ML16187A296
| ML16187A296 | |
| Person / Time | |
|---|---|
| Site: | Dresden, Quad Cities  |
| Issue date: | 06/23/2016 |
| From: | Wong W Structural Integrity Associates |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML16187A321 | List: |
| References | |
| 00517760, Rev 62 and 63, RS-16-130 | |
| Download: ML16187A296 (44) | |
Text
ATTACHMENT 2 Structural Integrity Associates, Inc. File No. 1400735.301, Revision 1, "Finite Element Model Development and Thermal/Mechanical Stress Analyses for the N1 Nozzle," dated June 23, 2016
'l:J Structural Integrity Associates, Inc. CALCULATION PACKAGE PROJECT NAME: N-702 Evaluation for Dresden and Quad Cities CONTRACT NO.: 00517760, Rev 62 and 63 CLIENT: PLANT: File No.: 1400735.301 Project No.: 1400735 and 1501375 Quality Program:
0 Commercial Exelon Corporation Dresden and Quad Cities Generating Stations CALCULATION TITLE: Finite Element Model Development and Thermal/Mechanical Stress Analyses for the Nl Nozzle Document Revision 0 1 Affected Pages 1 -21 A-1 -A-2 1, 6, 11 Revision Description Initial Issue Removed Proprietary Markings Project Manager Approval Signature
& Date Richard Bax 1/22/16 Wilson Wong 6/23/16 Preparer(s)
& Checker(s)
Signatures
& Date Responsible Engineer:
Wilson Wong 1/22/16 Responsible Verifier:
MinjiFong 1/22/16 Responsible Engineer:
Wilson Wong 6/23/16 Responsible Verifier:
1hinf Minji Fong 6/23/16 Page 1 of21 F0306-01R2 e Structural Integrity Associates, Table of Contents 1.0 1NTRODUCTION
.........................................................................................................
4 2.0 OBJECTIVES
................................................................................................................
4 3.0 ASSUMPTIONS
............................................................................................................
4 4.0 DESIGN INPUTS ..........................................................................................................
5 4.1 Nozzle Geometry ...............................................................................................
5 4.2 Material Properties
.............................................................................................
5 4.3 Transient Definitions
.........................................................................................
5 5.0 METHODOLOGY
........................................................................................................
6 5.1 Thermal Transient Selection
..............................................................................
6 5.2 Nozzle FEM and Load Case Evaluation
............................................................
6 5.3 Model Validation
...............................................................................................
7 5.4 Post-Processing
..................................................................................................
7 6.0 ANALYSIS ....................................................................................................................
7 6.1 Bounding Transient Selection
............................................................................
8 6.2 Unit Internal Pressure Analysis .........................................................................
8 6.3 Thermal Transient Analyses ..............................................................................
9 6.3.1 Thermal Analyses ...............................................................................................
9 6.3.2 Thermal Stress Analyses ....................................................................................
9 6.4 Model Validation
...............................................................................................
9
7.0 CONCLUSION
............................................................................................................
10
8.0 REFERENCES
............................................................................................................
11 APPENDIX A FILENAMES
...............................................................................................
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e Structural Integrity Associates, lnc.!Jl: List of Tables Table 1: Bounding Transients for Analysis [9, 12] .................................................................
12 Table 2: Material Properties for Carbon Moly Steel [ 4, 5]. .....................................................
12 Table 3: Material Properties for Austenitic Stainless Steel [ 4, 5] ...........................................
13 List of Figures Figure 1: Components Included in the Finite Element Model ...............................................
14 Figure 2: 3-D Finite Element Model Mesh for Analyses, Baseline Mesh ..............................
14 Figure 3: Path Locations for Through-Wall Stress Extractions
..............................................
15 Figure 4: Applied Boundary Conditions and Unit Internal Pressure ......................................
15 Figure 5: Total Stress Intensity Plot for Unit Internal Pressure ..............................................
16 Figure 6: Applied Thermal Boundary Conditions for Thermal Transient Analyses ..............
17 Figure 7: Applied Mechanical Boundary Conditions for Thermal Stress Analyses ..............
18 Figure 8: Temperature Contour for SCRAM Transient at Time=6000 sec ............................
18 Figure 9: Stress Intensity Plot for SCRAM Transient at Time=6000 sec ..............................
19 Figure 10: Total Stress Intensity Contours for Mesh Sensitivity Study -Unit Pressure ....... 20 Figure 11: Linearized Membrane-Plus-Bending Stress Intensity History for Path 1, SCRAM and Interruption of Feed Flow ................................................................
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1.0 INTRODUCTION
Exelon intends to extend the applicability of Code Case N-702 [1] at both Dresden Units 2&3 and Quad Cities Units 1&2 through the end of their respective periods of extended operation (PEO). The Code Case allows for the reduction of in-service inspection from 100% to 25% of all Reactor Pressure Vessel (RPV) nozzle inner radii and nozzle-to-shell welds that must be performed every 10 years, including one nozzle from each system and pipe size, except for the feedwater and control rod drive return nozzles. Technical documents BWRVIP-108
[2] and BWRVIP-241
[3] provide the technical basis for the code case, but only consider 40 years of plant operation.
In order to extend the applicability of Code Case N-702 [1], a probabilistic fracture mechanics (PFM) evaluation, consistent with the methods of BWRVIP-108
[2] and BWRVIP-241
[3], must be performed to ensure the probability of failure remains acceptable.
The Nl (Recirculation Outlet) and N2 (Recirculation Inlet) nozzles are identified as the bounding RPV nozzles when fluence is not considered
[3]. Since the Nl nozzle does not meet the NRC's additional requirements, outlined in Reference
[2] and has a larger r/t ratio (and thus higher pressure stress) than the N2 nozzle, a bounding approach is used in this calculation to qualify all applicable nozzles using the Nl nozzle geometry.
2.0 OBJECTIVES
The objectives of this calculation package are to: 1. Develop a Finite Element Model (FEM) for the Nl nozzle and to, 2. Determine the stresses caused by applicable Service Level A and B thermal transients and internal pressure.
The stress distributions obtained as output of this analysis will be used as input for a subsequent probabilistic fracture mechanics (PFM) evaluation to be performed in a separate calculation package. 3.0 ASSUMPTIONS The following assumptions are made in this evaluation:
- The Nl nozzle-to-safe end weld is not specifically modeled. Instead the material instantaneously transitions from the nozzle material to the safe end material.
Since the location of stress extraction (see Figure 3) is not near the transition, the effect of this assumption on the analysis output is minimal.
- Density and Poisson's ratio for all materials are considered temperature independent.
In addition, typical values are assumed. This is consistent with the manner in which these properties are presented in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code.
- The nozzle-to-vessel weld is assumed to have thermal properties equal to the vessel material, since all carbon moly steels have the same thermal properties in the ASME Code [ 4, 5]. The modeled location and shape of the nozzle-to-vessel weld are approximate and for visualization purposes only. File No.: 1400735.301 Revision:
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- The inside surfaces are subjected to an essentially infinite convective heat transfer coefficient (HTC) of 10,000 Btu/hr-ft2-°F to maximize through wall thermal gradients and thus thermal stresses.
4.0 DESIGN
INPUTS This section introduces the design inputs used for the analyses documented in this calculation package. The following items are discussed separately below:
- Nozzle Geometry,
- Material Properties,
- Transient Definitions.
4.1 Nozzle
Geometry The recirculation outlet nozzles from Dresden, Units 2 and 3 are identical to the ones from Quad Cities, Units 1and2 when the nozzle drawings are compared [6, 7, pg. A-17]. The RPV inside radius (IR) is 125.6875 inches (to base metal, not including the 0.1875 inch cladding [6, 7, pg. A-1 and A-8]), with a vessel wall thickness of 6.125 inches [6, 7, pg. A-1 and A-8], resulting in an outside radius (OR) of 131.8125 inches. The Nl nozzle has an IR of 12.95 inches with cladding (13.1375 inches without cladding), with an outside radius (OR) of 26.5 inches on the vessel side and 14.6875 inches on the safe end side of the nozzle. The nozzle to vessel blend radius is 3.0 inches on the inside and 5. 7 5 inches on the outside [ 6, 7, pg. A-17]. Figure 1 shows the nozzle geometry with key dimensions identified.
4.2 Material
Properties The material component identifications are as listed:
- RPV shell:
- Nozzle forging: SA-302 Grade Bin accordance with Code Case 1339 (evaluated as carbon moly steel) [8, Section 5.3 .3 .1.1], Mn-Mo steel [6, 7, pg. A-16],
[6, 7, pg. A-16] Figure 1 identifies the applicable material type on an image of the nozzle model. The material properties are taken from the Design Code of Construction
[8 Section 5.3.l], the 1965 Edition through the Summer 1965 Addenda of the AS1\t1E B&PV Code,Section III [ 4]. Since this Edition of the Code does not list thermal conductivity or diffusivity, these values are obtained from the 1971 Edition of the Code [5]. All material properties used in the calculations are presented in Table 2 and Table 3. 4.3 Transient Definitions The thermal transient definitions are obtained from Reference
[9]. Only normal and upset (Service Level A and B) transients for the RPV and Nl nozzle specific transients are considered
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e Structural Integrity Associates, Inc.cs: 5.0 METHODOLOGY This section describes the methodology used to perform the analyses documented in this calculation package, which follows the same procedure as BWRVIP-108
& 241 [2, 3]. The Nl (Recirculation Outlet) and N2 (Recirculation Inlet) nozzles are identified as the bounding RPV nozzles when fluence is not considered (geometry only) [3]. Since the Nl nozzle does not meet the NRC additional requirements outlined in Reference
[2] and has a larger r/t ratio (and thus higher pressure stress) than the N2 nozzle, a bounding approach is used in this calculation to qualify all nozzles except for the feedwater and control rod drive return nozzles using the Nl nozzle geometry.
The general analytical methodology is introduced followed by specific discussion of the modeling method. The following process is used: 1. Select limiting Service Level A/B transient from RPV and nozzle thermal cycle diagram (TCD), 2. Build nozzle FEM and analyze bounding thermal transient and pressure load case, 3. Perform mesh sensitivity check and model validation checks, 4. Extract stresses for subsequent calculation.
5.1 Thermal
Transient Selection A bounding transient is selected for evaluation rather than analyzing all Service Level A/B transients defined on the RPV and recirculation outlet nozzle TCD. Separate load cases are defined for the internal pressure and thermal portions of the transient.
All analyses are performed using linear elastic methods for material models and loading; therefore, a single "unit" pressure load case is evaluated from which stress distributions at any pressure can be scaled from the results of the "unit" load case. The bounding thermal transient is selected based on maximum temperature fluctuation and rate of change. 5.2 Nozzle FEM and Load Case Evaluation All structural analyses are performed using the finite element method. A three-dimensional (3-D) finite element model is constructed using the ANSYS finite element program [10]. The model is used for linear elastic pressure and thermal transient stress analyses.
It is developed as a symmetric quarter model using the dimensions given in Reference
[ 6, 7], and includes a local portion of the RPV shell, the Nl nozzle-to-vessel weld, the Nl nozzle, and a portion of the attached safe end, as shown in Figure 1. The Nl nozzle-to-safe end weld is not modeled because it is sufficiently far from the region of interest to introduce any significant influence.
The model is meshed with the SOLID45 element type, of which the thermal equivalent is SOLID70. The mesh used in this calculation is depicted in Figure 2. A 1000 psig "unit" internal pressure load case is evaluated from which nozzle stress distributions for any desired pressure can be obtained by linearly scaling the results of the "unit" load case. Nozzle File No.: 1400735.301 Revision:
1 Page 6 of21 F0306-0!R2 e Structural Integrity Associates, Inc.(/!. piping loads are not considered since stresses in the blend radius and nozzle-to-vessel weld locations caused by these loads are insignificant according to Reference
[11, Section 5.5]. 5.3 Model Validation The FEM built for this analysis is validated using three separate checks: 1. Mesh size check To ensure the mesh provides stress results in the region of interest which are insensitive to the local mesh refinement, a second model is created utilizing twice the mesh density in the to-vessel weld and blend radius regions. The acceptance criterion for adequate mesh density is a change in peak stress intensity between solutions ofless than 1 %. 2. Time step check Excessively large time steps during thermal transient stress analysis may cause stress peaks to be missed. This can be checked by ensuring there are no sharp changes when plotting linearized through wall stress intensity time history results, and that there are multiple time points located in the peak and valley regions. 3. Far field stress check If the model boundaries are sufficiently far from the region of interest and if the boundary conditions are correct, hoop stresses in the RPV shell should be within a few percent of a hand calculation approximation.
The hoop and axial stress due to internal and end cap pressure can be approximated using the formula for thin walled cylinders:
Pr/t for hoop stress and Pr/2t for axial stress, where P is the internal pressure, r is the mean radius, and t is the wall thickness.
5.4 Post-Processing In support of future PFM analysis, four through-wall stress paths, two each at 0° and 90°, are defined within the region of the Nl nozzle blend radius and nozzle-to-vessel weld, as shown in Figure 3. Since the model is symmetric, these paths also represent the stress at 180° and 270°, respectively.
Stresses from the thermal transient and pressure load cases are extracted and saved in * .csv file format which can be imported to an Excel workbook for further processing (see Appendix A for file listings).
6.0 ANALYSIS
This section documents the results of the analyses described in Section 5.0 above. The following items are discussed in separate sections below: 1. Bounding Transient Selection, 2. Unit Internal Pressure Analysis, 3. Thermal Transient Analyses, 4. Model Validation.
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6.1 Bounding
Transient Selection The nozzle specific transient "Interruption of Feed Flow" located between Zone 9 and 10 [9a] has a instantaneous temperature down shock to 70°F from the normal operating temperature, before instantaneously returning to the normal operating temperature after only 60 seconds, bounding all Service Level AIB recirculation outlet nozzle specific transients.
This transient only applies to Dresden 2 and 3, but will be considered for Quad Cities 1 and 2 for bounding purposes.
Due to the severity of the transient, the vessel "SCRAM" transient located in Zone 10 to 11 of Region B [9b] is selected to bound all vessel transients in Reference
[9b]. These transients are tabulated in Table 1, and includes the effects of extended power uprate (EPU) [12]. 6.2 Unit Internal Pressure Analysis A unit internal pressure, P = 1,000 psi, is applied to the interior surfaces of the model. For the pressure run, cladding was removed since cladding cannot be credited as structural material per the ASME Code Section NB-3122 [5]. An end-cap load is applied to the free end of the nozzle piping in the form of tensile axial pressure, as calculated below. where, Peel p End cap pressure on attached piping free end (psi) Internal unit pressure (psi) IR1 OR1 Inside radius of modeled nozzle piping (in)= 13.1375 inches (w/o cladding)
[6, 7, pg. A-17] Outside radius of modeled nozzle piping (in) =14.6875 inches [6, 7, pg. A-17] The internal pressure also induces an end-cap load on the axial free end of the modeled vessel shell, as calculated below. p = P*IR/ _ 1000*125.6875 2 _ lO 016 . ec 2 (OR2 2-/R2 2)-(131.8125 2-125.6875 2)-' psi where, P ec2 End cap pressure on vessel shell axial free end (psi) P Internal unit pressure (psi) IR2 Inside radius of modeled vessel shell (in)= 125.6875 inches (w/o cladding)
[6, 7, pg. A-1] OR2 = Outside radius of modeled vessel shell (in)= 131.8125 inches [6, 7, pg. A-1] Symmetric boundary conditions are applied at the vessel's circumferential free end and the overall model's two planes of symmetry.
The free end of the nozzle piping and axial free end of the vessel shell are coupled in their respective axial directions to simulate the remaining portions of the geometry not included in the model. The applied load and boundary conditions for this case are shown in Figure 4. A representative stress intensity contour plot for the unit pressure analysis is shown in Figure 5. File No.: 1400735.301 Revision:
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{J Structural Integrity Associates, /nc.(/f. 6.3 Thermal Transient Analyses The transients to be analyzed are tabulated in Table 1. In order to achieve a final steady state condition, an arbitrary time of 3,600 seconds is added after the last load step of each transient, followed by an imposed steady state load step (at an arbitrary 60 seconds after the 3,600 seconds of additional time). 6.3.1 Thermal Analyses Bulk fluid temperatures and heat transfer coefficients are applied to the inside and outside surface nodes of the model. The inside surfaces are subjected to a conservative high convective heat transfer coefficient (HTC) of 10,000 Btu/hr-ft 2-°F. The heat transfer coefficient for the RPV and nozzle external surfaces of 0.2 Btu/hr-ft 2°F, along with an ambient temperature of 100°F, are typical values [13, Section 4.6.9.3] for all times during the transient.
The symmetry planes, top/bottom surfaces of the RPV, and "cut" plane in the piping are treated as adiabatic.
Figure 6 depicts a representative plot of the thermal boundary conditions applied for the transient analysis.
6.3.2 Thermal
Stress Analyses Symmetric boundary conditions are applied at the vessel's circumferential free end and the overall model's two planes of symmetry.
The free end of the nozzle piping and axial free end of the vessel shell are coupled in their respective axial directions to simulate the remaining portions of the geometry not included in the model. Figure 7 shows a representative plot of the mechanical boundary conditions applied for the thermal transient stress analysis.
A representative temperature contour and total stress intensity contour plot at an arbitrary time step for the SCRAM transient are shown in Figure 8 and Figure 9, respectively.
6.4
- Model Validation A unit internal pressure load with end cap loads and boundary conditions, as discussed in Section 6.2, was applied to both the base and double mesh density models. Figure 10 depicts both meshed models and shows that there is less than 0.1 % difference in the maximum stress intensities at the nozzle blend radius. Therefore, the mesh density originally chosen for this calculation does not need further refinement and is adequate for this calculation.
To validate that the time step sizes chosen for the thermal transient solutions described in Section 6.3 were sufficient, the linearized through wall stress intensity time history results from the thermal transient load cases along the four paths shown in Figure 3 are reviewed to ensure a smooth stress history response is obtained.
Representative plots of the Path 1 linearized membrane-plus-bending stress intensity history are shown in Figure 11 and clearly show the time points and resulting smooth stress response.
Peaks and valleys are clearly represented by multiple time points ensuring no peaks were missed and time stepping is adequate.
Using the formulae for thin walled cylinders (precise for R/t>> 10), the hoop stress for the vessel should be: File No.: 1400735.301 Revision:
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SJ Structural Integrity Associates, Inc.rs: where, Pboop = p p = P
- r = 1000*128.7525
= 21 020 si hoop (t) (6.125) ' p Thin wall cylinder predicted hoop stress (psi) Internal unit pressure (psi) r t = Mean radius of modeled vessel shell (in)= 128.75 inches (w/o cladding)
[6, 7] = Wall thickness of vessel (in)= 6.125 inches (w/o cladding)
[6, 7] Axial stresses should be half the hoop stresses.
Nodal query of hoop stress indicate a variance of less than 1 % from the thin wall formulae at the vessel mid-wall, which is acceptable since the r/t ratio for the vessel is approximately
- 21.
7.0 CONCLUSION
Unit pressure and thermal transient stress analyses have been performed.
Stress results were extracted from all analyses for through-wall paths at locations of interest along the Nl nozzle blend radius and nozzle-to-vessel weld in support of future PFM calculations.
All of the stress results are stored in computer files for later use (see Appendix A for file listings).
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8.0 REFERENCES
- 1. Code Case N-702, "Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell Welds,Section XI, Division 1," February 20, 2004. 2. Safety Evaluation of Proprietary EPRI Report, "BWR Vessel and Internal Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Inner Radius (BWRVIP-108)," December 19, 2007, SI File No. BWRVIP.108P.
- 3. BWRVIP-241:
BWR Vessel Internals Project, Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii, EPRI, Palo Alto, CA. 1021005NP.
- 4. ASME Boiler and Pressure Vessel Code,Section III, 1965 Edition with Summer 1965 Addenda 5. ASME Boiler and Pressure Vessel Code,Section III, 1971 Edition. 6. Babcock & Wilcox Certified Design Document for Quad Cities I&II, Contract No. 610-0122-51/52, SI File No. CEC0-36Q-210.
- 7. Babcock & Wilcox RPV Certified Stress Reports: a. Dresden 2, Contract No. 610-0098-51, Rev. 1, SI File No. 1400735.205
- b. Dresden 3, Contract No. 610-0111-51, Rev. 2, SI File No. 1400735.206.
- 9. Thermal Cycle Diagrams:
- a. General Electric Drawing No. 158B7279, Sheet 1, Revision 1, "Nozzle Thermal Cycles (Recirculation Outlet)," SI File No. 1400735.203.
- b. General Electric Drawing No. 921D265, Sheet 1, Revision 1, "Reactor Thermal Cycles," SI File No. 1400735.209.
- 10. ANSYS Mechanical APDL and PrepPost, Release 14.5 (w/ Service Pack 1 UP20120918), ANSYS, Inc., September 2012. 11. BWRVIP-108NP:
BWR Vessel and Internals Project, Technical Basis for the-Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii. EPRI, Palo Alto, CA: 2007. 1016123, SI File No. BWRVIP.108NP.
- 12. General Electric Design Specifications
- a. Report No. 26A5588, Revision 0, "Reactor Vessel -Power Uprate, Quad Cities 1 & 2," SI File No. 1400735.201.
- b. Report No. 26A5587, Revision 0, "Reactor Vessel -Power Uprate, Dresden 2 & 3," SI File No. 1400735.201
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e Structural Integrity Associates, Inc.& Table 1: Boundin2 Transients for Analysis [9, 121 Inside Outside Time, Tnozzle, Tvessel, Pressure, Surface Surface Description h, h, sec OF OF psig Btu/hr-Btu/hr-ft 2-°F ft 2-°F 0 530 530 SCRAM 4320 400 400 8640 530 530 0 530 530 1000 10000 0.2 Interruption of Feed 0.1 70 530 Flow 60.l 70 530 60.2 530 530 Notes: 1. A total of 3,660 seconds are added to the end of each transient followed by an imposed steady state condition 60 seconds later (see Section 6.3 of this calculation).
- 2. Flow rates are not considered since the inside surface heat transfer film coefficients are essentially infinite.
Table 2: Material Properties for Carbon Moly Steel [4, 5] Temperature, Young's Mean Thermal Modulus, Expansion, OF x10 6 psi xl0*6 in/in/°F 70 29.9 6.10 200 29.5 6.38 300 29.0 6.60 400 28.6 6.82 500 28.0 7.02 600 27.4 7.23 Density, p = 0.283 lb/in 3 , assumed temperature independent.
Poisson's Ratio, u = 0.3, assumed temperature independent.
Thermal Conductivity, x10*4 Btu/sec-in-°F 8.33 8.07 7.69 7.31 6.90 6.55 Specific Heat, Btu/lb-°F 0.113 0.117 0.121 0.124 0.128 0.132 Note: Specific Heat values are derived from the equation shown in General Note (5) of Table 1-4.0 [5], Specific Heat= TC I (TD x density).
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e Structural Integrity Associates, Table 3: Material Properties for Austenitic Stainless Steel (4, 5] Temperature, Young's Mean Thermal Modulus, Expansion, OF xl0 6 psi xI0*6 in/in/°F 70 27.4 9.20 200 27.1 9.34 300 26.8 9.47 400 26.4 9.59 500 26.0 9.70 600 25.4 9.82 Density, p = 0.29 lb/in 3 , assumed temperature independent.
Poisson's Ratio, u = 0.31, assumed temperature independent.
Thermal Conductivity, x10*4 Btu/sec-in-°F 1.93 2.06 2.16 2.27 2.37 2.48 Specific Heat, Btu/lb-°F 0.111 0.115 0.117 0.120 0.123 0.125 Note: Specific Heat values are derived from the equation shown in General Note (5) of Table 1-4.0 [5], Specific Heat= TC I (TD x density).
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e Structural Integrity Associates , Inc. 125.6875 in. IR 13.1375 in. IR 0.1875 in. Cladding ER308L (Auste niti c SS) Figure 1: Components Included in the Finite Element Model (Inside radius (IR) dim e n sions do n ot include t he 0.1875 inch thick claddi n g.) WMlmi t-J>T !JN Ures Q.: Vi :-no1 !'1:xl<'!
I 31.l Figure 2: 3-D Finite Element Model Mesh for Analyses , Baseline Mesh F il e No.: 1400735.301 R evision: 1 Page 14 of21 F0306-0IR2 e Structural Integrity Assoc i ates , Inc. Figure 3: Path Locations for Through-Wall Stres s Extraction s WMll!S t"M.tJH -:0016.2 _0192.13 -756S.12 _6344_1 U"'s CJ; 1 Vi '°""' l1:xle 1 3J Figure 4: Applied Boundary Condition s and Unit Internal Pre ss ure (Un i ts for P r essu r e i n psi) F il e No.: 1400735.301 R ev i s i o n: 1 Page 15 of21 F0306-0 I R2 e Structural Integrity Associates , Inc. IDJ.1L s:i.:..: !'ITP',P.;
..... TMF SlNT (A.T.;J IOS-O QIX *.U'793 -..?..tl9.:ZJ *57061.Z 2059.21 Ures Q:: 'II Vi. Figure 5: Total Stress Inten s ity Plot for Unit Internal Pressure (Units for stress intensity in psi) F ile No.: 1400735.301 Revision:
1 Page 16 of2 1 F 0306-0 IR2 e Structural Integrity Associates , Inc . .
.oc 2: 44 .0°"2s 1 .JCMJ .Jl5004 .o: 7117 .JlY29 Ures rt; 1 Vi. moz 1-tde l 3lJ ll1>tl '['fi'[.11.M . r 100 Dres :il Vi a) Heat transfer coefficient (HTC) b) Bulk temperature (TBULK) 4 8/ .??J 53C Figure 6: Applied Thermal Boundary Conditions for Thermal Transient Analyses (E nd of Si n g le R e l ief o r S af et y V a l ve Bl ow d ow n S C RAM Tra n s i e nt s h ow n) (Units for HTC in Btu/sec-in 2-°F , TBULK in °F) File No.: 1400735.301 Revision:
1 Page 17 of21 F0306-0 I R2 e Structural Integrity Assoc i ates , Inc. Ores '11 Vi moz 1-tdel .lJ Figure 7: Applied Mechanical Boundary Conditions for Thermal Stre ss Analy s es ltD.11L.s::a.:..TI.:lt SIE!'-3 S\ll *' 1"ll-L'"t>.J.IU ID!? ::..'<<
4?9.?44 43 1.6 433.955 436.3:1 438.6 Dres <;I:. '11 Vi rro7 1-txlel 30 Figure 8: Temperature Contour for SCRAM Tran s ient at Time=6000 s ec. (Un i ts fo r temperature in °F) F il e No.: 1400735.301 Rev i s i o n: 1 Page 1 8 of21 F0306-0 I R2 e Structural Integrity Associates , Inc. 145.8C4 4910.: 3 967<.46 14438_8 19203.1 73 1.9 334%.1 2826;*4 43C24.7 Ores :-11 Vi '"'"'°' 1-trle 1 3J Figure 9: Stress Intensity Plot for SCRAM Transient at Time=6000 sec. (Units for stress intensity in psi) F il e No.: 1400735.301 Re v i s i o n: 1 P age 19 of2 1 F0306-0IR2 e Structural Integrity Assoc i a t es , Inc. ____ __, f ine Mes h Figure 10: Total Stress Intensity Contours for Mesh Sensitivity Study -Unit Pres s ure (Units for stress intens i ty in psi) F il e No.: 1400735.301 R ev i s i o n: 1 P age 20 of21 F0306-0 IR2
{J Structural Integrity Associates, 10000 9000 8000 *v; 7000 a. > 6000 +' *v; c: 5000 Q) +' E 4000 V\ V\ Q) ..... 3000 +' V\ 2000 -+-Sint{I) 1000 -+-Sint(O) 0 0 2000 4000 6000 8000 10000 12000 Time (s) SCRAM 25000 20000 *v; a. ?;" 15000 *v; c: Q) +' c: V\ 10000 V\ Q) ..... +' V\ 5000 -+-Sint{l) -+-Sint(O) 0 0 so 100 150 200 250 300 Time (s) Interruption of Feed Flow Figure 11: Linearized Membrane-Plus-Bending Stress Intensity History for Path 1, SCRAM and Interruption of Feed Flow Note: SINT(I) and SINT(O) refer to the membrane-plus-bending stress intensity on the inside and outside surfaces of the model , respecti v el y. File No.: 1400735.301 Revision:
1 Page 21 of21 F 0306-0 IR2 i)structural Integrity Associates, Inc.rs: File No.: 1400735.301 Revision:
1 APPENDIX A FILENAMES Page A-1 of A-2 F0306-0IR2 e Structural Integrity Associates, Inc.&. File Name STACK.INP Dres_QC_N1 .INP Dres_QC_N1_x2.INP MProp_Linear_Dres_QC.INP THM_Dres_QC_
1.INP THM_Dres_QC_2.INP Dres_QC_PRESS.INP Dres_QC_PRESS_X2.INP THM_Dres_QC_
1_mntr.inp THM_Dres_QC_2_mntr.inp CMNTR.MAC GenStress.mac GETPATH.TXT STR_*_COE_P?.CSV File No.: 1400735.301 Revision:
1 Description Controller input file to run thermal and mechanical analyses Input file to construct the 3-D model for linear-elastic analysis Input file to construct the 3-D model with Refined Mesh Input file of temperature dependent linear elastic material properties Analysis input file for SCRAM Transient Analysis input file for Int. Feed Flow Transient Analysis input file for Unit Internal Pressure Analysis input file for Unit Internal Pressure using Refined Mesh Load step definition file from thermal analysis -SCRAM Load step definition file from thermal analysis -Int. Feed Flow Thermal temperature time history extraction macro file to create THM
- mntr.inp files Path stress extraction macro file to extract .CSV files Through-wall path definition file Curve fit coefficients outputs of stresses in tabulated forms *=PRESS, Dres_QC_ 1, Dres_QC_2
? =path number 1-4 Page A-2 of A-2 F0306-0IR2 ATTACHMENT 3 Structural Integrity Associates, Inc. File No. 1400735.302, Revision 1, "Code Case N-702 Evaluation for Dresden and Quad Cities Recirculation Outlet (N1) Nozzle," dated June 23, 2016 l> Structural Integrity Associates, Inc. File No.: 1400735.302 Project No.: 1400735 and 1501375 CALCULATION PACKAGE Quality Program: Nuclear D Commercial PROJECT NAME: N-702 Evaluation for Dresden and Quad Cities CONTRACT NO.: 00517760 Rev 62 and 63 CLIENT: PLANT: Exelon Corporation Dresden and Quad Cities Generating Stations CALCULATION TITLE: Code Case N-702 Evaluation for Dresden and Quad Cities Recirculation Outlet (Nl) Nozzle Document Affected Project Manager Revision Pages Revision Description Approval Signature
& Date 0 1 -17 Initial Issue A-1 -A-2 Daniel Sommerville 1/22/16 1 3, 11, 12, 14 Removed Proprietary Markings and Changed References 6 and 15 Wilson Wong 6/23/16 Preparer(s)
& Checker(s)
Signatures
& Date Resuonsible Engineer:
Wilson Wong 1/22/16 Resuonsible Verifier:
Jim Wu 1/22/16 Resuonsible Engineer:
Wilson Wong 6/23/16 Resuonsible Verifier:
Jim Wu 6/23/16 Page 1of17 F0306-01R2 SJ Structural Integrity Associates, lnc.(g Table of Contents
1.0 INTRODUCTION
.........................................................................................................
3 2.0 OBJECTNE ..................................................................................................................
3 3.0 METHODOLOGY
........................................................................................................
3 3.1 Fatigue Cycles ...................................................................................................
4 3.2 Probabilistic Fracture Mechanics Evaluation
...................................................
.4 4.0 DESIGN INPUT ............................................................................................................
5 4 .1 Deterministic Parameters
...................................................................................
5 4.1.1 ISI. ......................................................................................................................
5 4.1.2 Stresses ..............................................................................................................
5 4.1.3 Fatigue Cycles ...................................................................................................
5 4.2 Rando1n Variables
.............................................................................................
6 4.2.1 SCC Initiation
....................................................................................................
6 4.2.2 SCC Growth .......................................................................................................
6 4.2.3 Fatigue Crack Growth ...................................................
- ...................................
7 5.0 STRESS RESULTS AND FATIGUE CYCLE LOADINGS .......................................
8 6.0 ASSUMPTIONS
............................................................................................................
9 7.0 RESULTS OF ANALYSES ........................................................................................
10
8.0 CONCLUSION
S
.........................................................................................................
10
9.0 REFERENCES
............................................................................................................
11 Appendix A LIST OF SUPPORTING FILES ......................................................................
A-1 List of Tables Table 1: Deterministic Parameter Summary ............................................................................
13 Table 2: Probability of Detection Distribution
[14] ................................................................
13 Table 3: Random Variables Parameter Summary for Nl Nozzle ............................................
14 Table 4: PoF for Period of Extended Operation
.....................................................................
15 List of Figures Figure 1: Stress Extraction Path Orientations in the Nl Nozzle ............................................
15 Figure 2: Pressure Stress Distributions for the N-702 Evaluation
..........................................
16 Figure 3: Full Power Thermal Expansion Stress Distributions for the N-702 Evaluation
..... 16 Figure 4: Bounding Transient Stress Distributions for the N-702 Evaluation
.......................
17 Figure 5: Weld Residual Stress Distributions for Paths 2 and 4 .............................................
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1.0 INTRODUCTION
Exelon intends to extend the applicability of Code Case N-702 [1] at both Dresden Units 2&3 and Quad Cities Units 1&2 through the end of their respective periods of extended operation (PEO). The Code Case allows reduction of in-service inspection from 100% to 25% of all nozzle blend radii and to-shell welds every 10 years, including one nozzle from each system and pipe size, except for feedwater and control rod drive return nozzles. Technical documents BWRVIP-108
[2, 3] and BWRVIP-241
[4] provide the basis for the code case, but only consider 40 year plant operation.
In order to extend the applicability of Code Case N-702, a probabilistic fracture mechanics (PFM) evaluation, consistent with the methods ofBWRVIP-108 and BWRVIP-241, is performed to ensure that the probability of failure remains acceptable.
The Nl (Recirculation Outlet) and N2 (Recirculation Inlet) nozzles are identified as the bounding nozzles when fluence is not considered[
4]. Since all 4 units under consideration have identical Nl and N2 geometries and the Nl nozzle has a larger r/t ratio (and thus higher pressure stress) than the N2 nozzle, a bounding approach is used in this calculation to qualify all applicable nozzles in which the Nl nozzle geometry is considered in conjunction with the bounding fluence from the applicable nozzles. The evaluation consists of two parts: Finite Element Model (FEM) Stress Analysis and Probabilistic Fracture Mechanics (PFM) Analysis.
The FEM stress analysis is performed in a separate calculation
[5] while this calculation package docwnents the PFM analysis.
2.0 OBJECTIVE
The objective of the evaluations documented in this calculation package is to perform a plant specific analysis of the bounding Dresden/Quad Cities Nl nozzle to extend applicability of the existing reljef request to 60 years of operation, or 54 effective full power years (EFPY). 3.0 METHODOLOGY This evaluation considers the nozzle-to-shell weld and nozzle blend radius on the Nl nozzle per Reference
[3] and [4] and confirms that the nozzle still meets the acceptable failure probability considering the bounding fluence at the end of the PEO. Reference
[6] shows the highest fluence at 5.59x10 17 n/cm 2. The acceptance criterion limits the difference in probability of failure per year due to the low temperature over pressure (LTOP) event to be no more than 5x 1 o-6 when changing from full (100%) service inspection to 25% inspection for the PEO. In this analysis, the conservative case of zero inspection for the first 40 years with 25% inspection for the PEO is used. If the resulting probability of failure per year due to an LTOP event (including 1 x 10-3 probability of LTOP event occurrence per year [3, pg 5-13]) is less than 5x 10-6 , then no comparison to the full inspection case is required.
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1 Page 3of17 F0306-01R2 SJ Structural Integrity Associates, Inc.lit 3.1 Fatigue Cycles In the FEM calculation
[ 5], two bounding transients were defined to conservatively include fatigue crack growth contributions from all the normal and upset (Service Levels A and B) thermal transients defined for the Nl nozzles. The nozzle specific transient "Interruption of Feed Flow" located between Zone 9 and 10 of the nozzle thermal cycles [11] has a 450°F temperature shock and returns to 520°F after only 60 seconds, bounding all Service Level A/B recirculation outlet nozzle specific transients.
This transient only applies to Dresden 2 and 3, but will be considered for Quad Cities 1 and 2 for bounding purposes.
Due to the severity of the transient, the vessel "SCRAM" transient located in Zone 10 to 11 of Region B [12] is selected to bound all vessel transients in Reference
[12]. The number of cycles used in the N-702 evaluation is defined in Section 5.0. 3.2 Probabilistic Fracture Mechanics Evaluation The probabilistic evaluation is performed for the case of25% inspection for the extended operating period (with zero inspection coverage conservatively assumed for the initial 40 years of operation).
For the nozzle blend radius region, a nozzle blend radius crack model [17] is used in the probabilistic fracture mechanics evaluation.
For this location and crack model, the applicable stress is the stress perpendicular to a path defined 90 degrees from the tangent drawn at the blend radius. For the nozzle-to-shell weld, either a circumferential or an axial crack, depending on weld orientation; can initiate due to either component fabrication (i.e. considering only welding process) or stress corrosion cracking.
From BWRVIP-05
[8], it is shown that the probability of failure for a circumferential crack is less than an axial crack, due to the difference in the stress (hoop versus axial) and the influence on the crack model. However, this probabilistic fracture mechanics evaluation for the nozzle and vessel shell weld considers both circumferential and axial cracks (depending on weld orientation).
An axial elliptical crack model with a crack aspect ratio of all= 0.5 is used in the evaluation for the nozzle-to-shell weld. The inspection probability of detection (PoD) curve from BWRVIP-05
[8] (Table 2) is utilized with a ten year inspection interval.
The calculation of stress intensity factor is at the deepest point of the crack. The approach used for this evaluation is consistent with the methodology presented in BWRVIP-05
[8]. A Monte Carlo simulation is performed using a variant of the VIPER program [9]. The Monte Carlo method can be used to solve probabilistic problems using deterministic computation.
A mean value, std devation, and distribution curve as defined in the random variables summary (Table 3) defines a set of possible inputs and their probabilities of occurring.
Using this domain of possible inputs, a set of inputs are generated for use in determining whether the nozzle will fail using conventional deterministic fracture mechanics methodology.
This is repeated 20 million times. The number of simulations in which the nozzle is determined to fail divided by the number of simulations run gives the probability of failure .. File No.: 1400735.302 Revision:
1 Page 4of17 F0306-01R2
Integrity Associates, lnc.oc The VIPER program was developed as part of the BWRVIP-05 effort for Boiling Water Reactor (BWR) reactor pressure vessel (RPV) shell weld inspection recommendations.
The software was modified into a separate version, identified as VIPERNOZ, for use in this evaluation.
The detailed description of the methodology incorporated in the VIPER/VIPERNOZ program is documented in References
[8] and [3]. The modified software for this project is identified as VIPERNOZ to distinguish from the original VIPER software, and is verified on a project specific basis [7] to ensure the modifications made to the VIPER software are fully quality assured. 4.0 DESIGN INPUT The plant specific input is described below. Section 4.1 presents all inputs modeled deterministically as constants while Section 4.2 describes the probabilistic treatment of inputs considered to be random variables (RV) in the VIPERNOZ code. 4.1 Deterministic Parameters Table 1 summarizes the dimensional and operational inputs used in the N-702 evaluation
[10, 11, and 12]. Subsections 4.1.lthrough
4.1.3 describe
the more detailed input parameters used for in service inspection (ISI) interval, stress distributions and fatigue cycles, respectively.
4.1.1 !SI In this analysis, the conservative case of zero inspection for the first 40 years with 25% inspection for the PEO is used. The probability of detection (POD) distribution function associated with inspection is shown in Table 2 [14]. 4.1. 2 Stresses Stresses due to vessel pressure and bounding thermal transients are determined in the Finite Element Model Development and Thermal Mechanical Stress Analyses for the Nl nozzle [5]. In that calculation package, through wall stress distributions are presented at four locations in the region of the Nl nozzle for use in the N-702 evaluation.
Figure 1 shows the locations and orientations of these four through-wall stress paths. For vessel pressure, an internal pressure of 1,000 psig is applied to the inside surfaces of the RPV and Nl nozzle FE model. A bounding transient is also analyzed and the maximum cyclic stress ranges, based on a linearized through wall stress distribution, are identified.
Figures 2 through 4 show the distributions of the stress component acting normal to the crack plane (e.g. hoop or axial depending on the Path location) for the unit pressure, full power thermal expansion (steady state first load step of transient analysis) and the bounding transient load case step, respectively.
Details of the analysis can be found in [5]. 4.1.3 Fatigue Cycles The thermal transients from Reference
[5] were obtained from the thermal cycle diagrams in References 11 and 12, and the number of cycles to be considered for fatigue are listed in Reference
- 18. The number of cycles considered for each bounding transient is explained in Section 5.0. File No.: 1400735.302 Revision:
1 Page 5of17 F0306-0!R2 SJ Structural Integrity Associates, Inc.@: 4.2 Random Variables Random variables (RV) used in the N-702 evaluation are summarized in Table 3. Subsections
4.2.1 through
4.2.3 describe the more detailed input parameters used for SCC Initiation, *SCC Growth and fatigue crack growth respectively.
Table 3 identifies the specific references for each RV used in this N-702 evaluation.
4.2.1 SCC Initiation The cladding stress corrosion crack (SCC) initiation model in the VIPERNOZ program is a power law relationship.
Since there is no cladding specific SCC initiation data, the cast stainless steel SCC data in a BWR environment is used as specified in Reference 8, Section 8.2.2.2, and used in References 3 and 4. This model has the form; T = 84.2*10 18 O"-lo.s (1) where: T = time, hours cr = applied stress, ksi The residual plot shows that a lognormal distribution produces the best fit for the data. The lognormal residual plot with the linear fit of the data is shown below: where: 4.2.2 SCC Growth <l> = 0.9248x -0.003 = (x -cr) Iµ cr = data mean µ = data standard deviation X = In (T actuat/T predicted)
(2) The SCC growth model in VIPERNOZ program is also a power law relationship
[15]. The relationship used is; da = 6.82*10-12 K 4 dt where: da/dt = stress corrosion crack growth rate, in/hr K = sustained crack tip stress intensity
- factor, (3) File No.: 1400735.302 Page 6of17 Revision:
1 F0306-0IR2
{J Structural Integrity Associates, lnc.18. The residual plot shows that a Weibull distribution produces the best fit for the data. The Weibull residual plot with the linear fit of the data is shown below: Y = 0.9085x -0.3389 where: y = ln (In (1/ (1-F) )) F = cumulative distribution from 0 to 1 X = 1n ((da/dt) actual/ (da/dt) predicted)
4.2.3 Fatigue
Crack Growth (4) The fatigue crack growth data for SA-533 Grade B Class 1 and SA-508 Class 2 (carbon moly steels) in a reactor water environments are reported in Reference
[16] for weld metal testing at an R-ratio (algebraic ratio ofKminlKmax, "R") of 0.2 and 0.7. To produce a fatigue crack growth law and distribution for the VIPERNOZ software, the data for R= 0.7 was fitted into the form of Paris Law. The R= 0.7 fatigue crack growth law was chosen for conservatism.
The curve fit results of the mean fatigue crack growth law is presented with the Paris law shown as follows: where da = 3.817*10-9(Af()2.n1 dn a = crack depth, in n =cycles = Kmax -Kmin, ksi-in°*5 (5) A comparison to the ASME Section XI fatigue crack growth law in a reactor water environment is documented in Reference
[ 14] and it shows a reasonable comparison where the Section XI law is more conservative on growth rate at high AK. Using the rank ordered residual plot, it is shown that a Weibull distribution is representative for the data. The Weibull residual plot with the linear curve fit of the data is shown below: y = -0.3712 + 4.15x where y = ln(ln(l/(1-F))
X = ln((da/dn)actuail(da/dn)mean) F = cumulative probability distribution (6) File No.: 1400735.302 Revision:
1 Page 7of17 F0306-0IR2 SJ Structural Integrity Associates, Inc.@: 5.0 STRESS RESULTS AND FATIGUE CYCLE LOADINGS The stress analyses for the nozzle-to-shell weld and the nozzle blend radius for the Nl nozzle are presented in Reference
[ 5]. The stress analyses are performed for the load cases of unit pressure, and the bounding normal and upset (Service Levels A and B) thermal transient.
The azimuthal locations evaluated are 0° and 90°, which also represents the symmetric un-modeled 180° and 270° locations of the nozzle.
Two through-wall sections are selected at each azimuthal location.
One is at the location of the weld between the RPV and nozzle and the other is at the blend radius location of the nozzle. The load cases analyzed for the Nl nozzle include: 1. Unit pressure (1000 psi) 2. SCRAM Transient (Zone 10 to 11) [12] 3. Interruption of Feed Flow Nozzle Transient (Zone 9 to 10) [11] For the thermal transients, only the maximum or minimum through-wall linearized membrane plus bending stress profiles that produce the largest stress ranges for thermal fatigue crack growth are used in the evaluation.
These through wall stress profiles are shown in Figures 3 and 4. The nozzle specific transient Interruption of Feed Flow (Zone 9 to 10, 80 cycles, 11, 13, 18) is the bounding Service Level A/B transient because it has a 450°F temperature shock and returns to 530°F after only 60 seconds. Reference 11 states that this transient is only applicable to Dresden, but will be considered in this analysis for both plants to bound the analysis.
The transient occurs in 20 out of the 80 cycles of the Interruption of Feed Flow transient.
The second part of the transient is bounded by the first shock, but occurs 60 times during the 80 cycles of the Interruption of Feed Flow transient.
The Improper Start of Cold Recirc Loop (Zone 13 to 14, 11) nozzle transient (10 cycles, 12) is also bounded by the first Interruption of Feed Flow shock. Thus, 90 cycles of the third load case analyzed in Reference
[5] will be considered for the 40 year revised design basis cycles. This amount to 23 cycles for each block of 10 years of operation over 60-years of operation Due to the severity of the transient, the Vessel SCRAM transient (Zone 10 to 11, 12) is selected to bound all other transients.
The number of thermal cycles for the SCRAM transient is considered to be the total number of cycles for Service Level A/B conditions and normal startup/shutdown cycles [11, 12, 18]. Therefore the transients bounded were:
- Plant startup (298 cycles), o Includes Turbine roll with Feedwater Injection
- Plant cooldown (286 Cycles), o Includes Reduction of power File No.: 1400735.302 Revision:
1 Page 8of17 F0306-01R2 SJ Structural Integrity Associates, Inc.lg
- Loss ofFeedwater Flow (80 Cycles),
- SCRAM (294 Cycles), This totals:
- 958 cycles for the design 40 years of operation and
- 1437 cycles when scaled for 60 years of operation, or 240 cycles for each block of 10 years of operation over 60-years of operation.
Weld residual stresses (WRS) are assumed present in the nozzle-to-shell welds. The WRS distribution at the nozzle/shell weld is assumed to be a cosine distribution through the wall thickness with 8 ksi mean amplitude and 5 ksi standard deviation.
Figure 5 shows the assumed cosine distribution and the 3rd order polynomial fit used in the evaluation for Paths 2 and 4. No WRS is present in the nozzle blend radius region. 6.0 ASSUMPTIONS The following assumptions used in the evaluation are based on previous BWRVIP development projects.
Details of each assumption are provided.
[8]. 2. One stress corrosion crack initiation and 0.1 fabrication flaws is assumed per nozzle blend radius as justified in BWRVIP-108NP
[3] and BWRVIP-108 SER [2]. 3. One stress corrosion crack initiation and 1.0 fabrication flaw is assumed per nozzle/shell weld as justified in BWRVIP-108NP
[3]. 4. The NRC Pressure Vessel Research Users' Facility (PVRUF) flaw size distribution is assumed to apply as justified in the W-EPRl-180-302
[14] report. 5. The weld residual stress distribution at the nozzle/shell weld is assumed to be a cosine distribution through the wall thickness with 8 ksi mean amplitude and 5 ksi standard deviation as justified in BWRVIP-108NP
[3]. 6. Upper shelf fracture toughness is set to 200 ksivlin with a standard deviation of 0 ksivlin for irradiated material [2]. 7. Standard deviation of the mean Krc is set to 15 percent of the mean value of the Krc as justified in BWRVIP-108 SER [2]. 8. No CMTR is available for the nozzle-to-vessel weld in the RPV fabrication records [21], thus the copper/nickel content and initial RTndt values are taken from Reference
[3]. 9. No CMTR is available for the Dresden nozzles. Typically when material chemistries are not available, generic industry values are used from Reference
[3] and [4]. However, since the values available for Quad Cities bound the generic values, the Quad Cities values are considered bounding for this calculation.
- 10. No copper content is available in the nozzle forging CMTR [21], thus the nozzle forging copper content of 0.09189% from Reference
[4] is used. 11. Zero inspection coverage conservatively assumed for the initial 40 years of operation File No.: 1400735.302 Revision:
1 Page 9of17 F0306-0IR2 SJ Structural Integrity Associates, Inc.cg 7.0 RESULTS OF ANALYSES The reliability evaluation is presented using plant specific inspection coverage.
The probabilities of failure (PoF) per year due to the limiting LTOP event with 25% inspection for the extended operating term (with zero inspection coverage for the initial 40 years of operation) are summarized in Table 4. The PoF per year for the nozzle blend radius and the nozzle-to-shell weld due to LTOP events are both less than the 5x1 o-6 per year acceptance criterion from Reference
[ 19].
8.0 CONCLUSION
S The probability of failure per reactor year for the nozzle-to-shell-weld and nozzle blend radii in the Dresden/Quad Cities Nl nozzle is below the acceptance criterion of sx10-6 per year. This analysis shows that the Nl nozzles meet the acceptable failure probability even when considering elevated fluence level, thus qualifying all Dresden 2/3 and Quad Cities 1/2 RPV nozzles with full penetration welds (except feedwater and control rod drive return nozzles) for reduced inspection using ASME Code Case N-702 to the end of the period of extended operation.
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9.0 REFERENCES
- 1. Code Case N-702, "Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell Welds,Section XI, Division 1," Febmary 20, 2004. 2. Safety Evaluation of Proprietary EPRI Report, "BWR Vessel and Internal Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Inner Radius (BWRVIP-108)," December 19, 2007, SI File No. BWRVIP.108P.
- 3. BWRVIP-108NP:
BWR Vessel and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii. EPRI, Palo Alto, CA: 2007. 1016123. 4. BWRVIP-241:
BWR Vessel Internal Project, Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii, EPRI, Palo Alto, CA. 1021005NP.
- 5. SI Calculation 1400735.301, "Finite Element Model Development and Thermal/Mechanical Stress Analyses for the Nl Nozzle," Revision 0, January 2016. 6. General Electric Fluence Evaluation GE-NE-0000-0011-0531-R3-NP, "Dresden and Quad Cities Neutron Flux Evaluation," Rev. 3, SI File No. 1400735.213.
- 7. SI Calculation 1400735.303, "Verification of Software VIPERNOZ Version 2.0," Revision 0, October 2015. 8. BWRVIP Report, "BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP-05)," Electric Power Research Institute TR-105697, September 1995. EPRI PROPRIETARY INFORMATION.
- 9. VIPER, Vessel Inspection Program Evaluation for Reliability, Version 1.2 (1/5/98), Stmctural Integrity Associates.
- 10. Babcock & Wilcox RPV Certified Stress Report for Dresden 3, Contract No. 610-0111-51, Rev. 2, SI File No. 1400735.206.
- 11. General Electric Drawing No. 158B7279, Sheet 1, Revision 1, "Nozzle Thermal Cycles (Recirculation Outlet)," SI File No. 1400735.203.
- 12. General Electric Drawing No. 921D265, Sheet 1, Revision 1, "Reactor Thermal Cycles," SI File No. 1400735.209.
- 13. General Electric Report No. 26A5588, Revision 0, "Reactor Vessel -Power Uprate," SI File No. 1400735.201.
- 14. SI Calculation W-EPRI-180-302, "Evaluation of effect of inspection on the probability of failure for BWR Nozzle-to-Shell-Welds and Nozzle Blend Radii Region," Revision 0. 15. NUREG/CR-6923, Appendix B.8, "Expert Panel Report on Proactive Materials Degradation Assessment," Published Febmary 2007. File No.: 1400735.302 Revision:
1 Page 11of17 F0306-0IR2 SJ Structural Integrity Associates, Inc.ca 16. Bamford, W. H., "Application of corrosion fatigue crack growth rate data to integrity analyses of nuclear reactor vessels," Journal of Engineering Materials and Technology, Vol. 101, 1979, SI File No. 1300341.213.
- 17. ASME Boiler and Pressure Vessel Code,Section XI, Appendix G, 2013 Edition. 18. Exelon Transmittal of Design Input (TODI) No: QDC-13-049, Rev. 0, SI File No. 1301300.205.
- 19. Technical Basis for Revision of Pressurized Thermal Shock (PTS) Screening Limit in the PTS Rule (10 CFR 50.61), NUREG-1806, Vol. 1, August 2007. 20. USNRC Report, "Final Safety Evaluation of the BWR Vessel Internals Project BWRVIP-05 Report," TAC No. M93925, Division of Engineering Office of Nuclear Reactor Regulation, Nuclear Regulatory Commission, July 28, 1998. 21. RPV Fabrication Records a. Quad Cities I, SI File No. 1400735.212.
- b. Quad Cities II, SI File No. 1400735.211.
- 22. EPRI Letter 2012-138, "BWRVIP Support of ASME Code Case N-702 Inservice Inspection Relief," August 31, 2012, SI File No. 1300341.213.
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1 Page 12of17 F0306-0IR2 SJ Structural Integrity Associates, Inc.cg Table 1: Deterministic Parameter Summary VIPERNOZ Variable Value Reference RPV Thickness 6.125 inches (excluding clad) [10]
- RPVRadius 125.6875 inches (to vessel surface) [10]
- Clad Thickness 0.1875 inches [10]* Operating Temperature 530 °F(Region B) [11,12, 13] LTOP Event Temperature 100 °F [20] Operating Pressure 1005 psig [11,12, 13] LTOP Event Pressure 1200 psig [20] *Note: It was determined in Reference
[5] that the Nl nozzles of all four units under consideration are identical.
Table 2: Probability of Detection Distribution
[14] Flaw Size, in. Cumulative POD 0.00 0.00 0.05 0.10 0.10 0.46 0.15 0.80 0.20 0.92 0.25 0.95 0.30 0.98 0.35 0.99 0.40 0.99 0.45 1.00 0.50 1.00 0.55 1.00 0.60 1.00 File No.: 1400735.302 Revision:
1 Page 13of17 F0306-0IR2 e Structural Integrity Associates, Inc.cg Table 3: Random Variables Parameter Summary for Nl Nozzle Random Parameter Mean Std Dev Distribution Ref. Flaw density, nozzle/shell 1 per weld -vi Mean Poisson [3,3,3] weld (fabrication)
Flaw density, nozzle/shell 1 per weld -vf Mean Poisson [3,3,3] weld (SCC initiation)
Flaw density, nozzle blend 0.1 per weld -vi Mean Poisson [2,2,3] radius (fabrication)
Flaw size (fabrication) n/a n/a PVRUF [3] Flaw size (stress corrosion)
Clad thickness n/a Constant [3,3] Weld residual stress, 8 inside surface 5 Normal [3,3,3] through-wall (ksi) cosine distribution Clad residual stress (ksi)* 32 5 Normal [3,3,3] Nl Nozzle %Cu 0.26 0.045. Normal [3,3,3] %Ni 1.2 0.0165 Normal [3,3,3] to shell Initial RT ndt weld (oF) -20 13 Normal [3,3,3] %Cu 0.09189 0.04407 Normal [2,2,3] Nl Nozzle %Ni 0.70 0.068 Normal [21,2,3] forging Initial RTndt 40 26.48 Normal [21,2,3] (oF) Lower Shell fast neutron 5.59el 7 0.2 (20%) n/a [6,3] fluence (n/cm 2) Krc upper shelf (ksi-vlin) 200 0 Normal [2,22,3] Residual sec initiation time (hr) 't = 84.2xl0 18 ( a}10*5 y=0.9248x-Lognormal
[2,3,3] 0.0003 K dependent Residual da/dt = 6.82xl o-12 (K)4 y=0.9085x-Weibull [15,3,3] SCCG (in/hr) K >50 ksi-vlin 0.3389 K independent da/dt = 2.8x10-6 , Na na [15] K <50 ksi-vlin sec threshold (ksi-vlin) 10 2 Normal [2,3,3] Fatigue crack growth (FCG) da/dn=3.82 Residual (in/cycle) x 10-9( dK)2.921 y=4.155x-Weibull [3,3,3] 0.3712 FCG threshold (ksi-vlin) 0 0 Normal [3,3,3] *Note: The mean clad stress used already includes the effects of post-weld heat treatment.
File No.: 1400735.302 Revision:
1 Page 14 of17 F0306-0IR2 S) Structural Integrity Associates , Inc.<< Table 4: PoF for Period of Extended Operation PoF per year due to LTOP Event Location Zero inspection Allowable PoF per for initial 40 years , y ear [19] 25% forPEO* Path 1 7.8 x 10-s Path 2 3.33 x 1 0-11 5.0 x 10-6 Path 3 < 8.33 x 1 0-13 Path 4 < 8.33 x 10-1 3 *Note: Values include I x 10-3 probability of LTOP event occurrence per yea r [3 , pg 5-1 3]. Perry -Vipcmoz M::lde l -30 F il e No.: 1400735.302 R ev i s i o n: 1 F i g ure 1: Stress Ex t ract i o n P at h Ori e nt a ti o n s in t h e Nl Nozz l e Page 1 5 of 1 7 F0306-0 JR2 S) Structural Integrity Associates , lnc.c 55 50 -+-Path 1 45 Path2 :":" (fj _..,_ Path 3 40 -(fj 0 Path 4 (fj 35 Q) ... -30 CJ) C) r:: 25 *;; *.::: 20 c (.) 15 () 10 . 5 0 -r--....---.----,--...---.--,....-...,....--.----.----,--....----,---t 0.0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 10.0 11.0 12.0 13.0 Path length (in) Figure 2: Press u re Stress Distr i butions for the N-702 Eval u at i o n (fj -5 (fj 0 (fj -en C) -5 r:: "> *.::: c -10 (.) u +-Path 1 Path 2 _..,_ Path 3 Path 4 -20 +--.----.--.....--...---.--,....-...,....--.----.----,--....----,---t 0.0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 10.0 11.0 12.0 13.0 Path length (in) Figure 3: Fu ll Power Thermal Expansion Stress Distrib u tions for the N-702 Eva l uation Fi l e No.: 1400735.302 Revision:
1 Page 16 ofl 7 F0306-0 IR2 e Structural Integrity Associates , Inc.<< 70 60 -+-Path 1 Path 2 -50 "' ._._ Path 3 Path 4 .x: -40 "' "' 30 .... en en 20 c: > *c:: 10 c .x: 0 (.) cu ... (.) 20 --.-----.---,---..-----.--,--......---.-----.---,---.,....----,---r 0.0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 10.0 11.0 12.0 13.0 Path length (in) F igure 4: Bounding Transient Stress Distributions for the N-702 Eva luation 12 ti-Cosine Stress Distr Poly. (Cosine Stress Distr) *u; .x: 6 -y = -4E-14x 3 + 1.543x 2 -10.41 Sx + 11.631 en Cl) 4 QI ... .... ,,, 2 ca :I 0 :2 Cl) -2 QI -0:: ---6 --10 0.0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 Weld Thickness (in) Fig ur e 5: Weld Residual Stress Distributions for Paths 2 and 4 File No.: 1400735.302 Revi s ion: 1 Page 17of17 F0306-0IR2 SJ Structural Integrity Associates, Inc.CE Appendix A LIST OF SUPPORTING FILES File No.: 1400735.302 Revision:
1 Page A-1 of A-2 F0306-0IR2 i} Structural Integrity Associates, Inc.IE. File Name Pathl.INP Path3.INP Path2.INP Path4.INP Pathl.OUT Path3.0UT Path2.0UT Path4.0UT VIPERNOZ v2.EXE ISPCTPOD.EXE FL WDSTRB.EXE File No.: 1400735.302 Revision:
1 Description VIPERNOZ input file for Path 1 at nozzle blend radii. VIPERNOZ input file for Path 3 at nozzle blend radii. VIPERNOZ input file for Path 2 at nozzle-to-shell-weld.
VIPERNOZ input file for Path 4 at nozzle-to-shell-weld.
VIPERNOZ output file for Path 1 at nozzle blend radii. VIPERNOZ output file for Path 3 at nozzle blend radii. VIPERNOZ output file for Path 2 at nozzle-to-shell-weld.
VIPERNOZ output file for Path 4 at nozzle-to-shell-weld.
VIPERNOZ executable program VIPERNOZ probability of detection curve input file VIPERNOZ flaw size distribution curve input file Page A-2 of A-2 F0306-0IR2