ML072410389
| ML072410389 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 08/22/2007 |
| From: | Laughlin G Constellation Energy Group |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML072410389 (23) | |
Text
Constellation Energy Nine Mile Point Nuclear Station P.O. Box 63 Lycoming, NY 13093 August 22, 2007 U. S. Nuclear Regulatory Commission Washington, DC 20555-0001 ATTENTION:
SUBJECT:
Document Control Desk Nine Mile Point Nuclear Station Unit No. 1; Docket No. 50-220 Submittal of Additional Engineering Evaluations for Two Reactor Pressure Vessel Weld Flaws in Accordance with Amended License Renewal Application Commitment (a) Letter from R. B. Abbott (NMPC) to Document Control Desk (NRC), dated September 14, 1999, Submittal of 1999 Inservice Inspection Summary Report and Flaw Indication Evaluations
REFERENCES:
(b) Letter from P. S. Tam (NRC) to J. H. Mueller (NMPC), dated May 5, 2000, Nine Mile Point Nuclear Station, Unit No. I - Evaluation of Flaw Indications in Reactor Pressure Vessel Welds (TAC No. MA65 10)
(c) Letter from J. A. Spina (NMPNS) to Document Control Desk (NRC), dated December 5, 2005, License Renewal Application (LRA) - Responses to NRC Requests for Additional Information Regarding LRA Parts 1, 2, 3 and 4 (TAC Nos. MC3272 and MC3273)
(d) NUREG-1900, Safety Evaluation Report Related to the License Renewal of Nine Mile Point Nuclear Station, Units 1 and 2, Volume 2, September 2006 By letter dated September 14, 1999 (Reference a), Niagara Mohawk Power Corporation (the previous licensee) submitted to the NRC for review and approval a structural evaluation of subsurface flaw indications found in two Nine Mile Point Unit 1 (NMP1) reactor pressure vessel (RPV) welds (RV-WD-140 and RV-WD-099) during refueling outage 15. The evaluations considered fatigue crack growth and irradiation embrittlement for up to 28 effective full power years (EFPY) of operation (i.e., the end of the original license term). The NRC staff concurred that continued operation with these flaws was acceptable until the end of the 28 EFPY in a safety evaluation dated May 5, 2000 (Reference b).
As discussed in the amended NMP1 License Renewal Application (LRA), Section 4.7.4, the analyses performed for these RPV weld flaws were considered to be time-limited aging analyses since the 4107
Document Control Desk August 22, 2007 Page 2 acceptability criteria were applicable only through the original 40-year license term. In a letter to the NRC dated December 5, 2005 (Reference c), Nine Mile Point Nuclear Station, LLC (NMPNS) made the following commitment regarding the RPV weld flaw evaluations:
"The RPV weld flaw evaluations will be revised to consider additional fatigue crack growth and the effects of additional irradiation embrittlement (for beltline materials) associated with operation for an additional 20 years (i.e., out to at least 46 EFPY) and submitted for NRC review and approval no later than 2 years prior to the period of extended operation. If the revised calculation shows the identified flaws cannot meet the applicable acceptance criteria, the indications will be reexamined in accordance with ASME Section XI requirements."
The above commitment was affirmed in the NRC's safety evaluation report related to license renewal of NMPI, as documented in Section 4.7.4 of NUREG-1900, Volume 2 (Reference d).
In accordance with the commitment stated above, this letter is submitting additional engineering evaluations that have been performed for the subject RPV weld flaws to project the evaluations to the end of the period of extended operation, by considering additional fatigue crack growth and the effects of additional irradiation embrittlement (for beltline materials) associated with operation for an additional 20 years (see Attachment 1). The evaluations utilized inputs from scoping pressure-temperature (P-T) curves that were based on projected fluence levels corresponding to 46 EFPY, using the current NMPNS Regulatory Guide 1.190 methods approved by the NRC. In addition, the evaluations used K1c (instead of Kla) for allowable fracture toughness consistent with the current NRC-approved NMPI P-T curve application of Code Case N-640 and IWB-3600 of the 2006 Edition of the ASME Code,Section XI. The additional engineering evaluations described in Attachment (1) are conservative and provide reasonable assurance that the flaws will remain acceptable and that structural integrity of the RPV will be maintained during the period of extended operation.
This letter contains no new regulatory commitments. Should you have any questions regarding the information in this submittal, please contact T. F. Syrell, Licensing Director, at (315) 349-5219.
Very truly yours, Ga Jay Laughlin Manager Engineering Services GJL/DEV
Attachment:
(1) Nine Mile Point Unit I - Additional Engineering Evaluations for Reactor Pressure Vessel Welds RV-WD-140 and RV-WD-099 for the License Renewal Period (Calculation SOVESSELM030, Revision 01, Disposition 01 B) cc:
S. J. Collins, NRC M. J. David, NRC Resident Inspector, NRC
ATTACHMENT (1)
NINE MILE POINT UNIT 1 ADDITIONAL ENGINEERING EVALUATIONS FOR REACTOR PRESSURE VESSEL WELDS RV-WD-140 AND RV-WD-099 FOR THE LICENSE RENEWAL PERIOD (Calculation SOVESSELM030, Revision 01, Disposition 01B)
Nine Mile Point Nuclear Station, LLC August 22, 2007
P~age I (Next) "t.
Engineering Lotae 1 NZ Services DISPOSITION COVER SHEET Last 7_7 Project: NINE MILE POINT NUCLEAR STATION Unit (1,2 or O=Both):
1 Discipline: Mechanical Title Calculation No.
Rev:
Disp SOVESSELM030 101 01B Reconciliation of Previous RPV Flaw Evaluations for the Ucense Renewal Period Originator
/.{
R. Corieri Date X,
(Sub)System(s)
RXVE Index No.
Reviewer G. Inch Date
//
ISO l;/-
7 DER. Evaluation or Change No.
Approver Bart)lini D7 NCTS # 504582.19 9__/_/_,
Safety Class: (SR*/NSRJQXX):
SR NMP Acceptance/Date
,) A
' II SR, attach or reference the associated Design Verification Report. (The attached calculation was performed by SIA under Safely Related PO 9 05-55640-002 and therefore was design verified by SIA under their SR QA Program).
Superseded Document(s): NIA Output provided? N If yes, group(s):
Y/N Description of Change SEE PAGE 2 Resolution SEE PAGE 2 Cross Reference Change(s):
- 1.
ATI-05-034-001
- 2.
MPM-405778
- 3.
NER-1 M-063 General Reference(s): NCTS # 504582-19 Confirmaton Required (Yes/No):...
N Final Issue Status Turnover See Page(s):
I APP Req'd (Yes/N/A): N/A 10 CFR50.59 Evaluation Number(s):
Compnent ID(s)(As shown in MEL):
Copy of Applicability Determination or 50.59 Screen Attached? Yes 0 No't N/A o
- 1 'No", location of AD/Screen?
RPV-NR02 Key Words: License Renewal, ASME, Reactor Vessel
ENGINEERING SERVICES CALCULATION CONTINUATION SHEET Page 2
(Next
.L Project: Nine Mile Point Nuclear Station Unit: 1 Disposition: 01B Originator/Date Reviewer/Date Calculation No.
Revision R. Corieril6-18-07 G. Inch/6-18-07 SOVESSELM030 01 Ref.
Table of Contents Description of Change & Resolution...........................................
Pages 2-3 SIA Calculation Package NMP-05Q-303......................................
Pages 4-16 SIA Letter GLS-07-022 - Flaw Proximity Assessment..................... Pages 17-20 50.59 S creen...........................................................................
P ages 21-22 Description of Change
Background:
In 1999, unacceptable indications in the RPV shell per ASME XI IWB-3500 were identified by Ultrasonic exams (UT) in axial weld RVWD-140 and shell-to-flange circumferential weld RVWD-099. The detected flaws are subsurface planar flaws located parallel to the centerline of the weld (i.e., the indications in RVWD-140 were axially-oriented and the indications in RVWD-099 were circumferentially-oriented). The flaw evaluations considered fatigue crack growth and irradiation embrittlement (only applicable for the beltline weld, RVWD-140) to 28 Effective Full Power Years (EFPY). The original flaw evaluations for these flaws were performed in revision 1 of this calculation SOVESSELM030 using a flaw handt ook developed by General Electric Nuclear Energy (GENE). The original flaw evaluations were submitted to the NRC for review and approval under NMPC letter dated September 14, 1999. The NRC reviewed the original evaluations and concurred that continued operation with these flaws is acceptable through 28 EFPY, the end of the current license term, as stated in the NRC SE dated May 5, 2000.
The original flaw evaluations determined that leak test and bolt-up conditions were the most limiting conditions for fracture analysis of the flaws. The leak test (i.e., ASME Xl Leakage Test) was identified as the limiting loading condition in the axial weld RVWD-140 and reactor vessel bolt-up was limiting for the shell-to-flange circ weld RVWD-099. In 2003, the NRC issued Technical Specification Amendment No.
183 which revised the NMP-1 reactor coolant system pressure-temperature limit curves and tabies in Tech Spec Section 3.2.2/4.2.2, "Minimum Reactor Vessel Temperature of Pressurization". The revised P-T curves were developed using Code Case N-640 "Alternative Reference Fracture Toughness for P-T Curves". Use of code case N-640 ultimately decreased the leak test temperatures, which decreases the fracture toughness. As such, the existing flaw evaluations were dispositioned in SOVESSELMO30-*01A (SIA calculation NMP-05Q-303 Revision 0) to reconcile the leak test conditions associated with -'he updated P-T limit curves. The calculation disposition concluded that the previously detected flaws remained acceptable when compared to the updated (lower) allowable flaw sizes at 28 EFPY. The minimum temperature for bolt-up (100 IF) remained unchanged by the Tech Spec amendment; however, the calculation disposition conservatively evaluated the flaws assuming a lower bolt-up temperature of 70 OF. The revised flaw evaluations were not submitted to the NRC for review and approval.
In 2005/2006 during the License Renewal application period, NMP1 committed to submit revised flaw evaluations for the subject flaws in the RPV shell welds to the NRC for staff review and approva. at least two years prior to entering the period of extended operation. The revised flaw evaluations were 1o HARPV\\RPV WELDS\\SOVESSELM030-01 B Continuation Sheets.doc NEP.DES-08 Rev 07
ENGINEERING SERVICES CALCULATION CONTINUATION SHEET Pag,__ 3 (Nex4 )
Project: Nine Mile Point Nuclear Station Unit: 1 Disposition: 01B Originator/Date Reviewer/Date Calculation No.
Revision R. Coderil6-18-07 G. Inch/6-18-07 SOVESSELM030 01 Ref.
consider additional fatigue crack growth and the effects of additional irradiation embrittlement ("or beltline materials) associated with operation for an additional 20 years (i.e., out to at least 46 EFPY).
This commitment is documented in NUREG-1900, Volume 2, Section 4.7.4.2. and in the Unit 1 UFSAR Appendix C, Section C.2.5. 1. In December 2005, the attached revision 1 of SIA calculation NMP-05Q-303 revised the allowable flaw sizes for fatigue and irradiation embrittlement for 46 EFPY. Since the limiting loading condition for flaws in axial weld RVWD-140 is the leak test condition, the P-T curve for leakage testing had to first be revised for projected fluence levels corresponding to 46 EFPY. The Reference 3 calculation documents "scoping" P-T limit curves for 46 EFPY. The scoping 46 EFPY P-T limit curves were developed using the same methodology as the existing P-T curves approved by the NRC in Tech Spec amendment 183. The scoping P-T curves were based on draft best estimate neutron fluence calculations available in 2005 when the attached calculation was originated. The draft neutron fluence calculations were performed in accordance with Regulatory Guide 1.190. Subsequently the neutron fluence calculations were finalized in MPM-405778, "Neutron Transport Analysis for NMP-I".
The 46 EFPY fluence exposures at weld RVWD-140 used in the attached SIA calculation was compared to the final fluence calculated in MPM-405778. This comparison determined that the fluence used in the attached calculation is conservatively bounded by the final fluence calculation. Since both the scoping P-T curve calculations and the attached SIA calculation used the higher draft fluen.,e values, the calculated allowable flaw sizes at weld RVWD-140 are deemed to be conservative. The allowable flaw sizes for shell-to-flange weld RVWD-099 were also recalculated assuming 20 additional years of fatigue crack growth. The revised allowable flaw sizes were conservatively determinec for a bolt-up temperature of 700F, although the current minimum Tech Spec bolt-up temperature is 100!"F.
Resolution & Conclusions The attached calculations provide "scoping" allowable flaw sizes for welds RVWD-140 and RVWD-099 out to 46 EFPY in accordance with NMPI's License Renewal commitment. The calculation concludes that existing flaws in the two welds are acceptable as compared to the 46 EFPY acceptance criteria.
The calculations while considered for information only are conservative and are only intended to provide reasonable assurance that the flaws will remain acceptable and the structural integrity of the RPV will be maintained during the period of extended operation. Final allowable flaw sizes will be re-calculated when P-T curves for the license renewal period are developed. The impact of the future P-T cur've development on the attached calculation will be re-visited at some future time under the NMP cesign change control process. The NMP1 fluence methods are based on approved Regulatory Guide 1.190 methods which require maintenance of fluence projections based on routine updates using actual core operating conditions and changes to ART as needed. In addition NMP1 is part of the BWRVIP ISP Program which requires review of the ART when ISP capsules are removed. Therefore the acceptability of the flaws is reviewed whenever the ART and/or P-T curves require adjustment.
The NRC commitment as documented in NUREG-1900 and the UFSAR also states that the flaws will be reexamined in accordance with ASME Section XI as necessary. Because the revised flaw evaluations contained herein demonstrate that the flaws are acceptable for additional 20-years, the current ASME XI inspection frequency of once/interval for examination category B-A pressure retaining welds is adequate.
H:\\RPV\\RPV WELDS\\SOVESSELM030-01 B Continuation Sheets.doc NEP-DES-08 Rev 07
p Z
CALCULATION FILE No.: NMP-05Q-303 STRUCTURAL INTEGRITY Associates PACKAGE PROJECT No.: NMP-09Q PROJECT NAME: Nine Mile Point FatiguePro CLIENT:
Constellation Energy Group (Nine Mile Point Unit 1)
CONTRACT NO.: 05-55640-002 CALCULATION TITLE: RPV Flaw Evaluation PROBLEM STATEMENT OR OBJECTIVE OF THE CALCULATION:
This calculation provides a reconciliation of the previous flaw evaluation performed for the NMP-1 RPV. The allowable flaw sizes computed by GE for 28 EFPY for the indications in question are first reproduced. This step ensures consistency in methodology application. Then, revised allowable flaw sizes are computed for 46 EFPY (projected end-of life value for 60 years of operation) using the appropriate revised pressure test temperature for comparison to the previously as-found indications.
Project Mgr.
Preparer(s) &
Document Affected R
Approval Checker(s)
Revision Pages Revision Description Signature &
Signatures &
Date Date 0
1-8 Original Issue G. L. Stevens G. L. Stevens 11/5/02 GLS 11/5/02 On Computer K. K. Fujikawa Files KKF 11/5/02 I - 13 Revised to evaluate license G. L. Stevens G. L. Stevens renewal operation for 60 years.
GLS 12/27/05 In computerfiles Revisions are marked brhyh..,!.,,w Irevision bars" in the right hand aiX?))0,5 files margin.
K. K. Fujikawa KKF 12/27/05 Page I of 13
~CWb*AL~AiCZ~- r~R Table of Contents 1.0 IN T R O D U C T IO N.......................................................................................................................
3 2,0 IN P U T S.......................................................................................................................................
- 3 3.0 BENCHMARK ANALYSIS......................................................................................................
5 4.0 R E V ISED A N A L Y SIS................................................................................................................
7 5.0 REVISED ANALYSIS #2....................................................................................................
10 6.0 C O N C LU SIO N S.......................................................................................................................
12 7.0 R E FE R EN C E S..........................................................
....................................................... 1:3 List of Tables Table 1. Flaw s To B e E valuated.........................................................................................................
3 Table 2. Reproduction of Original ART Calculations for 28 EFPY........................................................
5 Table 3. Reproduction of Revised ART Calculations for 46 EFPY........................................................
3 Table 4. Revised ART Calculations for 46 EFPY Using Reduced Fluence....................................
10 List of Figures Figure 1. Benchmark Results for Vessel Flange Horizontal Weld (Figure D-3 of [2])................ 5 Figure 2. Benchmark Results for Lower-Intermediate Course Weld (Vigure D-12 of [2])...............
5 Figure 3. Results of Revised Analysis #2 for Lower-Intermediate Course Weld..................................
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NMP-05Q-303 Page 2 of 13
1.0 INTRODUCTION
During past RPV weld examinations for Nine Mile Point Unit I (NMP-1), flaws were detected in the reactor pressure vessel (RPV) that required IWB-3600 evaluation [1]. The flaws were dispositioned via a RPV Flaw Handbook prepared for NMP-I by GE [2]. The flaw handbook determined the boltup and pressure test conditions to be limiting, so allowable flaw sizes were determined based on the pressure-temperature (P-T) curve values for pressure test conditions for 20.3 and 28 EFPY. Revision 0 of this calculation was performed to reconcile the prior GE flaw evaluation due to the revision of thie P-T curves to incorporate Code Case N-640 (i.e., application of K1.). The P-T curves were recently revised for license renewal operation [3]. This caused a change in the P-T curves, thereby changing the required pressure test temperature and, therefore, the resulting allowable fracture toughness. As a result, reconciliation of the prior RPV flaw evaluation was considered necessary.
In this calculation, the previous flaw evaluation for the RPV is reconciled. The allowable flaw sizes originally computed by GE for 28 EFPY for the indications in question are first reproduced. This step ensures consistency in methodology application. Then, revised allowable flaw sizes are computed for 46 EFPY (projected end-of life value for 60 years of operation) [3] using the appropriate revised pressure test temperature for comparison to the previously as-found indications.
This calculation details all inputs, methodology, and analysis results associated with the RPV flaw reconciliation analysis calculation.
2.0 INPUTS Reference [1] provides the flaws to be evaluated, as shown in Table 1.
Table 1. Flaws To Be Evaluated Flaw Flaw Handbook Flaw 112 Flaw Flaw GE Wall RPV Weld ID Flaw Id Orientation Figure No.
Depth, 2a Depth, a Length, L a/L Allowable Thick, t alt RVWD-099 109/139 Circ D-3 0.396 0.198 6.75 0.0293 1.20 7.2 0.0275 RVWD-099 1-112 Circ 0-3 0.594 0.297 1.25 0.2376 1.55 7.2 0.0413 RVWD-099 1.113 Circ D-3 0.594 0.297 3.25 0.0914 1.26 7.2 0,0413 RVWD-099 1-114 Circ 0-3 0.594 0.297 3.5 0.0849 1.24 7.2 0.0413 RVVVD-099 1-115 Circ 0-3 0.552 0.276 3.5 0.0789 1.23 7.2 0.0383 RVWD.099 1-116 Clrc 0-3 0.,552 0.276 2.5 0.11,04 1.31 7.2 0.0383 RVWD-099 221149 Clrc D-3 0.453 0.2265 7.75 0,0292 1.20 7.2 0.0315 RVWD-140 55 Axial D-12 0.396 0.198 13.75 0.0144 0.90 7.2 0.0275 RVWD-140 9-015÷016 Axial 0-12 0.424 0.212 3.0 0.0707 1.00 7.2 0.0294 The remaining inputs were obtained from Reference [2], as follows:
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NMP-05Q-303 Page 3 of 13
For GE Figure D-3 Flaws:
Base Metal Thickness = 7.125"
[2, Table A-I]
Clad Thickness = 0.2188"
[2, Table A-I]
RTNiyr = 407F
[2, Table A-2, All Vertical Welds in Upper Shell Course]
Adjusted Reference Temperature, ART = 40°F (Flange region not affected by fluence)
Condition Evaluated = Boltup
[2, Table A-4c]
Condition Temperature= 100'F
[2, Table A-4c]
Yield Stress = 50 ksi
[2, Table A-4c]
Flaw Orientation = Circumferential (see Table I above)
Stresses:
[2, Table A-5, Non-Beltline (near flange) Boltup]
Pressure Stress = 0.0 ksi a,, = 0.0 ksi Gb = 26.0 ksi Weld Residual Stress = 0 ksi Clad Residual Stress = 35.0 ksi Fatigue Crack Growth Cycles =18 cycles/EFPY [2, p. 10] = 18 9.7 = 175 cycles For GE Figure D-1 2 Flaws:
Base Metal Thickness = 7.125"
[2, Table A-I]
Clad Thickness = 0.2188"
[2, Table A-1]
RTNrr = 407F
[2, Table A-2, Weld @ 225']
ART = 1220F @ 1/4t
[2, Table A-3b, Weld @ 225°]
(The above ART value is reproduced in accordance with Reg. Guide 1.99, Rev. 2 [4] in Table 2 below.)
Condition Evaluated = Pressure Test (2, Table A-4e]
Condition Temperature = 260°F
[2, Table A-4c]
Yield Stress = 46.01 ksi
[2, Table A-4c]
Flaw Orientation = Axial (see Table I above)
Stresses:
[2, Table A-5, Vertical Welds Beltline]
Pressure Stress = 18.51 ksi am = 0.3 ksi rb =0.5 ksi Weld Residual Stress 8 ksi (bending)
Clad Residual Stress 17.11 ksi Fatigue Crack Growth C3cles = 18 cycles/EFPY [2, p. 101 18
- 9.7 = 175 cycles Flaw eccentricity ratio, e/t:
Flaw 55: (7.98"/2 - 2" - 0.396"/2) / 7.98" = 0.22
[1, pg. A6]
Flaw 9-015+016 = (8.00"/2 - 2.20" - 0.424"/2)/ 8.00" = 0.20
[1, pg. B6]
The new pressure test temperature from Figure 6 of Reference [3] is 1941F at the limiting 1/4t location for a leak'test pressure of 1,050 psig (per Constellation input). Note that any potential future increase:;
in the leak test pressure are bounded by this evaluation since a higher leak test pressure will yield a Revision 0
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r,.
higher temperature (and the lower temperature used in this evaluation is bounding because of the associated lower fracture toughness),
'Fable 2. Reproduction of Original ART Calculations for 28 EFPY NMP-1 RPV ARTNDT Calculation for 28 EFPY rNOTE: rhis catcuJarioa dupsieates the previous calculation, rro.m Reference (2) for 28 CFPY. and Is used for the Benchmar, Analnsa Only.)
Chemistry Chamlsary Adjulstents Fo: 11/t Initial RT. 0, Factor
&RT,'r Margin Term!
ARTor Location
(*F)
Cu(We%)
N1lwtv)
(IF) rF)
- a. *F) qg (IF) EFPY (IF)
Weld Q 225" - Weld Where Flaws Arte Localed 40 0.27 0.53 173.85 47.6 17.0 0.0 28.0 121.6 luse 11lhmtno Plate G-307-4?
clýme ry) 7 Flange Horlzoneto Weld 40 0.0 0.0 o 00 40.0 Fluencs Information:
WOt Thtdnrss t (inches)
Fluence at I0 Attonuation. 1/41 Flueanc @ 114t FRunce Factor, FF Locatilon Full 1141 EFPY mknh "42*
in/iO fM i/.,0c WNld 0 225" 7.344 2.000 28.0 7.16E+17 0.810 4.43E+17 0.27 Flanne Horizontal Woel 7.344 2.000 280 1,00E-00 11.619 6.19E-01 f,0 9 3.0 BENCHMARK ANALYSIS As a first step, the allowable flaw sizes originally developed in Reference [2] were reproduced to substantiate the methodology used. This was accomplished using the SI Program APPENDA [51, which is an in-house, verified computer program for performing flaw tolerance analysis of reactor vessel shells. APPENDA uses the same methodology outlined in the Reference [21 report. The. inputus described above were input to APPENDA for each of the flaws.
The values of the flaw eccentricity ratio, e/t, that were used to develop Figures D-3 and D-12 of Reference [2] were not documented in Reference [2]. Therefore, a range of e/t values was evaluated with the APPENDA program until the previous results were identically matched. Values of e/t of
-0.17 and -0.38 were determined for Figures D-3 and D-12, respectively.
The results are shown in Figure 1 (corresponding to Figure D-3 of Reference [2]) and Figure 2 (corresponding to Figure D-12 of Reference [2]). The flaws identified in Table I are also included in Figures 1 and 2.
The APPENDA input files for these two cases are D3C.IN and D 12A.IN, respectively, and are included in the computer files associated with this calculation. The results are documented in output summary files D3C.SUM and DI2A.SUM, which were incorporated into Excel spreadsheets "Allowable Flaw Sizes (D3).xls" and "Allowable Flaw Sizes (D12).xls". All of these files are also included in the computer files associated with this calculation.
f:
- Zt
?
Figure 1. Benchmark Results for Vessel Flange Horizontal Weld (Figure D-3 of 121)
NMP-1 RPV Flaw Evaluation (Non-Bemflne. Vessel Flange Hortzontal Weld Subsurface Flawl 2¢.5
.2 0
.1
.*.2 0.5 Flow A.P00t Ratio lAL)
Figure 2. Benchmark Results for Lower-Intermediate Course Weld (Figure D-12 of [21)
N1MP.1 RPV Flaw Evaluation (Lower4ntermedlate Course at 225% Subsurface Flaw) 12 1.0 0.4
-1W03600 Aftwabic Flow fta*
11 GE Flaw. Hanabeak Reuail Xt Ahowbtie lnrw eli. for T *1941 (oal - -0.3a)
-A~oawbo F~w~lt lT
- 194F (on.o.022)
. 4"bal~tSOcTIA U0U 000 0.Cs 0.10 0.10 020 0.2, Fla. A.peat Ratio I'A.)
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4.0 REVISED ANALYSIS Revised allowable flaw sizes were determined for a boltup temperature of 70'F (for information only),
and the revised pressure test temperature of 194°F. The following revised inputs apply:
For GE Figure D-3 Flaws:
Condition Temperature = 70'F Fatigue Crack Growth Cycles = 18 cycles/EFPY x (46 - 20.3*) EFPY = 463 cycles
(* EFPY level at the time of the Reference 12] analysis.)
All other inputs remain the same.
The revised allowable flaw sizes are shown in Figure 1 as "IWB-3600 Allowable Flaw Sizes for 70OF Boltup". The APPENDA input file for this case is D3D.IN, and is included in the computer files associated with this calculation. The results are documented in output summary files D3D.SUM, which was incorporated into Excel spreadsheet "Allowable Flaw Sizes (I)3).xls". All of these files art; also included in the computer files associated with this calculation.
For GE Figure D-12,F1aws:
Condition Temperature = 1941F Fatigue Crack Growth Cycles= 18 cycles/EFPY x (46 - 20.3*) EFPY = 463 cycles
(* EFPY level at the time of the Reference (21 analysis,)
Stresses: All remain the same except the yield and clad residual stresses:
Yield Stress = 47.66 ksi ner calculation below:
Clad Residual Yield Stress:
T =
194 "F
(new pressure test temp.)
YS at 100"F =
50.00 inhn-'F (Table A-4c of Reference [2J)
YS at 260°F =
46.01 in/in--F (Table A-4c of Reference [2])
YS at T -
47.66 "F
(interpolated)
Stress = 23.00 ksi, per calculation below:
Clad Residual Stress:
T 194
- F (new pressure test temp.)
a, at 70°F 35.0 ksi (pg. A-4 of Reference [21)
Ess @ 70*F = 28,300 ksi (Reference (61)
Ess @ 200*F = 27,600 ksi (Reference [6])
Ess @ T 27,632 ksl (interpolated)
Aet for AT = 1080°F = 2.70E-06 Inlin-*F (pg. A-4 of Reference (21)
Aa for AT = -177"F = 2.44E-06 lnlin-'F (pg. A-5 of Reference (2])
AT = 70 - T =
-124
'F Ae for AT - 2.45E-06 In/in.'F (interpolated) ac at T =
23.00 ksi NOTB: The cladding stress has no effect on the subsurface flaws evaluated in this caiculation, out It Is incluudu for comp cWiness.
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ART = 167.9 0F @ 1/4t
[3, Table 3, Plate G-307-4]
(NOTE: The above ART value is reproduced in accordance with Reg. Guide 1.99, Rev. 2 [4) in Table 3.)
All other inputs remain the same.
Table 3. Reproduction of Revised ART Calculations for 46 EFPY NMP-1 RPV ARTNDr Calculation for 46 EFPY (NOre: This catulat/oUa dupltc, a(
he Cate uon from Referenco 131 for 46 EFPY. andls used for the Revised Analys is.
Chemistry Chemistry Adiustments For 114t Inlltal RT,,r Factor M1tT,0 MargIn Terms ART,,
Location (Ff)
Cu (wt %)
INI (wt%)
(F)
(*F) o.=(4F) :a, i(*F)
(-F)
Weld @ 226", Weld Where Flaws Ae Located 40 0,27 0.53 173.85 93.9 17.0
' 0.0 40.0 157.9 lug*li I~mlng I=tO.3"-
.hml*YY Flange Horizonujl W'teld 40 0.0 0.0 0.0 44.0 40.0 Wall Thlctneee. t (Inchoe Fluonce at I AllerntJfcon, 1141 Fluence 0 1141 Fluence Facto. FF Location FUl:
1141 EFPY Olecm-.')nc flOnl Weld 0 225' 7.125 1.761 45.0 2.7iE÷18 0.652 1.77! *16 0.54 Flange Horzonltl Weld 7.125 1.781 46.0 1.00*.00 0.652 6.52F 01 0.00 The revised allowable flaw sizes are shown in Figure 2 as "Allowable Flaw Sizes for T = 194°F (e/t
-0.38)". Two other cases were run for the actual flaw eccentricity ratios, e/t = -0.22 and -0.20, which are also shown in Figure 2 as "Allowable Flaw Sizes for T = 194 0F (e/t = -0.22)" and "Allowable Flaw Sizes for T = 194'F (c/t = -0.20)", respectively. The APPENDA input file for this case is D12C.IN, and is included in the computer files associated with this calculation. The results are documented in output summary files D I2C.SUM, which was incorporated into Excel spreadsheet "Allowable Flaw Sizes (D12).xls". All of these files are also included in the computer files associated with this calculation.
Based on the results shown in Figures 1 and 2, the following conclusions can be made with respect to the revised RPV flaw evaluation:
The allowable subsurface flaw sizes for the Non-Beltline, Vessel Flange Horizontal Weld Region at 46 EFPY are reduced for a boltup temperature of 70*F. This is a result of the lower allowable stress intensity factor, K1,, at 70°F versus 100°F. For the limiting flaw eccentricity ratio of -0.17, which is the basis for the original flaw diagram in Reference [2], the as-found indications are acceptable compared to these lower allowable flaw sizes.
/
The allowable subsurface flaw sizes for the Lower-Intermediate Course at 2250 Region at 46 EFPY are reduced for the revised pressure test temperature of 194'F. This is a result of the lower allowable stress intensity factor, K16, at 194°F versus the temperature of 2607F used in the Benchmark Analysis, and also because the fluence is higher for 46 EFPY compared to the fluence for 28 EFPY used in the Benchmark Analysis. For the limiting flaw eccentricity ratio Revision 0
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of -0.38, which is the basis for the original flaw diagram in Reference [2], the as-found indications are unacceptable compared to these lower allowable flaw sizes.
v" For the Lower-Intermediate Course at 2250 Region at 46 EFPY, and using the actual flaw eccentricity ratios of -0.22 (Flaw 55) and -0.20 (Flaw 9-015+016), the as-found indications are also unacceptable compared to the lower allowable flaw sizes for the revised pressure test In order to show acceptability of the flaws in Figure 2, additional analysis using the following items (some of which are identified in Section B-7 of Reference [2]) will be performed:
o Use Kl0 for the allowable fracture toughness. A Code change to KI0 has just passed the ASME Board for implementation in IWB-3600 of ASME Code Section XI. This change should be published in the 2006 Addenda of the Code.
o A revised ART value specific to the weld location will be used. Per Constellation input, the peak fluence for the upper plate from Reference [3], as used in Table 3 and this revised analysis, is 2.33 times higher than the peak fluence for the Weld @ 225'.
Evaluation considering the above two items is performed in the next section.
.1L 5.0 REVISED ANALYSIS #2 For the D-12 flaws, revised allowable flaw sizes were determined for the revised pressure test temperature of 194°F using K1, and a revised ART value specific to the weld location being evaluated.
The following revised inputs apply:
For GE Figure D-12 Flaws:
All inputs are the same as identified in Section 4.0, except:
ART= 137.3 0F @ 1/4t (see Table 4)
(NOTE: Thc above ART value is reproduced in accordance with Reg. Guide 1.99, Rev. 2 [4] in Table 3.)
Use Kic instead Kia for the allowable fracture toughness.
.VOTE: A project-specific revised version of the APPENDA program, called APPENDA2, was used for the K1, change. The only technical change made for the software was that the data statement defining K&,, was replaced with the data statement for KI,. Thus, any program outputs still identifies that KI, is being used, but in reality the values used are Kl. Verification of this can be found by viewing the output in the *.OUTfile.
Table 4. Revised ART Calculations for 46 EFPY Using Reduced Fluence NMP-1 RPVARTNDT Calculation for 46 EFPY Using Lower Fluence I flOflTF This en rcalsu/on uSBS ucedtlefluence oi1h Wnltd 226, using a tacrorJ of L3 forg YFPV rd Js ued for oRvls*dAniys 02) )
chemistry C~herletrV Adjuste~nwts For 114t initlal RT..,
Factor ART,4 0,
Margin ;!n~
Location (T
Cu (wt AI I N1 (wV%) __!_
P (T)
- o.,F)
IbF EFPY
(-F)
W4c)d a@225' - Weld Where Flaws Are Located 40 0.27 0.53 173.85 6313 17.0 0.0 48.0 137.3 (use lfiinPlarl1ite G3-307-4 ch~emistry)
I_________
FlaNe I iorizontal WtId 40 0.0 I
~~lLf MLIbuf
.t*,Luence aL_*
"ce alID Aternuation, 1!41 Fluen..ce Q 1/41 Fluenr Factor, FF Locallon Ftll 1/4t EFPY (nlcmn)
P2-.
(nfcrn)
Weld Q 225' 7.125 I.T61 46.0 1.10E,10 00652 7.58Et 17 0.36 i
Flante Hoirtontl Wald J. 125 1781 48.0 1.00EnO 0652 65*E,.01 0.00 Note:
The weld at 2250 is actually at the RPV 450 azimuth which is the lowest fluence azimuth in the quadrant. The fluence at the 225' azimuth is 2.33 times less that the peak flDuence. Therefore, the peak fluence is estimated above based on 2.71x1008/2.33 ý 1.16x]010.
The revised allowable flaw sizes are shown in Figure 3 as "Allowable Flaw Sizes for T = 194°F (e/t =
-0.38)", "Allowable Flaw Sizes for T = 194°7 (c/t = -0.22)", and "Allowable Flaw Sizes for T = 194°F (e/t = -0.20)". The APPENDA2 input file for this case is DI2D.IN, and is included in the computer files associated with this calculation. The results are documented in output files D 12D.SUM and D 12D.OUT. D I2D.SUM was incorporated into Excel spreadsheet "Allowable Flaw Sizes (D 12),xls".
All of these files are also included in the computer files associated with this calculation.
Revision M-"
0 1
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~&\\Q.1%4 i p
Figure 3. Results of Revised Analysis #2 for Lower-Intermediate Course Weld NMP-1 RPV Flaw Evaluation (Lower-Intermediate Course at 225°, Subsurface Flaw, Using Reduced Fluence and K1j) 0.
1.4 1.2 0.4 Flaw 9-015+016 777 IWB-3500 Acceptance Criteria IWB-3600 Allowable Flaw Size B
GE Flaw Handbook Results 0.2 X
Allowable Flaw Size for T = 194"F (ett = -0.38)
-0 Allowable Flaw Size for T = 194°F (elt = -0.22) - Allowable Flaw Size for T = 194°F (e/t - -0.20) 0.0 000 0.05 0.10 0.15 0.20 0.25 Flaw Aspect Ratio (alL)
.KeVISIOD I
U I
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6.0 CONCLUSION
S Based on the results shown in Figures 1, 2, and 3, the following conclusions can be made with respect to the RPV flaw evaluation:
/
Based on the results of Figures 1 and 2, the previous allowable flaw sizes were identically reproduced using the APPENDA program. Therefore, the methodology of evaluation for the current analysis is considered validated.
V Based on the results of Figure 1, the allowable subsurface flaw sizes for the Non-Beltline, Vessel Flange Horizontal Weld Region at 46 EFPY are reduced for a boltup temperature of 701F. This is a result of the lower allowable stress intensity factor, Kia, at 70TF versus 100°F.
For the limiting flaw eccentricity ratio of -0.17, which is the basis for the original flaw diagranm, in Reference [2], the as-found indications are acceptable compared to these lower allowable flaw sizes.
" Based on the results of Figure 2, the allowable subsurface flaw sizes for the Lower-Intermediate Course at 2251 Region at 46 EFPY are reduced for the revised pressure test temperature of 194TF. This is a result of the lower allowable stress intensity factor, Kla, at 194°F versus the temperature of 260'F used in the Benchmark Analysis, and also because the fluence is higher for 46 EFPY compared to the fluence for 28 EFPY used in the Benchmark Analysis. For the limiting flaw eccentricity ratio of -0.38, which is the basis for the original flaw diagram in Reference [2], the as-found indications are unacceptable compared to these lower allowable flaw sizes when K1a is used.
V Based on the results of Figure 2, for the Lower-Intermediate Course at 2250 Region at 46 EFPY, and using the actual flaw eccentricity ratios of -0.22 (Flaw 55) and -0.20 (Flaw 9-015+016), the as-found indications are also unacceptable compared to the lower allowable flaw sizes for the revised pressure test temperature of 194*F when Kia is used.
V Based on the results of Figure 3, for the Lower-Intermediate Course at 225' Region at 46 EFPY, and using the actual flaw eccentricity ratios of -0.22 (Flaw 55) and -0.20 (Flaw 9-015+016) as well as the weld-specific fluence and K,*, the as-found indications are acceptable compared to the allowable flaw sizes for the revised pressure test temperature of 194TF.
It is therefore concluded that the previously detected RPV flaws are dispositioned for the revised pressure test temperature of 194°F and are therefore acceptable for 46 EFPY (60 years of operation).
C i (4.
L-
7.0 REFERENCES
- 1. Niagara Mohawk Nuclear Engineering Calculation No: SOVESSELM030, Revision 1, "RPV Weld Flaw Evaluation Using GE Nuclear Energy NMP 1 RPV Flaw Evaluation Handbook (GENE-B 13-01805-124, Rev. 0)," SI File No. NMP-05Q-205.
- 2.
Niagara Mohawk Power Corporation Nuclear Engineering Report No. NER-I M-063, Revision 0, "GE Nuclear Energy RPV Flaw Evaluation Handbook for NMPI, GENE-B 13-01805-124," SI File No. NMP-05Q-203.
- 3.
AT[ Consulting Report No. ATI-05-034-001, "Calculation of P-T Operating Limit Curves for License Renewal for Nine Mile Point Units 1 and 2," October 2005, SI File No. NMP-09Q-291.
- 4.
USNRC Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," U. S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, (Task ME 305-4), May 1988.
- 5.
APPENDA, Computer Program for Performing Flaw Tolerance Analysis of Reactor Vessel Shells, Version 1.1, SIR-94-044, Revision 2, 4/24/95, Structural Integrity Associates.
- 6.
ASME Boiler and Pressure Vessel Code, Section HII, Appendix 1, 1989 Edition.
Revision 0
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NMP-05Q-303 Page 13 of 13
Sgr tucuuel Wfagvify AssLciateCs, Inlc.
55&S~d:V(2 ~) <)( &
t2~L r7 of 2)2-Suite 350 Centennial, CO 801 12-3bil Phone:
303-792.01)77 Fax:
1,03-792-2,68 www~stucikt.conm gswven@utfuctint.coil July 9, 2007 GLS-07-022 Mr. Roy Corieri Constellation Generation Group Nine Mile Point Nuclear Station Engineering Services Bldg. 1 348 Lake Road Oswego, NY 13126
Subject:
Flaw Proximity Assessment lior Nine Mile Point RPV Flaw Evaluation Rcferenccs:
I Structural Integrity Associates Calculation No, NMP-05Q-303, Revision I, "RPV Flaw Evaluation," 12/27/2005.
- 2. Niagara Mohawk Power Corpoxration Nuclear Engineering Report No. N ER-IM -063, Revision 0, "GE Nuclear Energy RPV Flaw Evaluation Hlandbook fbr 'NMPI, GENE-BI 3-01805-124," SI File No. NMP-05Q-203.
- 3. Niagara Mohawk Nuclear Engineering Calculation No. SOVESSELM030, Revision 1, "R*PV Weld Flaw Evaluation Using GE Nuclear Energy NMI" 1 PIV [Flaw Evaluation Hlandbook (GENE-B 13-01805-124, Rev. 0)," SI File No. NMP-O5Q-205.
- 4. ASME Boiler and Pressure Vessel Code,Section XI, "Rules ibr In-service lnspcction of Nuclear Power Plant Components," 1989 Edition.
Dea' Roy:
Per your request, this letter is to clarify and provide suppolling documentation ofea flaw proximity assessment related to the reactor pressure vessel (RP\\V) flaw evaluation documentedl in the Referetnce [I calculation.
BACKGROUND The Reference [I] calculation was completed in 2005 as a part olfConstellation's license renew.-l Ol'firts for the Nine Mile Point Nuclear Power Station, Unit I (NMP-t), That calculation provides a recon1ciliation Of'the previtus flaw evaluation perlimrled *br the NMP-I RIPV. An RPV flaw evaluation handbook was originally developed by GE in the Reftrence [21 document. Subsequent to that, during RPV exams fbr NMP-I in 2000, flaws were detected, and evaluation ofthose flaws was pertbi-ncd by Constellation fbr 28 effective full power years (EFPY) in the Reference [3] document. The Reli:rcnce Austin, TI Chiarlotte, NC H. Stornontoo, CT San Jose, CA 512-533 9191 704-597-55A W6f-59943250
.I08-978-8200 Silver Sping, MaD
- Sunrise, Ft.
301-4414;?00 Sr5M.5 M2.20 Uniontown, OH Mhiller CA 330.89%9753 562-944.-D210
Mr. Roy Coricri 2L.Iuly 9, 2007 GLS-07-022 Page 2 of"4 Flaw Proximity Asscssnienl for Nine Mile Point 101V Fl1w Evaluation
[I ] calculation provided a reviscd flaw evaluation for 46 EFPY (projected end-of life value for 60 years of operation).
Per Reisrence [3], the original flaw evalu, ation handbook work was done in accordance with the I c'83 Edition, Summer 1984 Addenda of Section XI ofthe ASME Code. The Rcfcrcnce [I] evaluation was pcr-frmed in accordance with the 1989 Edition of Section XI of the ASME Code, anti it was showa through a benchmark evaluation in the Reference [I J calculation that the previous work was identically reproduced. Theelbrc, the 1989 Edition ol'Section XI of the ASME Code is used in this assessment.
EVAILIJATION SI did not explicitly perl.brm a flaw proximity check in the Reference [I ] calculation at the end oftthe evaluation period (i.e., at 46 EFPY). It was implicitly assumcd that such a flaw proximity check was unnecessary, since all flaws evaluated were subsurlace (i.e., not exposed to the reactor environment),
and therefor'e the fatigue crack growth was negligible due to use of the ASME air fatigue crack grcwth curve. In this letter, it is further demonstrated that all relevant ASME Code rules wcrc satisfied ba:icd on a conservative assessment of the fatigue crack growth considering operation through 46 EFPY.
Note that the calculation that follows represents a bounding crack growth calculation fbr all possible allowable subsurlice flaw sizes bfr the RVWD-099 flaw diagram. This is because flaw diagrams must evaluate bounding crack growth ahead of time, since it is not known beforehand the size of a flaw dat may he detected during exa minat ions.
For flaw RVWD-099, the crack growth calculated in Reference [I I may be obtained ti'om supporti;.g tile "1)3D.SUM", which is identified in Section 4.0 of Refierence [I ], and is available in the supporting files associated with Reference [I]. The crack growth is 1.30x 10" inch fbr all subsurfiace liaws. F:or. thc litniting case, the temperature, ", is 70'F, as noted froi,, input file "D3D.IN" associatcd with Reference 1.1], where a constant through-wall temperature gradicilet was provided for the limiting stress distribution.
Thereobre, ihe crack growth is identical for all subsurface Ilaws since the temperature is unilbrm tl.rough the wall thickness.
The following provides the supporting calculation for this value (refer to Section XI of the ASME Code, Appendix A 141 tbr the equations used):
Fcr T,- 70'1, and a material RlNI) ofr40'F (per p. 4 or'Reference [Ij). the critical fracture toughness, Kl1. 1, is:
K l,,.liri,ni, -: 26.78 + 1.223 e (0.
<14 r
5
,(T.
'4 161)1 = 26.78 + 1.223 e 10.01 5*(T7 4'4 '(0)) = 46.0 ksi-hnch The allowable K1,, is:
IR fu r (o Appendix (;, G-21 10 of Re6,ercnce 1.41 Ikr the KI, equation, wihic incluiodes correction of the cxponen ial couelicic nt frtm. "1.233" to "1.223" per Figure 4-I ofWRC-175. This correclion was made in tater edlions of Section Xl.
Structural Intolrity Associates, Inc.
Mr. Roy Corieri qI July 9, 2007 GLS-07-022 Page 3 of 4 Flaw Proximily Assessment for Nine Mile Point RPV Flaw Evailuation Klua-llowabic = Kia-,'ilical / (Safety Factor) = 46.0 /,10 = 14.55 ksilinch Per A-4300 of Appendix A, assuming the load fluctuates between the above value and zero, the stress intensity factor range, AK, is:
AK-K,,m, - K.1,*, = 14.55 - 0 = 14.55 ksi',inch.
Per A-4300 of Appendix A, for a subsurface flaw, the air crack growth relationship applies:
da/dN = Co AK' The following coefficients were used2:
n = 3.07 Co = 1.99x10"11 S S = 25.72 (2.88 - R) 3" 7 = 3.703449 (assuming worst-case R = I)
Co 1.99x 1010 (3.703449) = 7.37x10 0" da/dN = 7.37x10w" (14.55)3.07 = 2.74x10,6 in/cycle Using dN = 463 cycles from p. 7 of Reference [1]:
da = da/dN * (No. of applied cycles) = 2.74x10-6 in./cycle (463 cycles) 1.30x 10"3 inch As stated previously, the above calculation is NOT for the actual flaw; rather, it is a bounding calculation made for the flaw handbook that bounds all possible allowable flaw sizes for the given temperature. Therefore, it provides a conservative and bounding value of crack growth compared to that for the actual flaw size.
From Table 1 of Reference [I], the smallest flaw depth of any flaws for RPV weld RVWD-099 is 0.396". Thus, the crack growth value calculated above represents 1.30xl 0,3/0.396 = 0,003 = 0.3% of the crack depth, which is insignificant. With respect to length, the SI flaw evaluation computer prograrn performs calculations assuming a constant flaw aspect ratio, a/I. Therefore, the percent change in flaw length is identical to the flaw depth, as demonstrated below.
For the same flaw, Table I of Reference [I] reveals an aspect ratio of 0.0144 and a flaw length, t, of 13.75". The change in length, Ae is therefore:
a/e = constant = (0.396/2) / 13.75 = 0.0144 = [(0.396 +/-1 1.30xl0"3)/2] / (13.75 + Ae) or Ae = 0.045-inch 2 The coefficients shown were obtained from the 1992 Addenda to Section XI and are used by Si's flaw evaluation computer program, as they are more recent and provide more conservative estimates of fatigue crack growth than the values specified il the 1989 Edition.
Structural Integrity Associate,;, Inc.
Mr. Roy Coricri July 9, 2007 GLS-07-022 Page 4 of 4 Flaw Proximity Assessment for Nine Mile Point RPV I'lm' Evaluation The change in length value calculated above represents 0.045 I/13.75 = 0,003 = 0.3% ofthe crack lengl h, which is insignificant.
Therefore, the change in both the depth and length is insignificant, so the flaw proximity does not diff cr significantly over the life oe the flaw.
Similar calculations to the above apply lbr the flaw diagram associated with flaw RVWD-140, wit) the exception that the limiting stress case tbr flaw RVWD-140 has a througtlwall tempcrature variation.
Theretore, depending on the flaw location within the wall thickness (based on t/t), the value of temperature, T, in the above calculation will be different. Similar results were achieved (a maximum fatigue crack growth value of 1.28x 10' inch was obtained in supporting lilt "'1)12C.SUM" tbr the other flaws evaluated in Refirence [ 1I).
CONClUAJSION Based on the eva*luation above, the change in flaw siZC (IuC to fatigue crack growth for the flaws evaluated in the Rcfrence [I I evaluation is negligible, so the initial flaw proximity check pciformeld on the flaws remains valid at the end of the evaluation period (46 EFPY).
Please do not hesitate to contact me if l you have any quest ions.
Prepar-ed BY:
Gary L.
evens, 1.E Senior Associate Gary I.- Stevens, P. E.
Senior Associate Karen K. Fujikawa,Qi r.
E.
Senior Associate AppjW() ted 8)':
cc:
NMP-05Q-406 Strac~turaI InlcgritY Associates, i0C.