NG-18-0036, Supplemental Information, Fifth Inservice Inspection Interval Program Plan, Relief Request RR-03
| ML18086A495 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 03/27/2018 |
| From: | Dean Curtland NextEra Energy Duane Arnold |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NG-18-0036 | |
| Download: ML18086A495 (26) | |
Text
NEXTera.
E N ERGY~
DUANE ARNOLD March 27, 2018 NG-18-0036 10 CFR 50.55a U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Duane Arnold Energy Center Docket No. 50-331 Renewed Op. License No. DPR-49 Supplemental Information, Fifth lnservice Inspection Interval Program Plan, Relief Request RR-03
References:
- 1) Letter. Curtland (NextEra) to U.S. NRC, "F)fth lnservice Inspection Interval Program Plan," dated March 7, 2017 (ML17069A172)
- 2) Letter. Curtland (NextEra) to U.S. NRC. "Response to Request for Additional Information, Fifth lnservice Inspection Interval Program Plan, Relief Request RR-03," dated October 26, 2017 (ML17300A195)
- 3) Letter, Curtland (NextEra) to U.S. NRC, "Response to Second Request for Additional Information, Fifth lnservice Inspection Interval Program Plan, Relief Request RR-03," dated January 26, 2018(ML18026A779).
In the Reference 1 letter, NextEra Energy Duane Arnold, LLC (hereafter NextEra Energy Duane Arnold) submitted our Fifth lnservice Inspection Interval Program Plan pursuant to 10 CFR 50.55a, which included Relief Request RR-03. In Reference 2, NextEra submitted additional information regarding Relief Request RR-03. Reference 2 included a Probabilistic Fracture Mechanics (PFM)
Evaluation, Calculation 1701150.301R1, performed by Structural Integrity Associates, Inc. (SI) . NextEra has since been notified by SI that the VIPERNOZ software version 2.0 used for the PFM calculations in the evaluation contained an error related to the behavior of Stress Corrosion Crack Growth. This error also NextEra En ergy Du ane Arnold , LLC , 3277 DAEC Ro ad, Palo, IA 52324
Document Control Desk NG-18-0036 Page 2 of 2 affected the calculation performed to determine the operating probabilities of failure submitted in Reference 3.
SI has corrected this error and rerun the analysis, which is enclosed with this letter. Probabilistic Fracture Mechanics (PFM) Evaluation, Calculation 1701150.301 R2, performed by SI, supersedes, in its entirety, the PFM Evaluation provided in Reference 2. The analysis, using the corrected software, indicates the probabilities of failure will increase slightly, but do not change the conclusions documented in Reference 3.
This letter does not contain any new or revised commitments.
If you have any questions or require additional information, please contact Michael Davis at 319-851-7032.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on March 27, 2018 Dean Curtland Site Director NextEra Energy Duane Arnold, LLC Enclosure cc: NRG Regional Administrator NRG Resident Inspector NRG Project Manager
Enclosure to NG-18-0036 Supplemental Information, Fifth lnservice Inspection Interval Program Plan, Relief Request RR-03 23 pages follow
~Structural Integrity Associates, Inc. File No.: 1701150.301 Project No.: 1701150 CALCULATION PACKAGE Quality Program Type: [8J Nuclear D Commercial PROJECT NAME:
Duane Arnold N702 Evaluation CONTRACT NO.:
02370494, Revision 0 CLIENT: PLANT:
NextEra Energy Duane Arnold Energy Center CALCULATION TITLE:
Probabilistic Fracture Mechanics Evaluation for Duane Arnold Recirculation Outlet (NI) Nozzles Project Manager Preparer(s) &
Document Affected Revision Description Approval Checker(s)
Revision Pages Signature & Date Signatures & Date 0 I - 21 Initial Issue Resuonsible Engineer:
A A-2 Wilson Wong Wilson Wong WW 9/27/17 WW 9/27/17 Resuonsible Verifier:
Stan Tang SST 9/27/17 1 1, 8, 10, 15, 18 Addressed additional Resuonsible Engineer:
client editorial comments Wilson Wong Wilson Wong WW 9129117 WW 9/29/17 Resuonsible Verifier:
Jim Wu JW 9129117 Resuonsible Engineer:
vJ~vJ~ w~w~
2 11, 14, 16 Updated calculation A-2 results in Table 2&5 based on CAR-18-012 Wilson Wong Wilson Wong 3120118 3/20/18 Resuonsible Verifier:
- ~Atr-Stan Tang 3/20/18
e Structural Integrity Associates, Inc.@;.
Table of Contents
1.0 INTRODUCTION
....... ... ..... .. ........ ................ ................................................................ 4 2.0 OBJECTIVE .................................... .................................................. .. .. .. .................. .... 4 3.0 METHODOLOGY ................................................ .... ................ .................................... 4 3.1 Fatigue Cycles ................................... ..... .. ....... ..... .. ........ .... ............................... 5 3 .2 Probabilistic Fracture Mechanics Evaluation .......................... .......... ............. ... 5 4.0 DESIGN INPUTS ......... ............... .. ... ... .......... .............. .................................................. 6 4.1 Deterministic Para1neters ...... .............................. .. ............................ ... ..... ...... ... 6 4.1.1 In-Service Inspection ..... .. ....... .. ........ .... ............ ... .. .... .... ........... ... .. .. .... ......... .. ... 6 4.1. 2 Stresses .. .. .. ........... ... .. ......... ........... .. ..... .. .. ... ......... .... ..... .......... .... ....... ...... ... .... .. 6 4.2 Random Variables ...................................... .. ...... .. ............................................. 6 4.2.1 Material Chen1istry .......... .. .. .. .. .... .............................. .... .. ......... ..................... .. .. 7 4.2.2 Fluence . ....... .. .............. ... ..... .... ........ .. ........ .................................. .............. ..... ... 7 4.2.3 SCC Initiation ..... ..... ........ ... ... ..... ... ... ... .. ... .... ........... .. .. ..... .. ... .. .... ............ ......... . 7 4.2.4 SCC Growth ........ .... ...... .... ... .. ........... .. ... .. ... ...... ............... ............... .. ............... .. 7 4.2.5 Fatigue CrackGrov.1th ... .... .. .. .. ....... ........ ............. .. ......... ...... ...... ............ .. .. .... ... 8 5.0 STRESS RESULTS AND FATIGUE CYCLE LOADINGS ................................ ....... 9 6.0 ASSUMPTIONS ......................................................................... ................................... 9 7.0 RESULTS OF ANALYSIS ........... ....... ........................................................................ . 9
8.0 CONCLUSION
S .............. .. ......................................................................... ........... ..... 10
9.0 REFERENCES
............................... .. ............. .... ......... ... .... ................................... ....... 11 Appendix A LIST OF SUPPORTING FILES ...................................................................... A-1 File No.: 1701150.301 Page 2of21 Revision : 2 F0306-01R3
e Structural Integrity Associates, Inc.rs; List of Tables Table 1: Lumping of Thermal Cycles .......................... ........................................................... 13 Table 2: Random Variables Parameter Summary ................................................................... 14 Table 3 Deterministic Parameter Summary ............................................................................ 15 Table 4: Probability of Detection (PoD) Distribution [1 l] ...................................................... 15 Table 5: Duane Arnold PoF for Period of Extended Operation ............................................. 16 List of Figures Figure 1: Stress Extraction Path Orientations in the Nl Nozzle ...... ................ .... ................ .. 17 Figure 2: Unit Pressure Stress Distributions ........................................................................... 18 Figure 3: Through-wall Stress Distributions, Steady State Operating Conditions ................. 19 Figure 4: Through-wall Stress Distribution, Loss of Feedwater Pumps Transient ................ 20 Figure 5: Weld Residual Stress Distributions for Path 2 ........................................................ 21 File No.: 1701150.301 Page 3 of21 Revision: 2 F0306-01R3
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1.0 INTRODUCTION
In response to the U.S. Nuclear Regulatory Commission (NRC) request for additional information (RAI) on Duane Arnold's relief request (MLl 7069Al 72 dated March 7, 2017) based on American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code Case (CC) N-702 [I], this calculation addresses Sections A.3 and A.4 of BWRVIP-241 Appendix A[4]. The Code Case allows reduction of in-service inspection from 100% to 25% of all nozzle blend radii and nozzle-to-shell welds every 10 years, including one nozzle from each system and pipe size, except for feedwater (FW) and control rod drive (CRD) 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. To extend the applicability of Code Case N-702, a probabilistic fracture mechanics (PPM) evaluation, consistent with the methods ofBWRVIP-108 and BWRVIP-241, is performed to ensure that the probability of failure (PoF) remains acceptable. The NI (recirculation outlet) nozzles are identified as the bounding nozzles when fluence is not considered in BWRVIP-241 [4].
The evaluation consists of two parts: Finite Element Model (FEM) Stress Analysis and Probabilistic Fracture Mechanics (PPM) Analysis. The FEM stress analysis was previously performed for a separate DAEC project calculation [5], but new path stress results at the locations of interest have been generated.
This calculation package documents the new path stress results as well as the PPM analysis.
2.0 OBJECTIVE The objective of the evaluation documented in this calculation package is to perform a plant specific analysis of the DAEC recirculation outlet (NI) nozzles to extend applicability of the existing reliefrequest to 60 years of operation.
3.0 METHODOLOGY This evaluation considers the nozzle-to- shell weld and nozzle blend radius on the Duane Arnold NI nozzles 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. The probability of failure expressed in this calculation is equivalent to the through-wall cracking frequency.
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 5xl o-6 when changing from essentially 100% in-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 a postulated LTOP event (including 1x 10-3 probability ofLTOP event occurrence per year [3, pg. 5-13]) is less than 5x10-6, then the inspection reduction based on Code Case 702 can be extended to 60 years.
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3.1 Fatigue Cycles The thermal transients and number of cycles were obtained from Reference [6]. The number of cycles used in the PFM evaluation for the 10-year inspection interval is shown in Table 1. Transients were grouped into four groups based on severity, with the transients listed in bold font being the bounding transient in each group.
3.2 Probabilistic Fracture Mechanics Evaluation The probabilistic evaluation is performed for the case of 25% 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 from Reference [14] 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. The probability of failure for a circumferential crack is less than an axial crack, due to the difference in the stress (axial versus hoop) and the influence on the crack model. However, this probabilistic fracture mechanics evaluation for the nozzle to vessel shell weld considers both circumferential and axial cracks (depending on weld orientation).
An axial elliptical crack model with a crack aspect ratio of a/l = 0.5 is used in the evaluation for the nozzle-to-shell weld. The inspection probability of detection (PoD) curve from BWRVIP-05 [9, Table 4]
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 [9]. A Monte Carlo simulation is performed using a variant of the VIPER program [10]. The Monte Carlo method can be used to solve probabilistic problems using deterministic computation. A mean value, standard deviation, and distribution curve as defined in the random variables summary (Table 2) defines a set of possible inputs and their probabilities of occurring. Using this domain of possible inputs, a new set of inputs for each simulation are generated for use in determining whether the nozzle will fail using conventional deterministic fracture mechanics methodology. This is repeated 2 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.
The VIPER program was developed as pait of the BWRVIP-05 effo1t 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 [9] and [3].
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SJ Structural Integrity Associates, Inc.rs; 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 [8] to ensure the modifications made to the VIPER software are fully quality assured.
4.0 DESIGN INPUTS The plant specific input is described below. Section 4.1 presents the 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 3 summarizes the dimensional and operational inputs used in the N-702 evaluation [7, 6].
Subsections 4.1.1 through 4.1.2 describe the more detailed input parameters used for in service inspection (ISI) interval, stress distributions and fatigue cycles, respectively.
4.1.1 In-Service Inspection ln 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 4 [11].
4.1.2 Stresses Stresses due to vessel pressure and thermal transients were determined in a previous calculation [5]. From the results of that calculation package, new through wall stress distributions are generated for use in the N-702 evaluation. Figure I shows the locations and orientations of these new through-wall stress paths.
For vessel pressure, an internal pressure of l,000 psig was applied to the inside surfaces of the RPV and NI nozzle FEA model. Per Reference [6], the axisymmetric model has the effect of modeling the cylindrical RPV as spherical, which requires a correction factor to account for the 3-D effects of two intersecting cylinders. As calculated in Reference [6], the calculated pressure stress for the DAEC axisymmetric model needs to be multiplied by 2.572/1.5 = 1.715.
The maximum cyclic stress ranges for the four considered transients, based on a linearized through wall stress distribution on the new paths, are identified. Figure 2 through Figure 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 SCRAM transient, respectively. Details of the analysis can be found in [5].
4.2 Random Variables Random variables (RV) used in the N-702 evaluation are summarized in Table 2. Subsections 4.2.l through 4.2.5 describe the more detailed input parameters used for SCC Initiation, SCC Growth and fatigue crack growth respectively.
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4.2.1 Material Chemistry The weld chemistries for the Nl nozzles and nozzle to vessel welds are presented in Table 2 along with the standard deviation and distribution used in the PFM evaluation.
4.2.2 Fluence.
The bounding nozzle fluence for extended operation is listed in Reference [18]. The Nl6 nozzle has the highest fluence for both the nozzle blend and nozzle to shell weld. Therefore, the Nl 6 fluence values shown in Table 2 are used for analysis.
4.2.3 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 9, Section 8.2.2.2, and used in References 3 and 4.
This model has the form; (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:
<l> = 0.9248x - 0.0003 (2) where: <D = (x - cr) Iµ cr = data mean
µ = data standard deviation X = In (Tactua1/Tpredicted) 4.2.4 SCC Growth The SCC growth model in VIPERNOZ program is also a power law relationship [12]. The relationship used is; da = 6.82*10-12 (K) 4 (3) dt where: da/dt = stress corrosion crack growth rate, in/hr File No.: 1701150.301 Page 7 of21 Revision: 2 F0306-01R3
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K = sustained crack tip stress intensity factor, ksi\iin 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 (4) where: y =In (In (1/ (1-F) ))
F = cumulative distribution from 0 to 1 X =In ((da/dt) actual I (da/dt) predicted) 4.2.5 Fatigue Crack Growth The fatigue crack growth data for SA-533 Grade B Class 1 and SA-508 Class 2 (carbon-molybdenum steels) in a reactor water environments are rep011ed in Reference [13] for weld metal testing at an R-ratio (algebraic ratio of Kmin/Kmax, "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:
da = 3.817*10-9 (M)2.921 (5) dn where a = crack depth, in n =cycles LiK = Kmax - Kmin, ksi-in°* 5 A comparison to the ASME Section XI fatigue crack growth law in a reactor water environment is documented in Reference [11] and it shows a reasonable comparison where the Section XI law is more conservative on growth rate at high LiK.
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 (6) where y = ln(ln(l/(1-F))
x = ln((da/dn)actuatl(da/dn)mean)
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F = cumulative probability distribution 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 were performed in Reference [5] . Two through-wall paths are selected, one at the location of the weld between the RPV and nozzle and the other at the blend radius location of the nozzle, as shown in Figure 1.
For the thermal transients considered, 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. Example minimum and maximum through wall stress distributions for the Loss of Feedwater Pumps transient is shown in Figure 4.
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 Path2 . 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.
- 1. Flaws are assumed to be aligned parallel with the weld direction as justified in BWRVIP-05 [9].
- 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 one 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-EPRI-180-302 [11] repo1i.
- 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 ksiv'in with a standard deviation of 0 ksiv'in for un-irradiated material consistent with BWRVIP-108 [2].
- 7. Standard deviation of the mean Kie is set to 15 percent of the mean value of the Kie as justified in BWRVIP-108 SER [2].
- 8. Zero inspection coverage conservatively assumed for the initial 40 years of operation.
7.0 RESULTS OF ANALYSIS 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 File No.: 1701150.301 Page 9of21 Revision: 2 F0306-0IR3
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(with zero inspection coverage for the initial 40 years of operation) are summarized in Table 5. The PoF per year for the nozzle blend radius and the nozzle-to-shell weld due to LTOP events are both less than the Sx 10-6 per yearNRC safety goal from Reference [15].
8.0 CONCLUSION
S The probability of failure per reactor year for the nozzle-to-shell-weld and nozzle blend radii in the Duane Arnold NI nozzles are below the acceptance criterion of 5 x I o- 6 per year. Since this analysis considered the bounding nozzle geometry (NI), additional thermal cycles for the period of extended operation, and the bounding nozzle fluence (N 16) at the end of the period of extended operation, the NI nozzles and all other applicable nozzles are qualified for reduced inspection using ASME Code Case N-702 through 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," February 20, 2004.
- 2. Safety Evaluation of Proprietary EPRI Report, "BWR Vessel and Internal Project, Technical Basis for the Reduction oflnspection 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, ProbabWstic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Rad;;, EPRI, Palo Alto, CA.
1021005. EPRI PROPRIETARY INFORMATION.
- 5. SI Calculation DAEC-20Q-303, "Stress Analysis of Reactor Recirculation Outlet Nozzle," Revision 3.
- 6. SI Calculation DAEC-20Q-302, "Design Inputs and Methodology for ASME Code Fatigue Usage Analysis of Reactor Recirculation Outlet Nozzle," Revision 2.
- 7. Duane Arnold Drawing No. APED-Bl 1-2655-95-5, (Chicago Bridge & Iron Company, Contract No.
68-2967, Drawing No. 30, Revision 7) "Recirculation Outlet Nozzle Mk NIA," SI File No. DAEC-20Q-217.
- 8. SI Calculation 1701150.302, "Verification of Software VIPERNOZ Version 1.1," Revision 1.
- 9. BWRVIP Report, "BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP-05)," Electric Power Research Institute TR-105697, September 1995. EPRI PROPRIETARY INFORMATION.
- 10. VIPER, Vessel Inspection Program Evaluation for Reliability, Version 1.2 (1/5/98), Structural Integrity Associates.
- 11. 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.
- 12. NUREG/CR-6923, Appendix B.8, "Expert Panel Report on Proactive Materials Degradation Assessment," Published February 2007.
- 13. 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.
- 14. Delvin, S. A., Riccardella, P. C., 'Fracture mechanics analysis of JAERI model pressure vessel test,'
ASME PVP Conference, Paper 78-PVP-91, 1978.
- 15. 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.
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TAC No. M93925, Division of Engineering Office of Nuclear Reactor Regulation, Nuclear Regulatory Commission, July 28, 1998.
- 17. EPRl Letter 2012-138, "BWRVIP Support of ASME Code Case N-702 Inservice Inspection Relief,"
August 31, 2012, SI File No. 1300341.213.
- 18. Design Input Request Response, Revision 0, SI File No. 1701150.202.
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Table 1: Lumping of Thermal Cycles 60 Year Internal Total Cycles Number Transient Name Cycles Cycles Considered 3 Normal Startup 212 1 424 21-23 Shutdown 212 1 4 Turbine Roll and Increase to Rated Power 176 1 9 Lossof Feedwater Heaters Turbine Trip 25% Power 6 2 12 Turbine Generator Trip 30 1 329 13 Reactor Overpressure 1 1 15 SCRAM Other 110 1 17 Improper Startup 5 1 5 11 Loss of Feedwater Pumps 8 3 26 14 SRV Blowdown 2 1
- Note: Transients listed in bold are the bounding transients of its group.
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Table 2: Random Variables Parameter Summary Random Parameter Mean Std Dev Distribution Ref.
Flaw density, nozzle/shell 1 per weld --JMean Poisson [3,3,3]
weld (fabrication)
Flaw density, nozzle and nozzle/shell weld (SCC 1 per weld --JMean Poisson [3,3,3]
initiation)
Flaw density, nozzle blend 0.1 per nozzle --JMean Poisson [2,2,3]
radius (fabrication)
Flaw size (fabrication) n/a n/a PVRUF [31 Flaw size (stress corrosion) Clad thickness n/a Constant [3,3]
8 Weld residual stress, tlu*ough-inside surface 5 Normal [3,3,3]
wall (ksi) cosine distribution Clad residual stress (ksi)* 32 5 Normal [3,3,3]
% Cu 0.03 0.045 Normal [19,3 ,3]
Nl nozzle to %Ni 0.91 0.0165 Normal ri 9,3,31 shell weld Initial RTndt 10 13 Normal [19,3,3]
(oF)
% Cu 0.09189 0.04407 Normal rI9,2,3l Nl nozzle %Ni 0.93 0.068 Normal r19,2,31 forging Initial RTndt 40 26.48 Normal [19,2,3]
(of)
Nozzle forging fast neutron 2.0lel8 0.2 (20%) n/a [ 18,3]
fluence at OT (n/cm 2)
Nozzle to shell weld fast 3.27e18 0.2 (20%) n/a [18,3]
neutron fluence at OT (n/cm 2)
Kie upper shelf (ksi--Jin) 200 0 Normal [2,17,3]
Residual sec initiation time (Ju-) 't = 84.2xl0 18(oJ 10*5 Lognormal [2,3,3]
v=0.9248x-0.0003 K dependent Residual da/dt = 6.82x10- 12(K) 4 Weibull [12,3,3]
y=0.9085x-0.33 89 K >50 ksi--Jin SCCG (in/Ju')
K independent da/dt = 2.8x10-6 , na na [12]
K <50 ksi--Jin sec tlu*eshold (ksi--Jin) 10 2 Normal [2,3,3]
Fatigue crack growth (FCG) da/dn=3.817 Residual xi o-9c llK)2.921 Weibull [3,3,3]
(in/cycle) y=-0.3712+4. l 55x FCG tlu-eshold (ksi--Jin) 0 0 Normal [3,3,3]
- Note: The mean clad stress used already includes the effects of post-weld heat treatment.
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Table 3 Deterministic Parameter Summary VIPERNOZ Variable Value Reference RPV Thickness 5.25 inches (excluding clad) [7]
RPVRadius 92.5 inches (to vessel surface) [7]
Clad Thickness 0.1875 inches [7]
Operating Temperature 533 °F (Region B) [6]
LTOP Event Temperature 100 °F [16]
Operating Pressure 1025 psig [6]
LTOP Event Pressure 1200 psig [16]
Table 4: Probability of Detection (PoD) Distribution [11]
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.: 1701150.301 Page 15of21 Revision: 2 F0306-01R3
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Table 5: Duane Arnold PoF for Period of Extended Operation Zero inspection for initial 40 years, 25% fo.r PEO Nl Nozzle Location PoF per year due PoF per year due to Allowable PoF per to LTOP Event
- Normal Operation year r15]
Blend Radius i.02 x 10-1 2.50 x 10-8 5.0 x 10-6 Nozzle to Shell Weld < 8.33 x 10-9 < 8.33 x 10-9
- Note: Values include I x 10-3 probability ofLTOP event occurrence per year [3, pg 5-13].
File No.: 1701150.301 Page 16of21 Revision: 2 F0306-0IR3
SJ Structural Integrity Associates, Inc.fit l EUM'NTS MllT NUM PA'IH Duane Arnold Recirc. Outlet Nozzle, Transient 4 Stress Figure 1: Stress Extraction Path Orientations in the Nl Nozzle File No.: 1701150.301 Page 17of21 Revision: 2 F0306-01R3
e Structural Integrity Associates, Inc.is:
45 40 - P1
-"' 35
~
30 -
- P2 Cl>
a..
Cl) 25 C'l c:
- 20
'i::
c
~
(.) 15 C'a a..
u 10 5
0 0.0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 Path length (in)
Figure 2: Unit Pressure Stress Distributions File No.: 1701150.301 Page 18of21 Revision: 2 F0306-01R3
SJ Structural Integrity Associates, Inc.'"'
5 - -
~
(1) i...
en Cl c:
-5 - -
'i:: -10 c
~
CJ - P1
<<I i... -15 -
()
~--- - - -- - - ------- ~-- - *- - - - P2 -
-25 +-~~~~~-.-~~~~~---.~~~~~~....-~~~~~-.-~~--f 0.0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 Path length (in)
Figure 3: Through-wall Stress Distributions, Steady State Operating Conditions File No.: 1701150.301 Page 19 of21 Revision: 2 F0306-01R3
SJ Structural Integrity Associates, Inc.~
30 25 20
- 15
(/)
.lll::
10
( /)
(/)
5
...<LI en 0 Cl c: -5
- s;
'i:: -10 c
.lll:: -15 - P1 MIN CJ ctl
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-20
- P2MIN
-25
~ P1 MAX
-30
-35 - - - - - - - ~ P2MAX
-40 0.0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 Path length (in)
Figure 4: Through-wall Stress Distribution, Loss of Feedwater Pumps Transient File No.: 1701150.301 Page 20 of21 Revision: 2 F0306-01R3
e Structural Integrity Associates, lnc.oc 12
~ Cosine Stress Distr 10
- 8
- Poly. (Cosine Stress Distr)
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~ E-15x3 + 2.5028x2 - 13.265x + 11.631 6
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- s 0 "C
Q) c:::
-2
-4
-6
-8
-10 0.0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 Weld Thickness (in)
Figure 5: Weld Residual Stress Distributions for Path 2 File No.: 1701150.301 Page 21 of21 Revision: 2 F0306-01R3
Appendix A LIST OF SUPPORTING FILES File No.: 1701150.301 Page A-1 of A-2 Revision: 2 F0306-01R3
File Name Description Pathl.INP VIPERNOZ input file for Nl Path 1 at nozzle blend radii.
Path2.INP VIPERNOZ input file for Nl Path 2 at nozzle to shell weld Pathl.OUT VIPERNOZ output file for Nl Path 1 at nozzle blend radii.
Path2.0UT VIPERNOZ output file for Nl Path 2 at nozzle to shell weld VIPERNOZlpl.EXE VIPERNOZ executable program ISPCTPOD.EXE VIPERNOZ probability of detection curve input file FLWDSTRB.EXE VIPERNOZ flaw size distribution curve input file DAEC Stresses.xlsx Stress post-processing spreadsheet File No.: 1701150.301 Page A-2 of A-2 Revision: 2 F0306-01R3