ML12033A178
| ML12033A178 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 07/29/2011 |
| From: | Xcel Energy |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| L-MT-12-002, SIA 1000847.301 CA 11-003, QF-0549 (FP-E-CAL-01), Rev 7 | |
| Download: ML12033A178 (23) | |
Text
ENCLOSURE 5 MONTICELLO NUCLEAR GENERATING PLANT LICENSE AMENDMENT REQUEST REVISE THE TECHNICAL SPECIFICATIONS TO INCLUDE A PRESSURE TEMPERATURE LIMITS REPORT CALCULATION CA 11-003 EVALUATION OF ADJUSTED REFERENCE TEMPERATURES AND REFERENCE TEMPERATURE SHIFTS Non-Proprietary Version (SIA No. 1000847.301)
(23 pages follow)
Document Information NSPM Calculation (Doc) No: 11-003 I Revision: OA
Title:
Evaluation OF Adjusted Reference Temperatures and Reference Temperature Shifts Facility: Z MT [] PI Unit: [1 E12 Safety Class:
N SR El Aug Q Z Non SR Special Codes: F-Safeguards El Proprietary Type: Calc Sub-Type:
[NOTE:
Print and sign name in signature blocks, as required.
Major Revisions Z
N/A EC Number:
F1 Vendor Cale Vendor Name or Code:
Vendor Doc No:
Description of Revision:
Theacetalfollowinga calculationa lgbeQandreorattachments have been reviewed and deemed
[
Prepared by: (sign),
/ (print)
Date:
Reviewed by:(sign)
/ (print)
Date:
Type of Review: El Design Verification E] Tech Review El Suitability Review Method Used (For DV Only): El Review EL Alternate Calc nI Test Approved by: (sign)
I(print)
Date:
Minor Revisions ii N/A EC No: 18522
{ Vendor Calc:
Minor Rev. No: OA Description of Change: The revision is to update the references section of the calculation and include the non-proprietary copies of the calculation as well as the affidavits from vendors who have proprietary information in the proprietary copy.
Pages Affected: 12,13 The following calculation and attachments have been reviewed and deemed acceptable as a legible QA record Prepared by: (sign)
I 11A)
/,}jw I (print) By Vendor Date: 7/29/2011 Reviewed by: (sign)
/,(print)
Wynter McGruder Date: 916/2011 Type of Review-El Design Verification El Tech Review M Suitability Review Record Retention: Retain this form with the associated calculation for the life of the plant.
Method Used (For DV Only): El Review [] Alternate Calc 111 Test Approved by: (sign)
Z-Y/
(print) Paul Young Date: //-,
.//
Record Retention:.Retain this form with the associated calculation for the life of the plant.
This reference table is used for data entry into the PassPort Controlled Documents Module reference tables (C012 Panel). It may NOTE:
also be used as the reference section of the calculation. The input documents, output documents and other references should all be listed here. Add additional lines as needed by using the "TAB" key and filling in the appropriate information in each column.
Reference Documents (PassPort C012 Panel from C020)
Controlled*
Document Name Document Doc Ref Type**
Doc? + Type Number Rev INPUT OUTPUT 1
DIT 17500-2: Elevation information for M NGP vessel, BWRVIP 17500-2 N/A X
199NP 2
DIT 18522-1: GE Non-Proprietary Document Statement 18522-1 N/A X
3 Email from W. McGruder (Xcel Energy) to E. Houston ($I) Subject N/A N/A X
Contains EPRI Proprietary information, July 21, 2011 4
5
.6 7
8 10_
11 12 13 14 15 16 17 Controlled Doc marked with an "X" means the reference can be entered on the C012 panel in black. Unmarked lines willbe yellow. If marked with an 'W", also list the Doc Type, e.g., CALC, DRAW, VTM, PROC, etc.
Record Retention: Retain this form with the associated calculation for the life of the plant.
Mark with an WX" if the calculation provides inputs and/or outputs or both. If not, leave blank. (Corresponds to PassPort "Ref Type" codes: Inputs I Both =
"ICALC", Outputs "OCALC", Other I Unknown = blank)
Other PassPort Data Associated System (PassPort C011, first three columns)
OR Equipment References (PassPort C025, all five columns):
Facility Unit System Equipment Type Equipment Number MT 1
RPV Superseded Calculations (PassPort C019):
Facility Calc Document Number Title N/A Description Codes - Optional (PassPort C018):
Code Description (optional)
Code Description (optional)
Notes (Nts) - Optional (PassPort X293 from C020):
Topic Notes.
Text.
talc Introduction (pcopy directly from the calculation. Intro Paragraph or E See write-up below E] (Specify)....
Record Retention: Retain this form with the associated calculation for the life of the plant.
Monticello Specific Information ZYES
PLEX, RATE Structural Code(s) (See MT Form 3805):
Calculation:
Z No Require Fire Protection Review? (Using MT Form 3765, "Fire Protection Program Checklist", determine if a Fire Protection Review is required.) If YES, document the engineering review in the EC. If NO, then attach completed MT Form 3765 to the associated EC.
[
No Affect piping or supports? (If Yes, Attach MT Form 3544.)
Z No Affect IST Program Valve or Pump Reference Values, and/or Acceptance Criteria? (If Yes, inform IST Coordinator and provide copy of calculation.)
Record Retention: Retain this form with the associated calculation for the life of the plant.
Structural Integrity Associates, Inc.
File No.: 1000847.301 CALCULATION PACKAGE Project No.: 1100730-:
Quality Program: N Nuclear E Commercial PROJECT NAME:
Monticello P-T Curves Revision According to the PTLR Methodology.
CONTRACT NO.:
1005, Release 29 CLIENT:
PLANT:
Xcel Energy, Inc.
Monticello Nuclear Generating Plant CALCULATION TITLE:
Evaluation of Adjusted Reference Temperatures and Reference Temperature Shifts Non-Proprietary Version Document Affected Project Manager Preparer(s) &
Revision Pages Revision Description
- Approval, Checker(s)
Signature & Date Signatures & Date 0
1 - 11 Initial Issue A-i -A-3 Eric J. Houston Mark J. Jaeger 10/09/2010 10/08/2010 Vikram Marthandam 10/08/2010 I - 13 Added additional A-i - A-4 evaluation for two Eric J. Houston Nadia Crisan Computer files intermediate fluence EJH 1/11/2011 NC 1/07/2011 values corresponding to 36 and 40 EFPY Eric J. Houston EJH 1/07/2011 Mark J. Jaeger 1/07/2011 2
1-13 Modify References and A-i - A-4 Identification of Computer files Proprietary Information Eric J. Houston Eric J. Houston EJH 7/29/2011 EJH 7/28/2011 Mark J. fae7er 7/29/2011 Page -1 of 13 F0306-OIRI
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Table of Contents 1.0 IN TR O D U CTIO N...................................................................................................
3 2.0 M ETH O D O LO G Y..................................................................................................
3 3.0 D ESIGN IN PUT...................................................................................................
5 4.0 A SSU M PTION S....................................................................................................
7 5.0 CA LCU LA TIO N S...................................................................................................
8 6.0 C O N C L U SIO N S...........................................................................................................
8 7.0 REFEREN CE S.................................................................................................
12 APPENDIX A: FLUENCE CALCULATION...............................................................
A-i List of Tables Table 1: ART Values for MNGP RPV Components at 36 EFPY........................................
9 Table 2: ART Values for MNGP RPV Components at 40 EFPY...................................... 10 Table 3: ART Values for MNGP RPV Components at 54 EFPY...........................................
11 File No.: 1000847.301 Revision: 2 Page 2 of 13 F0306-OIR1
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1.0 INTRODUCTION
Radiation embrittlement of reactor pressure vessel (RPV) materials causes a decrease in fracture toughness. Nuclear Regulatory Commission (NRC) Regulatory Guide 1.99, Revision 2 (RG 1.99) describes general procedures to evaluate the effects of neutron irradiation embrittlement on the alloy steel used in RPV's. In order to perform this evaluation, RG1.99 requires calculation of Adjusted Reference Temperature (ART) and Reference Temperature Shift (ART, T) values [1]. The ART values are then used to determine the local fracture toughness of the RPV wall, according to ASMIE Code,Section XI, Non-mandatory Appendix G [2] evaluations.
The purpose of this calculation is to develop ART and ARTNDT values for the Monticello Nuclear Generation Plant (MNGP). In accordance with RG1.99, the ART and ARTNDT values are developed for all Reactor Pressure Vessel (RPV) plates, welds and nozzles exposed to fluence levels greater than 1.Oxl017 n/cm 2 [1]. This value is considered a lower bound, below which material effects due to irradiation are negligible, based on 10CFR50 Appendix H, Section III.A [3]. Based on updated fluence calculations, ART and ARTNDT values are provided at 36, 40, and 54 effective full power years (EFPY).
The reported valued for 54 EFPY are applicable until the end of MNGP's extended operation period (60 years). The intermediate values at 36 and 40 EFPY are provided due to operational challenges presented by the leak test temperature at 54 EFPY.
The application of assumptions is indicated throughout the document using a set of braces containing the appropriate reference number; for example, Assumption #3 would be indicated as {3, Section 4.0}.
2.0 METHODOLOGY When surveillance data is limited or not available, RG1.99 [1] specifies that ART is calculated with the following equation:
ART = InitialRTNDT + ARTNDT +Margin (1)
The "Initial RTNDT" term refers to the reference temperature of nil ductility transition for the non-irradiated material.
The reference temperature shift, ARTNDT, is defined in RG1.99 [1] as the shift in the reference temperature resulting from neutron irradiation. ARTNDT is calculated from the product of the chemistry factor (CF) and fluence factor (FF) as follows:
The CF is a function of the weight percent copper (Cu) and weight percent nickel (Ni) of the weld and base metal (plate or forging) materials. Tables 1 and 2 of RG1.99 [1] provide the standard CF values used in this calculation.
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The FF is based on the accumulated fast neutron exposure (E > I MeV), and is typically corrected by the thickness at the location of interest. The FF can be read directly from Figure 1 of RG1.99, or calculated using the following equation [1]:
FF = f o
28-010 log(f)
(3)
Due to attenuation effects, the fluence decreases with distance into the RPV wall. Per RG1.99 [1], the calculated or measured fluence from the inside surface of the RPV is attenuated using the following formula:
f s=fs e-0 24x (4)
Where:
f
=
fast neutron fluence (1019 n/cm2, E > 1 MeV) fuf = fast neutron fluence at the RPV inside surface (i.e., at base metal / cladding interface, same units as f) x
= depth into the RPV wall from the inside surface (inches)
For ASME Code,Section XI, non-mandatory Appendix G [2] evaluations, the "x" value is taken at one-quarter of the base metal thickness (1/4t). The fast neutron fluence can be attenuated through the stainless steel cladding on the inside surface of the RPV. By design, however, the cladding is treated purely as a lining, and not as a load-bearing member. Thus, for the purposes of this evaluation, the inside surface neutron fluence is considered to be at the base metal / cladding interface.
Margin (M), a conservative term defined in RG1.99 [1], accounts for uncertainty in the initial reference temperature and for variance in ARTNDT. The margin is calculated using the following formula:
Margin= 2.4o-1
+ 0-"
(5)
Where:
a,
= the standard deviation for the initial RTNDT (OF)
GA the standard deviation for ARTNDT (0F)
RG1.99 [1] states that the standard value of GA is 28 OF for welds and 17 OF for base metal (plates or forgings), and GA need not exceed 0.5 times the mean reference temperature shift (0.5* ARTNDT).
The a1 term, which is related to the uncertainty in the precision of the Initial RTNDT, is applied for values that are determined by measurement and also when generic or default values are used. For MNGP components where a aYl value is not explicitly identified, oy is assumed to be equal to 0 OF { 1, Section 4.0}.
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When surveillance data exists (e.g., the ISP Representative Material or other Supplemental Surveillance Program (SSP) material) containing an identical match for the heat number of the vessel beltline material being evaluated, a separate procedure is used to evaluate the ART. This procedure first determines the credibility of the data and, using best estimate chemistry values, calculates a fitted CF.
The fitted CF is then compared to the Table CF (defined above in Equation 2), and the greater of the two is used in subsequent ART calculations. If the surveillance data is credible, the margin (GA) may be cut in half, as specified in RG1.99 [1]. Detailed procedures to evaluate surveillance data in the manner described above can be found in Section 3 of Reference [4].
3.0 DESIGN INPUT The fluence values obtained from Reference [5] for 54 EFPY account for extended power uprate (EPU) operation at 2004 MWT from Cycle 22 on. Specific fluence values for RPV beltline components (i.e.
shell plates, welds, and nozzles) are not provided, as the table in Section 4.0 of Reference [5] merely lists the peak fluence value of 6.43x1018 n/cm2 at the inner surface of the RPV. Location specific fluence values for all EFPY must be calculated, along with peak fluence values for 36 and 40 EFPY.
The required input into these calculations is the flux and axial distribution of relative flux for both pre-EPU and post-EPU operation. The pre-EPU input is obtained from Reference [17] and the post-EPU input is obtained from Reference [5]. The fluence values are calculated in Appendix A, and summarized in Table A-1. Note that although references [5 and 17] are listed as proprietary in Section7, no proprietary information for those documents was used herein [20].
The MNGP RPV is constructed of a series of plates, numbered 10 through 17 from top to bottom [7].
Two plates are joined at each elevation via circumferential and vertical welds. According to Section 4.0 of Reference [5], at 54 EFPY the upper elevation of the RG1.99 fluence threshold (1.0x10 17 n/cm 2) is 168.7 inches above the bottom of active fuel (BAF). Reference [19] specifies that the BAF is at an elevation of 207.5 inches in the RPV, so the top of the beltline at 54 EFPY is at an elevation of 376.2 inches. Reference [7] specifies that the weld separating the lower intermediate shell plates (14 and 15) from the upper intermediate shell plates (12 and 13) is located at an elevation of 366.125 inches.
Therefore, the upper intermediate plates must be included in the ART evaluation.
The chemical composition of the MNGP RPV plates is obtained from several sources. The nickel content of the lower plates (C2193-1 and A0946-1) and upper intermediate plates (C2613-1 and C2089-1)is obtained from Reference [8]. The copper content of the lower plates is obtained from Table 4-1 of Reference [9]. Copper content is not available for the upper-intermediate plates; for conservatism, the bounding value of 0.35% copper specified in Section 1.1 of RG1.99 [1] is applied to these components-
{2, Section 4.0}.
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Reference [10] specifies updated copper and nickel values for the lower intermediate plates (C2220-1 and C2220-2); these values supersede an prior information for these components. Reference [10] also specifies
], which exceeds the default chemistry factor specified in the tables of Reference [1]. According to the discussion in the Attachment to Reference [10], [
Therefore, the aA margin term is cut in half for the lower intermediate plates.
Initial RTNTT values for the MNGP RPV plates are obtained from Table 5-1 of Reference [11]. In certain cases, multiple values are provided, based on different evaluation methods that are equally relevant. In such cases, it is assumed that selecting the minimum reported value is applicable for the ART calculations {3, Section 4.0}.
The vertical and circumferential welds that join the RPV plates must also be considered during the ART evaluation. Information on specific welds is not available; rather, Reference [12] provides parameters for a bounding beltline weld. Chemical composition information for the beltline weld is provided in Table 4-1 of Reference [12]. As described in Sections 3.1 and 3.2 of the same document, the Initial RTNDT value for the bounding beltline weld is calculated from 45 tests performed on a sample specimen.
The average calculated value is -65.6 OF, with a standard deviation of 12.7 'F. For the ART evaluation, these values are applied as the Initial RTNDT and a7, respectively.
According to the drawing in Reference [7], the centerline N-2 recirculation inlet nozzles in the MNGP RPV are located at an elevation of 186 inches above the bottom of the reactor vessel. According to Section 4.0 of Reference [5], at 54 EFPY the lower elevation of the 1.0x10 17 n/cm 2 fluence threshold will be 19.2 inches below the bottom of active fuel (BAF). This corresponds to an RPV elevation of 188.3 inches. However, the elevation of the uppermost blend radius of the N-2 nozzle is 204.3 inches, as shown in Reference [19]. Therefore, the N-2 nozzles must be included in the ART evaluation.
Similar to the upper intermediate shell plates, documentation of the copper content of the N-2 nozzles is not available. Section 3.2 of Reference [13] provides a conservative estimate of copper content based on a statistical evaluation of beltline nozzles in other BWR plants {4, Section 4.0}. Note that although
- Reference [13] is an EPRI proprietary document, EPRI does not consider the copper content of the N-2 nozzles proprietary [21 ]. Nickel content for each nozzle is identified in the RPV test reports in Reference [14]. The average of the reported values is 0.86%; this value, the best-estimate nickel content, is used to determine an N-2 ART value. The Initial RTNDT value is obtained from Table 5-2 of Reference [11], where a value of 40 OF is common to all of the N-2 nozzles.
Based on the boundary of the extended beltline [5,19] and examination of the RPV drawing [7], the N-2 nozzle is the only forged nozzle in the extended beltline at 54 EFPY. There are no instrument nozzles in the extended beltline at 54 EFPY.
The design inputs described above are replicated in the ART calculation results in Table 1 for 36 EFPY, Table 2 for 40 EFPY, and Table 3 for 54 EFPY.
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4.0 ASSUMPTIONS The assumptions made in order to define the evaluation approach and perform the analysis are summarized in the following list. The application of these assumptions is indicated throughout the document using a set of braces containing the appropriate reference number; for example, Assumption
- 3 would be indicated as {3, Section 4.0}.
- 1. According to RG1.99, the al term is equal to the standard deviation of the Initial RTNDT when that quantity is estimated from physical measurements [1]. However, for the MNGP evaluation, a number of components do not have a measured Initial RTNDT; rather, a bounding value is estimated via alternative means. Values calculated by this method include substantial conservatism, rendering it unnecessary to create additional conservatism via the a, term.
Consequently, for MINGP ART calculations, C7I is set equal to zero unless the Initial RTNTT for the component in question is estimated directly from measured data.
RG1.99 states that in cases where chemical composition is unknown, a conservative value of 0.35% copper may be used [1]. This approach is used herein to evaluate the ART values for the upper intermediate plates.
- 3. The Initial RTNDT values listed in Tables 5-1 and 5-2 of Reference [11] are calculated by one of four different methods, as described in the footnotes accompanying the tables. In many cases, the values reported in Reference [11] have been conservatively increased from the estimated value. Additionally, multiple evaluation methods are often applicable for a particular RPV component. All of the methods are valid, so it is assumed that the minimum initial RTNDT value reported for each component may be used for the ART evaluation. The values obtained by application of this assumption are consistent with those in MNGP's licensing basis documents.
- 4. Documentation of the copper content of the MNGP N-2 nozzles in unavailable. However, this information is available for beltline nozzles at other BWR plants. Section 3.2 of Reference [13]
offers an estimate of the copper content in nozzle forgin'gs by means of statistical evaluation of available industry forging data. It is assumed that this approach is conservative and therefore applicable for the purposes of MNGP ART calculations.
- 5. MNGP intends to implement EPU after the Spring 2011 refueling outage. In order to calculate future fluence values, it is necessary to calculate the EFPY for which the RPV is exposed to pre-EPU flux. Based on MNGP's operational history [15], an 81% cumulative load factor is assumed from MNGP startup up to April of 2011 (approximate end of the next refueling outage).
Thus, the EFPY corresponding to the end of the next refueling outage (41.25 years, from December 1970 to April 2011) is 33.4 (see Appendix A).
- 6. Intermediate fluence values are calculated for 36 and 40 EFPY. Reference [5] adds a 30% factor on the flux to account for unknown future operation. In projecting the fluence at 36 and 40 EFPY, this factor is removed from pre-EPU fluence calculations. Past operation is assumed to be bounded by pre-EPU flux without the 1.3 bounding factor.
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5.0 CALCULATIONS The methodology in Section 2.0 is used to evaluate the ART and ARTNDT values for MNGP, based on the design inputs in Section 3.0 and consistent with the assumptions in Section 4.0. The fluence estimates calculated in Appendix A and presented in Table A-I are applied where appropriate. The design inputs, intermediate calculations, and resultant ART values are provided for 36 EFPY in Table 1, 40 EFPY in Table 2, and 54 EFPY in Table 3.
6.0 CONCLUSION
S This document contains ART and ARTNDT values calculated in accordance with RG1.99 [1] for all MNGP plates, welds, and forgings exposed to fluence greater than 1.0X10 17 n/cm2. Design inputs are collected from a variety of sources, as discussed in Section 3.0. The calculated ART and ARTNDT values are provided for 36 EFPY in Table 1, 40 EFPY in Table 2, and 54 EFPY in Table 3.
The bounding ART value for the RPV plates and welds is 147.4 'F at 36 EFPY, 156.0 -F at 40 EFPY, and 186.6 'F at 54 EFPY. The ART value for the N-2 nozzles is 106.1 'F at 36 EFPY, 110.0 'F at 40 EFPY, and 125.2 'F at 54 EFPY.
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Table 1: ART Values for MNGP RPV Components at 36 EFPY RTNDT.('F)
~Cieristry FactorI cWi~wt %/01i; (W, "/)(
,%djqistm1ents For 14~
A RT,,, P!W rgin~eii cA(F) rý(?F Upper/Int Shell 1-12 Upoer/Int Shell 1-13 C2089-1 N/A C2613-1 N/A 0.0 27.0 0.35 1
0.50 1V3i n 4049 199.50 28.0 27.9 14.0 13.9 0.0 0.0 56.1 82.7 Lower/lnt Shelll114 C2220-1 N/A 1
27.0 r
F-.:...Lower/lntShell 1--5 C2220-2 NiA 27.0 103.4 8 8.5 I
1 1 147.4 103.4 8.5 0.0 147.4 Lower Shell 1-16 I nwp r Shill 1-17 A0946-1 I
N/A C2193-1 N/A 27.0 0.0 U.4 I U.bb 47.3 57.1 17.0 17.0 0.0 0.0 1.08.3 91.1 118.50 Chemi*+try justernts For 114t i
ao I NDT Margin Terms ARTNDT (F J(0F I-_
(
..,F ýJ SI 65ZZ.6 010 134.90 771.5_
28.01 12.7 73.4
-.-Ajustments~'-For1,,t ARTND Magn Aem RT~NID I (0F
.ý #A(f) f F) :.('F)D
-32.1
.16.0
'0.0-104.1 Fluence*atl/4t iFluence Factor,_FF,
... n 2,
(0.28 010 log f) 2 uppertint oneii FIlz D.Uo.+
I.ZOo I.
rz( I I U I.Oo Upper/Int Shell 1-13 5.063 1.266 1.97E+17 0.738 Lower/lnt Shell 1-14 5.063 1.266 2.77E+18 0.738 Lower/Int Shell 1-15 5.063 1.266 2.77E+18 0.738 Lower S hell 1-16 5.063 1.266 1.85E+18 0.738 Lower.Shell 1-17 5.063 1.266 1.85E+18 0.738 1.454E+17 1.454E+17 2.044E+18 2.044E+18 1.365E+18 1.365E+18 0.141 0.141 0.575 0.575 0.482 0.482 Limiting Weld - Beltline 5.063 1.266
.2.77E+18 0.738 2.044E+18 0.575 Bounding N-2 Nozzle 5.063 1.266 4.27E+17 0.738 3.151 E+1 7 0.226 File No.: 1000847.301 Revision: 2 Page 9 of 13 F0306-OIRI
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Table 2: ART Values for MINGP RPV Components at 40 EFPY Ceity Chemistry Adj -ustmentsFor 114t
- Hea, o
otNme Initial RTNDT (OF)
Factor' ARTNDT IMa'rgin'Terms ATO
_________~~
(wt%)[N i (wt%)
(0 )~
(OF)
&A(0F) cri(OF)
(0F)
Upper/Int Shell 1-12 Upper/Int Shell 1-13 C2089-1 C2613-1 N/A N/A 0.0 27.0 0.35 LowerllntShelll'14
' C2220-lI=
N/A 1
27.0 Lower/!ni Shell I-5:
C2220-2*,.
N/A 27.0 0.50 0.49 0.56 0.50 199.50 31.0 198.25 30.8 15.5 15.4 0.0 0.0 61.9 88.6 112.0 18.5 0.0 1560 112.0
-8.5
. 0.0 156.0 Lower Shell 1-16 A0946-1 C21 93-1 N/A N/A 27.0 0.0 U.14 0.17 118.50 51.9 62.7 17.0 17.0 0.0 0.0 112.9 96.7 C213-N4 0.0_
No.-
-Filler Material Initial RTNDT (OF)
Chemistry---
Cu(wt%) Ni(wt%)
Chemistry Factor:
- ff)j Adjustments For 1/4t ART'or Margin Terms ARTNDT
=("F)
E8018N
-:6.6y:J j0.1..o0
- 0.99 134.90 _
.9J 28.0 12.7I 79..8 Plate Location Initial RTN -'D (OF)
Chemistry-Cu (wt%:U/) NI (wt%)
Chemistry Factor 1' (OF)
Adjustments_:For !,1/4t-g"
'A'RTNIT MriTem AT (F
)
Va (OF)
(F Bounding N-2 Nozzle E21VW Y Plate 1-16 /1-17 40.0 0.18
.0.86 141.90 36.0 117.0
_0.0 110.0.:
,.Wall Thickness (in)
-ý..
,Fluence at ID ;ý-.Attenuation, 1/4t Fluence at /4t Fluence FactorFF 2-.
0.24x-2 02
ý10!g Full 1/4t!ý (n/cm )
e (ncm)(.8010ogf upperinnz nei F-iz Upper/nt Shell 1-13 Lower/Int Shell 1-14 Lower/Int Shell 1-15 Lower Shell 1-16 Lower Shell 1-17 5.063 5.063 5.063 5.063 5.063 5.063 1.266 1.266 1.266 1.266 1.266 1.266 2.30E+17 2.30E+17 3.36E+18 3.36E+18 2.28E+18 2.28E+18 0.738 0.738 0.738 0.738 0.738 0.738 1.698E+17 1.698E+17 2.48E+1,8 2.48E+18 1.683E+18 1.683E+18 0.155 0.155 0.622 0.622 0.529 0.529 Limiting Weld - Beltline 5.063 1.266 3.36E+18 0.738 2.48E+18 0.622 Bounding N-2 Nozzle 5.063 1.266 5.23E+17 0.738 3.86E+17 0.254 File No.: 1000847.301 Revision: 2 Page 10 of 13 F0306-OIRI
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Table 3: ART Values for MNGP RPV Components at 54 EFPY
- Lower/intS hellH-14 C222N A27.0 LoWer/2t shell 1-15 C22 "N/A 27.0 Lower Shell 1-16 A0946-1 N/A 27.0 1 nwair.qhPll L17 fl91OA-1 W/A 00 0.35 0.50 199.50 0.35 0.49 198.25 fli A M-0.14 0.56 9 8.20 n017 n rm 1 1A ri 1426 85* *0.0 186.6, 142.6 85 0.0 186.6-68.2 17.0 1 0.0 1129.2 823 170 n00 1183.
I I;IUUII I
F 40.0'
,---j_, 0.18
.I upper/Int *i1neii I-z12 Upper/Int Shell 1-13 Lower/Int Shell 1-14 Lower/Int Shell 1-15 Lower Shell 1-16 Lower Shell 1-17 5.063 1.266 5.063 1.266 5.063 1.266 5.063 1.266 5.063 1.266 4.06E+17 6.43E+18 6.43E+18 4.46E+18 4.46E+18 U. Io0 0.738 0.738 0.738 0.738 0.738 Z.lor_-I-i I 2.996E+17 4.746E+18 4.746E+18 3.292E+18 3.292E+18 U.L I U 0.219 0.792 0.792 0.694 0.694 Limiting Weld -Beltline 5.063 1.266 6.43E+18 0.738 4.746E+18 0:792 Bounding N-2 Nozzle 5.063 1.266 1.01E+18 0.738 7.454E+17 0.361 File No.: 1000847.301 Revision: 2 Page II of 13 F0306-OIRI
Structural Integrity Associates, Inc.
7.0 REFERENCES
- 1. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," May 1988.
- 2. American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, Rules for In-Service Inspection of Nuclear Power Plant Components, Appendix G, "Fracture Toughness Criteria for Protection Against Failure," 2004 Edition.
- 3. U.S. Code of Federal Regulations, Title 10, Energy, Part 50, "Domestic Licensing of Production and Utilization Facilities," Appendix H, "Reactor Vessel Material Surveillance Program Requirements," January 1, 2004 Revision.
- 4. BWRVIP-135, Revision 2." BWR Vessel and Internals Project, Integrated Surveillance Program (ISP) Data Source Book and Plant Evaluations. EPRI, Palo Alto, CA, 2009. 1020231. EPRI PROPRIETARY MATERIAL, SI File NO. B WRVIP-01-335P.
- 5. GE Hitachi Nuclear Energy Report No. 0000-0076-7052-RO, "Monticello Neutron Flux and Fluence Evaluation for Extended Power Uprate," Revision 0, December 2007, GE PROPRIETARY MATERIAL, SI File No. 1000847.204P.
- 6. Not Used.
- 7. Chicago Bridge and Iron Company Drawing No. 1, Revision 8, "General Plan, 17'2" I.D. x 63'-
2" Ins Heads Nuclear Reactor," NX-8290-13, SI File No. NSP-21Q-210.
- 8. Chicago Bridge and Iron Company Drawing No. R-7, Revision 0, "Skirt Knuckle, Heads & Shell
& Misc Heat Number Summaiy for 17'-2" ID x 63'-2" INS. HDS. Nuclear Reactor," NX-8290-133, SI File No. NSP-21Q-213.
- 9. GE Nuclear Energy Report No. SASR 88-99, "Implementation of Regulatory Guide 1.99, Revision 2 for the Monticello Nuclear Generating Plant," Revision 1, January 1989, SI File No.
NSP-21Q-202.
- 10. Letter from B. Carter (EPRI) to D. Potter (MNGP), "Evaluation of the Monticello 3000 Surveillance Capsule Data," BWR Vessel and Internals Project (BWRVIP), March 23, 2009, EPRI PROPRIETARY MATERIAL, SI File No. 1000207.202P.
- 11. Structural Integrity Associates, Inc. Report No. SIR-97-003, "Review of the Test Results of Two Surveillance Capsules, and Recommendations for the Materials Properties and Pressure-Temperature Curves to be used for the Monticello Reactor Pressure Vessel," Revision 3, March 1999, SI File No. NSP-21Q-401.
- 12. GE Nuclear Energy Report No. SASR 87-61, "Revision of Pressure-Temperature Curves to Reflect Improved Beltline Weld Toughness Estimate for the Monticello Nuclear Generating Plant," Revision 1,.December 1987, SI File No. NSP-21Q-201.
- 13. BWRVIP-1 73: BWR Vessel and Internals Project, Evaluation of Chemistry Data for BWR Vessel Nozzle Forging Materials. EPRI, Palo Alto, CA: 2007. 1014995. EPRI PROPRIETARY MATERIAL, SI File No. BWRVIP-01-373P.
File No.: 1000847.301 Page 12 of 13 Revision: 2 F0306-OIRI
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- 14. "Pressure Vessel Record, Exhibit D, Certified Test Reports," Author, Date, and Revision Not Identified, SI File No. NSP-21Q-233.
- 15. "Power Reactor Information System -PRIS." International Atomic Energy Agency (IAEA).
2009, accessed December 1, 2010, http://www.iaea.org/dbpage/.
- 16. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," March 2001.
- 17. GE Hitachi Nuclear Energy Report No. 0000-0024-2200-RO, "Monticello Neutron Flux and Fluence Evaluation," Revision 0, February 2004, GE PROPRIETARY MATERIAL, SI File No. 1000720.209P.
- 18. Not Used.
- 19. Design Information Transmittal from Monticello Nuclear Generating Plant to Structural Integrity Associates, DIT No. 17500-2, June 6, 2011, SI File No. 1100730.201.
- 20. Design Information Transmittal from Monticello Nuclear Generating Plant to Structural Integrity Associated, DIT No. 18522-01, July 22, 2011, SI File No. 1100730.201.
- 21. Email from W. McGruder (Xcel Energy) to E. Houston (SI), Subject :Contains EPRJ proprietary information, July 21, 2011, SI File No. 1100730.201.
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Appendix A:
FLUENCE CALCULATION File No.: 1000847.301 Revision: 2 Page A-I of A-4 F0306-O1RI
Structural Integrity Associates, inc.
METHODOLOGY Reference [5] presents the peak fluence at the RPV inside surface for 54 EFPY, which is applicable until the end of MNGP's extended operation period (60 years). This peak fluence value is calculated using both the pre-EPU and post-EPU flux, with EPU implementation conservatively modeled at 28.82 EFPY
[5]. Therefore, a linear interpolation of intermediate fluence values (i.e. 36 EFPY and 40 EFPY) based
.on 0 EFPY and 54 EFPY is overly conservative. Additionally, Reference [5] does not report specific fluence values for the RPV beltline components (i.e. shell plates, welds, and nozzles). The peak fluence at 54 EFPY is overly conservative for locations with an accumulated fluence nearer to the lower bound value of 1.Oxl017 n/cm2 in Reference [1]. Therefore, elevation-specific fluence values are calculated at 36 EFPY, 40 EFPY, and 54 EFPY in this Appendix. Note that the flux values in Reference [5] are calculated in accordance with NRC Regulatory Guide 1.190 [16].
The calculated peak fluence values at the RPV inner surface [5, 17] include an additional factor (F) of 1.3 to account for potential variation in future operation. In reproducing the 54 EFPY fluence values below, this factor is conservatively applied to both pre-EPU and post-EPU operation, consistent with Reference [5]. However, for the intermediate fluence calculations at 36 EFPY and 40 EFPY, this factor is only applied to the post-EPU fluence calculation (i.e. future operation). Past operation is assumed to be bounded by the pre-EPU flux {6, Section 4.0}.
The fluence calculations for MNGP follow these steps:
- 1. Benchmark the 54 EFPY peak fluence value calculated in Reference [5].
- 2. Calculate peak fluence for intermediate EFPY
- 3. Calculate location specific fluence for intermediate EFPY DESIGN INPUT The following inputs are required to calculate the intermediate fluence values:
- Current availability: MiNGP's cumulative load factor was 80.6% at the end of 2009 [15]. A cumulative load factor of 81% is assumed up to the next scheduled refueling outage {5, Section 4.0}. Based on a total of 41.25 years of operation (December 1970 to April 2011), the EFPY at the next scheduled refueling outage is taken as 33.4.
" Pre-EPU peak flux at RPV inner surface = 2.26x10 9 n/cm2-s [17, Section 3.1].
- Post-EPU peak flux at RPV inner surface = 3.70x10 9 n/cm2-s [5, Section 3.1].
" Pre-EPU axial distribution of relative flux at RPV inner surface at peak azimuth [17, Table A-2].
- Post-EPU axial distribution of relative flux at RPV inner surface at peak azimuth [5, Table A-2].
The bounding elevations of various RPV components are obtained by selecting the location of highest fluence for a particular component. Elevation is given relative to bottom of active fuel (BAF) using RPV geometry information in Reference [7] and elevations given in Reference [19].
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Elevations:
o Upper Intermediate Shell Plates = 158.6 inches o Lower Intermediate Shell Plate (elevation not required, this component receives peak flux) o Bounding Weld (elevation not required, this component receives peak flux) o Lower Shell Plate = 27.1 inches o N-2 Nozzles = -3.2 inches CALCULATIONS The following equation, which is consistent with the methodology of References [5 and 17], is used to calculate the fluence:
Fluence =
EFPY I. Rflux) pre-EPU + (flux EFPY. F. Rflux)posEPu (365.24.24.602)
Where:
flux
= peak flux for either pre-EPU or post-EPU EFPY = EFPY for either pre-EPU or post-EPU F
= factor to account for potential variation in operation Rflx
= relative flux, based on axial elevation above BAF For 54 EFPY, the following values are used in the equation above:
pre-EPU Flux = 2.26x10 9 n/cm2-s EFPY = 28.82 years F= 1.3 post-EPU Flux = 3.70xl 09 n/cm2-s EFPY = 25.18 years F= 1.3 The Rfu* term is dependent on both axial elevation above BAF and operating condition (i.e. pre-EPU or post-EPU). Therefore, fluence values are calculated for a range of elevations. A peak fluence value of 6.436x10 18 n/cm2 is obtained at an elevation of 80.95 inches above BAF, which compares well to the value of 6.43x10 18 n/cm2 calculated in Reference [5]. In order to maintain consistency with the peak end-of-life fluence, the elevation-specific fluence calculations at 54 EFPY use the inputs given above.
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For 36 EFPY, the process above is repeated using the following:
pre-EPU Flux= 2.26x10 9 n/cm2-s EFPY = 33.4 years F=I post-EPU Flux = 3.70x1 09 n/cm2-s EFPY = 2.6 years F= 1.3 For 40 EFPY, the process is identical except that the post-EPU EFPY is 6.6 years.
The results of the fluence calculations are presented in Table A-I. Calculation details are provided in the Excel spreadsheet 1000847.301.R1 Supporting File.xls, which is included with the electronic supporting files for this calculation package.
Table A-i: Fluence Values for RPV Components Component Fluence RPV Component 36 EFPY 40 EFPY 54 EFPY n/cm2 n/cm2 n/cm2 Upper Intermediate Shell Plates 1.97x10 17 2.30x10 17 4.06x10 17 (1-12 and 1-13)
Lower Intermediate Shell Plates 2.77x10" 3.36x101 6.43x108 (I-14 and 1-15)
Lower Shell Plates 1.85x101 2.28x10*
4.46x1018 (I-16 and 1-17)
Limiting Weld 2.77x108 3.36x1018 6.43x101 N-2 Nozzles 4.27x10 17 5.23x10 17 1.O1xl018 File No.: 1000847.301 Revision: 2 Page A-4 of A-4 F0306-O1R1