.com
{}ggg 1
LtA Ic N b -0IE2 cbv v 1% y I'l et 2f Arr-tsc (me er
~Tb3 i /6h8 227 W :.%
CALCULATIONAL
SUMMARY
SHEET (CSS) mm
??M,AT,Q,lj r
DOCUMENT IDENTIFIER TITLE Chemistry Factor For Weld Wire Heat 61782 PREPARED BY:
REVIEWED BY:
NAME M.J.DeVan NAME J. B. Hall h
SIGNATURE
)
.t SIGNATURE 17
\\
TITLE Engineer IV DATE g/gj/4 g TITLE Engineer ill DATE gg,y COST CENTER 41020 REF. PAGE(S) 7 TM STATEMENT: REVIEWER INDEPENDENCE FURPOSE AND
SUMMARY
CF RESULTS:
j l
PURPOSE:
e Th3 purpose of this calculation is to provide an evaluation of the weld wire heat 61782 surveillance data for assessing
{
th2 integrity of the Point Beach Unit i reactor vessel.
SUMMARY
OF RESULTS 4
in accordance with the NRC guidelines provided at the November 12,1097 meeting, the surveillance data for weld wire h:at 61782 was evaluated with respect to the Point Beach Unit 1 reactor vessel. The R. E. Ginna plant-specific weld m:tal surveillance data was used in the assessment because it requires the least amount of adjustments to the mrersured data. The plant-specific weld metal surveillance data from R. FL Ginna was determined to be credible with r:spect 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 criculated to be 161.1*F.
THE FOLLOWING COMPUTER CODES HAVE BEEN UsED IN THIS DOCUMENT:
CODE / VERSION / REV CODE IVERSION / REV THIS DOCUMENT CONTAINS ASSUMPTIONS THAT MuST BE VERIFIED PRIOR TO USE ON SAFETY.RELATED WORK YES (
)
NO ( X}
PAGE 1
OF 7
sjc c g;-yragy r --
P9e 20 of 25
- FRASfA TOME Attackment Tb5et).i}1g TECHNOLOGIES FTI NON-PROPRIETARY 32-5002/3.t-00 RECORD OF REVISIONS REVISION DESCRIPTION l
00 Original Release l
l l
l 1
)
\\
i
)
i i
i PREPARER:
M. J. DeVan DATE: 08/20/98
^ REVIEWER:
1.B. Hall DATE: 08/20/98 Page 2
P9 2.icf 25 FRAMA TOhfE mka "TDS Qi5l48 TECHNOLOGIES FTI NON-PROPRIETARY 33.5002134-00 TABLE OF CONTENTS TITLE PAGE 1.0 In trod u c ti o n............................................................,........................
2.0 -
S umrnary o f Results............................................................................................. 4 1
3.0 Assumptions,......................................................................................................4
]
1 4.0 Evaluation and Use of Weld' Wire Heat 61782 Surveillance Data............................. 4 4.1. C re dib ili ty Assessment..................................................................................... 5
{
4.2. Determination of Chemistry Factor............................................................... 6 i
5.0 References..................................................................................................7 i
1 l
i 4
]
l PREPARER:
M. J. DeVan DATE: 08/20/98 REVIEWER:
- 1. B. Hall.
DATE: 08/20/98 Page 3
.~G_ic 98-clM. Rev.C pq, 11 4 2.9 FRAMA TOME m&w TDS 9phE
+
1 TECHNOLOGIES FTI NON-PROPR1ETARY 32-5002184-00 1.0 Introduction
\\
Both Regulatory Guide 1.99, Revision 2* and 10 CFR 50.61* state that surveillance data (if available) be censidered 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 against five criteria and the calculation of the chemistry factor based on the surveillance data. The NRC provided guidance on performing evaluation 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.*
This calculation provides the evaluation and use of weld wire heat 61782 surveillance data for assessing the integrity of the Point Beach Unit I reactor vessel. The guidelines, provided i
by the Staffin the November 12,1997 meeting, are used to determine the chemistry factor.
2.0 Summary of Results In accordance with the NRC guidelines provided at the November 12,1997 meeting, the stuveillance data for weld wire heat 61782 was evaluated with respect to the Point Beach Unit I reactor vessel. The R. E. Ginna plant-specific weld metal surveillance data was used in the assessment because it requires the least amount of adjustments to the measured data.
'Ite 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 161.laF.
3.0 Assumptions No major assumptions are contained in this report.
4.0 Evaluation and Use of Weld Wire Heat 61782 Surveillance Data The lower shell longitudinal weld, found in the Point Beach Unit I reactor vessel, was fabricated using the weld wire heat 61782. The best estimate copper and nickel chemical compositions for this wire heat are as follows:
Cu = 0.23 wt%
Ni = 0.52 wt%
Weld wire heat 61782 surveillance data are not available from the Point Beach Unit 1 plant-specific surveillance program, but are available from other sources. The available weld wire heat 61782 surveillance data are presented in Table 4-1.
PREPARER:
M. J. DeVan DATE: 08/20/98 REVIEWER:
- 1. B. Hall DATE: 08/20/98 Page 4
~
6tc af,cist h,,g c FR,0fA TOSfE Paq( 1.3 'S 2r TECHNOLOGIES FTI NON PROPRIETARY kqch w T S *l'h 500M t-00
' Table 4-1. Available Surveillance Data For Weld Wire Heat 61782 Inadiacon Measa-2 Capsule ID Cu Ni Temperaturt Fluence
,iRTm. ('F)
(including source) wt%
wt%
(*F)
(x10 n/cm )
(TANTO 3
B&WOG; Capsule DBI l.GI 0.27 0.59 556 1.03 141 sA.!!35: ONs-2 Nozzle Belt Dropout Matl.
R. E. Ginna: Capsule v 0.24 0.52 545 0 556 146 sA-1036: Plant specific RVSP Material R. E. Ginna Capsule R 0.24 0.52 545 1.15 167 sA-1036: Plant specific RVsP Material R. E. Ginna Capsule T 0.24 0.52 545 l.97 169 sA-l036: Plant specific RVsP Material R. E. Ginna: Capsule s 0.24 0.52 545 3.37 222 sA-1036: Plant specific RVsP Material The B&WOG Capsule DB1-LG1 was irradiated in the Davis-Besse reactor vessel which Babcock & Wilcox is the NSSS vendor. The weld metal surveillance data for R. E. Ginna was irradiated as part of their plant-specific surveillance program.
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 is lower for R.
E. Ginna than Davis-Besse. In addition, Point Beach Unit I and R. E. Ginna have the same NSSS vendor, Westinghouse. Therefore, only the R. E. Ginna weld metal 61782 surveillance data is used in assessing the integrity of the Point Beach Unit I vessel.
4.1. Credibility Assessment Since the R. E. Ginna weld metal 61782 surveillance data are from one (1) source, a best-fit line is determined relating the measured (unadjusted) ARTNor to the fluence factor (determined from the capsule fluence). Table 4-2 presents the credibility assessment for weld wire heat 61782 using only the R. E. Glnna plant-specific weld metal surveillance data.
Slope of the Best-Fit Line = 159.O *F i
l l
PREPARER:
M. J. DeVan DATE: 08/20/98 REVIEWER:
.J.B. Hall DATE: 08/20/98 Page 5
FRAMA TOME y 2 4 cf 2 f w,,,,,,,2-5002/ N-009l,gj9g 395 TECHNOLOGIES FTI NON-PROPRIETARY 3
Table 4-2 Credibility Assessment for Weld Wire Heat 61782 Meas.
Predicted (Measured -
Irred.
ARTer.
$Te from Predicted)
Capsule Cu Ni Chem.
Temp.
Fluence
(*F)
Best Fit Une
- STm Designation wt%
wt*'.
Factor
(*n Factor frANE
(*R fan R. E. Ginnt Capsule V 0.24 0.52 161.4 545 0.836 146 132.9 13 i Plant. specific RVsP Matenal R. E. Ginnt Capsule R 0.24 0.52 161.4 545 1.039 167 165.2
!8 Plant specific RVSP Matenal R. E. Ginnt Capsule T 0.24 0.52 161.4 545 1.185 169 188.4 19 4 Plant-specific RVsP Matenal R. E. Ginnt Capsule s 0.24 0.52 161.4 545 1.349 222 214.5 7.5 Plant soccific RVsP Matenal where: Predicted ARTa or = (Slopewpd
- Fluence Factor v
These data are credible since the scatter is less than 28aF for all surveillance capsule data points.
4.2. Determination of Chemistry Factor The surveillance data chemistry factor is determined from a best fit line through the surveillance data adjusted to account form 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., Point Beach Unit 1).
The operating cold leg temperature (TPiant) for the Point Beach Unit I reactor vessel is 538aF, and the R. E. Ginna surveillance capsules have a irradiation temperature (TCapsule) of 545aF that is greater than Triant. Herefore for the capsules with TCapsuie greater than 538aF (i.e.,
Triant), the ARTNor. measured must be adjusted by increasing the measured ARTsor by 1.0aF for each degree difference in irradiation temperature to yield the temperature adjusted ARTsor (i.e., ARTNDT. Temp. Adjusted).
To account for the differences 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 (i.e., Point Beach Unit 1). To obtain the " ratio and temperature" adjusted ARTsoT, the surveillance data are adjusted as follows:
' C F,ua y wrcac=. 3 7
Ratio lTemperatureAdjustedARTuor==
- 6RTso r.ra-o A=~4 (p
r 4
The assessment of the surveillance data for weld wire heat 61782 with respect to Point Beach Unit I using the R. E. Ginna plant-specific weld metal surveillance data is presented in Table 4-3.
PREPARER:
M.1. DeVan DATE: 08/20/98 REVIEWER:
- 1. B. Hall DATE: 08/20/98 Page 6
'~
^
C_a te. 96-ci n b, c Pay 2 5 c f z i FRAMA TOME mg a,a ns gj,,jy TECHNOLOGIES FTI NON-PROPRIETARY 32-5002/W-00 l
Table 4-3. Surveillance Data Assessment for Weld Wire Heat 61782 For Point Beach Unit 1 Temp..
Irrad.
Capsule Fluence Meas.
Temp.
Rauo Temp.
Fluence Factor
.iRTuor, (F)
Adjusted Adjusted Caosule Cu Ni CF m
(n/cmh (fn (TANE ART.,er. A aRTwor. m R. E. Ginnt Capsuie V 0.24 0.52 161.4 545 0.556 0836 146 153 149 2 j
Plant Specific RVsP Matt.
1 R. E. Ginnt Capsule R 0.24 0.52 161.4 545 1.15 1.039 167 174 169.7 l
Plant specific RVsP Matt.
l R. E. Ginnt Capsuic T 0.24 0.52 161.4 545 1.97 1.185 169 176 171.6 l
Plant-Specific RVsP Mad.
l R. E. Ginnt Capsule s 0.24 0.52 16l.4 545 3 87 1.349 222 229 l
223.3 Plant-Specific RVsP Matt.
I Point Beach Unit 1:
0.23 0.52 157.4 538 Vessel Average The best-fit line is determined relating the " ratio and temperature" adjusted aRTsoT to the fluence factor (determined from the capsule fluence). The slope of this best-fit line is the chemistry factor calculated from surveillance data, CFsurv. oata.
CFsan. cata = 161.10F
5.0 REFERENCES
- 1. U. S. Nuclear Regulatory Commission, " Radiation damage to Reactor Vessel Material, "
Regulatorv Guide 1.99 Revision 2. May 1988.
- 2. Code of Federal Regulations, Title 10, " Domestic Licensing ofproduction and Utili:ation Facilities," Part 50.61, " Fracture Toughness Requirementsfor Protection Against Pressurised Thermal Shock, " Federal Register, December 19,1995.
- 3. Memorandum from Keith R. Wichman to Edmund J. Sullivan, " Meeting Summaryfor November 12,1997, Meeting with Owners Group Representatives and NE1 Regarding Review ofResponses to Generic Letter 92-01, Revision 1, Supplement 1 Responses. "
dated November 19,1997.
PREPARER:
M. J. DeVan DATE: 08/20/98 REVIEWER:
- 1. B. Hall DATE: 08/20/98 Page 7