ML12230A098
| ML12230A098 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 08/07/2012 |
| From: | Wu J Structural Integrity Associates |
| To: | Division of Operating Reactor Licensing |
| Tam P | |
| Shared Package | |
| ML12230A095 | List: |
| References | |
| TAC ME9160 1101463.301, Rev 2 | |
| Download: ML12230A098 (14) | |
Text
Accession No. ML12230A098 File No.: 1101463.301 Project No.: 1101463 CALCULATION PACKAGE Quality Program: Nuclear Commercial PROJECT NAME:
Monticello N2 Nozzle Code Case N-702 Relief Request CONTRACT NO.:
00001005 Release 0032 CLIENT: PLANT:
XCEL Energy Monticello Nuclear Generating Plant CALCULATION TITLE:
Evaluation of effect of inspection on the probability of failure for BWR Recirculation Inlet (N2) nozzle-to-shell-welds and nozzle blend radii region at Monticello Project Manager Preparer(s) &
Document Affected Revision Description Approval Checker(s)
Revision Pages Signature & Date Signatures & Date 0 1 - 10 Initial Issue Jim Wu Jim Wu A A-2 2/2/12 2/2/12 Stan Tang 2/2/12 1 1 - 11 Add References 20 and Jim Wu Jim Wu A A-2 Updated Item 7 of the 6/27/12 6/27/12 Assumption Section Stan Tang 6/27/12 2 1 - 11 Add Responses to NRC A A-2 RAI Jim Wu Jim Wu JW JW 8/07/12 8/07/12 Stan Tang SST 8/07/12 Page 1 of 11 F0306-01R1
Table of Contents 1.0 OBJECTIVE .................................................................................................................. 3 2.0 METHODOLOGY ........................................................................................................ 3 3.0 DESIGN INPUT ............................................................................................................ 3 4.0 ASSUMPTIONS............................................................................................................ 4 5.0 SOFTWARE MODIFICATIONS ................................................................................. 4 6.0 FATIGUE CRACK GROWTH ..................................................................................... 5 7.0 STRESS RESULTS AND FATIGUE CYCLE LOADINGS ....................................... 5 8.0 PROBABILISTIC FRACTURE MECHANICS EVALUATION ................................ 6 9.0 RESULTS OF ANALYSES/CONCLUSIONS ............................................................. 7 9.1 Nozzle Blend Radii ............................................................................................ 8 9.2 Nozzle-to-Shell Weld ........................................................................................ 8
10.0 CONCLUSION
S ........................................................................................................... 8
11.0 REFERENCES
.............................................................................................................. 8 APPENDIX A COMPUTER FILES .................................................................................... A-1 List of Tables Table 1: Monticello Weld Chemistry ..................................................................................... 10 Table 2: Probability of Failure Results Summary for Nozzle Blend Radii ............................ 10 Table 3: Probability of Failure Results Summary for Nozzle-to-Shell Weld ......................... 11 File No.: 1101463.301 Page 2 of 11 Revision: 2 F0306-01R1
1.0 OBJECTIVE This evaluation is to justify the reduction of in-service inspection of the nozzle-to-shell-weld and the nozzle blend radii in the recirculation nozzle (N2) at Monticello Nuclear Generating Plant per code case N-702 for the extended period of operation. The N-702 code case, with appropriate technical justification, could be used as an alternative to the requirements of ASME Section XI, Examination Category B-D.
2.0 METHODOLOGY The approach was based on the methodology presented in Reference 1. A Monte Carlo simulation was performed using a variance of the program, VIPER [2] with some modifications as described in the following sections. The VIPER program was developed as part of the program in Reference 1 for the BWR reactor pressure vessel (RPV) shell weld inspection recommendations. The software was modified into a separate edition, identified as VIPERNOZ, for use in this evaluation.
The detailed description of the methodology incorporated in the VIPER/VIPERNZ program is documented in References 1 and 11.
3.0 DESIGN INPUT This analysis is intended for evaluating the reduction of inspection based on the probability of failure in the nozzle-to-shell-weld and nozzle blend radius in N2 nozzles at Monticello Nuclear Generating Plant.
Some of the input is based on the prior analyses on BWR fleet per References 3 and 4. Others were from Monticello plant specific described below.
Vessel Wall Thickness = 5.0625 [5]
Vessel Radius = 103.188 [5]
Vessel Clad Thickness = 0.125 [5]
Vessel Operating Temperature = 549°F [10, Page 91]
Operating Pressure = 1025 psig [10, Page 91]
Radius to Nozzle-to-shell Weld = 18.25 [6, Figure 1], [17]
End of Life Fluence (54 EFPY/60 years) for N2 nozzles = 1.01x1018 n/cm2 [7]
The weld chemistries are presented in Table 1.
For the nozzle blend radius region, since the nozzle is a forging, the number of fabrication flaws was assumed to be 0.1 flaws per nozzle. In the weld between the vessel shell and the nozzle, the number of fabrication flaws was assumed to be 1 per nozzle-to-shell-weld.
File No.: 1101463.301 Page 3 of 11 Revision: 2 F0306-01R1
All random variables were summarized in Table 2 of Reference 8. Most of the input is obtained from Reference 1, except standard deviation for %Cu and %Ni for nozzle blend radii. They are 0.0447 and 0.068 for %Cu and %Ni, respectively and were obtained from BWRVIP-173 [18].
4.0 ASSUMPTIONS The following assumptions are used in the evaluation based on References 8 and 20:
(1) Fabrication flaw is assumed only due to the weld process in the nozzle-to-shell-weld (2) One stress corrosion initiated flaw and 0.1 fabrication flaw per nozzle blend radius (3) One fabrication flaw and one stress corrosion initiated flaw per nozzle/shell weld (4) Flaw size distribution, PVRUF, is assumed.
(5) Residual stress at the nozzle/shell weld is assumed cosine distribution through the wall thickness with a mean of 8 ksi at surface.
(6) The standard deviation for surface residual stress is assumed to be 5 ksi.
(7) Lower bound constant upper shelf fracture toughness is set to 200 ksiin per Reference
- 20. This is considered conservative, since the 2004 Edition of the ASME Code,Section XI, Appendix A specifies the maximum KIC to be 220 ksiin.
(8) Standard deviation of the mean KIC is set to 10 percent of the mean value of the KIC per Reference 20. This is consistent with the current PTS rule as stated in Reference 20.
5.0 SOFTWARE MODIFICATIONS Several modifications were made to VIPER in order to include the capability to perform the evaluation for nozzle bend radii. The modifications were:
(1) Include fatigue crack growth analysis (2) Option to perform stress corrosion crack growth and/or fatigue crack growth (3) User defined flaw size distribution (4) User defined probability of detection (PoD) curves for inspection.
(5) User defined event occurrence time (6) User defined distribution for selected random parameters (7) User input number of printout for failed and non-failed vessels.
(8) The constant for margin term for upper bound values of adjusted reference temperature required by Appendix G to 10 CFR Part 50 is a user input.
(9) Preservice inspection is eliminated.
(10) Initial flaw size to include clad thickness is a user option.
(11) Improvement in data structure for analysis results.
The modified software for this project is identified as VIPERNOZ to distinguish from the original VIPER software in Reference 1.
File No.: 1101463.301 Page 4 of 11 Revision: 2 F0306-01R1
6.0 FATIGUE CRACK GROWTH The fatigue data for A533-B-1 and A508-2 in reactor water environment are reported in Reference 12 for weld metal testing at R = 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 a form of Paris Law. The R= 0.7 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 (K ) 2.927 (1) dn where a = crack depth n = cycle K = Kmax - Kmin A comparison to the ASME Section XI [4] fatigue crack growth law in reactor water environment was done in Reference 8, it shows a very reasonable comparison where Section XI is more conservative on growth rate at high K.
Using the rank ordered residual plot, it was shown that a Weibull distribution was more representative for the data. The Weibull residual plot with the linear curve fit of the data is shown below:
y = -0.3712 + 4.15x (2) where y = ln(ln(1/(1-F))
x = ln((da/dn)actual/(da/dn)mean)
F = cumulative probability distribution 7.0 STRESS RESULTS AND FATIGUE CYCLE LOADINGS The stress analyses for the nozzle/shell weld and the nozzle blend radius for the N2 nozzles were presented in Reference 6. The stress analyses were performed for the load cases of unit pressure, and the relevant thermal transients for the N2 nozzles. The through-wall sections were selected based on the thermal transient results. The azimuth locations evaluated were 0, 90, 180 and 270 of the nozzles.
Two through-wall sections were selected. Section C is at the location of the weld between the RPV and nozzle. Section D is at the blend radius location of the nozzle.
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The load cases analyzed for the N2 nozzles include:
(1) Unit pressure (2) Unit axial load (3) Unit in-plane moment (4) Unit out-of-plane moment (5) Thermal transients depending on the nozzles as described in the following sections For the selected sections in the N2 nozzles (nozzle blend radius and nozzle-to-vessel shell weld), stresses due to the nozzle axial and moment loads are small compared to the pressure and thermal loadings.
Therefore, these load cases were not used in the evaluation.
The thermal transients for the recirculation inlet nozzle are the heat up and sudden pump start of cold recirculation loop. The pressure is maintained at 1050 psig for the sudden pump start transient.
For the thermal transients, only the maximum or minimum through-wall stress profiles that produce the largest stress ranges for thermal fatigue crack growth are presented and used in the evaluation. The maximum stress among the four azimuth locations was used.
In this section, the maximum stress is at the 90 and 270 in the hoop direction for the combination of pressure and thermal stresses.
The thermal cycles for recirculation inlet nozzle are the number of heat/shutdown cycles (288 cycles Reference 16 for an end of operation time of 60 years), and the number of sudden pump start of cold recirculation loop cycles (10 cycles per page 12 of Reference 13 for an end of operation time of 60 years).
8.0 PROBABILISTIC FRACTURE MECHANICS EVALUATION The probabilistic evaluation was performed for the case of 25% inspection for the extended operating term (70% inspection coverage for the initial 40 years of operation at the nozzle blend radii and 47%
inspection coverage for the initial 40 years of operation at the nozzle-to-shell weld location per Reference 19).
For the nozzle blend radius region, a nozzle blend radius crack model, [14] was used in the probabilistic fracture mechanics evaluation for the reliability of the in-service inspection program. For this location and crack model, the applicable stress is the stress perpendicular to any path cut along the nozzle longitudinal axis. Therefore, the maximum stress among the four azimuth locations (0, 90, 180 and 270) was selected.
For the nozzle-to-vessel shell weld, either a circumferential or an axial crack could be initiated due to either component fabrication (i.e. considering only welding process) or stress corrosion cracking. From File No.: 1101463.301 Page 6 of 11 Revision: 2 F0306-01R1
Reference 1, it is shown that the probability of failure for a circumferential crack is much less than an axial crack, due to the difference in the stress (hoop versus axial) and the influence function of the crack model. It is also shown in the through-wall stress plots in Reference 3 that the difference between the thermal hoop stress and the thermal axial stress is not as much compared to the difference between the pressure hoop stress and the pressure axial stress. Therefore, the probabilistic fracture mechanics evaluation for the nozzle and vessel shell weld would concentrate on the axial crack.
For the nozzle-to-vessel shell weld, the following crack model was used in the evaluation:
(a) Axial elliptical crack model with a crack aspect ratio of a/l = 0.2 The inspection PoD curve is the user input of Figure 42 of Reference 8, with an inspection interval every 10 years.
The crack size distribution, PVRUF, is shown in Figure 43 of Reference 8.
The calculation of stress intensity factor is at the deepest point of the crack.
The probability of failure was obtained due to a low temperature over pressurization (LTOP) event at
-3 88 F and 1150 psi, [15]. The probability of the LTOP event is 1x10 per year [15].
The probability of failure was also presented for plant normal operation.
The analyses were performed using VIPERNOZ, a superset of the program VIPER, [2], with the modifications as described in Section 5.
The number of simulations was 1 million.
9.0 RESULTS OF ANALYSES/CONCLUSIONS Safety evaluation of proprietary EPRI report, dated December 19, 2007 states that performing the PFM analysis only for recirculation inlet nozzle (N2) is acceptable it has been demonstrated that the recirculation inlet nozzle is limiting among the four nozzles evaluated [11]. This conclusion is applicable to both nozzle-to-shell weld and nozzle blend radii. In addition, increased inside surface fluence on reactor vessel results in decreases of fracture toughness. The N2 nozzles are the only components that have accumulated significant fluence, and the thermal transients introduced to the N2 nozzle are as, or more severe, than the transients experienced by the other applicable nozzles. Based on the limiting fluence and stress cases of the N2 nozzle, the results from this analysis shall bound all MNGP nozzle penetrations to the Reactor Pressure Vessel File No.: 1101463.301 Page 7 of 11 Revision: 2 F0306-01R1
9.1 Nozzle Blend Radii The reliability evaluation is presented using plant specific inspection coverage. The probabilities of failure (PoF) are summarized in Table 2. The in-service inspection of 25% inspection coverage for the period of extended operation of 20 years (70% inspection coverage for initial 40 years of operation) is used at the nozzle blend radius. The average failure probability due to normal operation is 1.17x10-7 per year. The average failure probability due to the LTOP event is 1.05x10-6 per year. The PoF for both normal operation and LTOP event are less than 5x10-6 per year criteria from Reference 21. Therefore this analysis demonstrates the acceptability of reduced in-service inspection per ASME Code Case N-702 at the nozzle blend radii of the recirculation nozzle (N2) at the Monticello Nuclear Generating Plant for the extended period of operation.
9.2 Nozzle-to-Shell Weld The reliability evaluation is presented using plant specific inspection. The probabilities of failure (PoF) are summarized in Table 3. The in-service inspection of 25% inspection coverage for the period of extended operation of 20 years (47% inspection coverage for initial 40 years of operation) is used at the nozzle-to-shell weld. The average failure probability due to normal operation is 2.25x10-6 per year. The average failure probability due to the LTOP event is 1.69x10-9 per year. The PoF for both normal operation and LTOP event are less than 5x10-6 per year criteria from Reference 21. Therefore this analysis demonstrates acceptability of reduced in-service inspection per ASME Code Case N-702 at the nozzle-to-shell weld of the recirculation nozzle (N2) at the Monticello Nuclear Generating Plant for the extended period of operation.
10.0 CONCLUSION
S The probability of failure per reactor year for Normal Operation and LTOP Event at the nozzle-to-shell-weld and nozzle blend radii in the recirculation inlet nozzle at Monticello Nuclear Generating Plant are below the criteria of 5x10-6 per year. The Monticello Nuclear Generating Plant is justified to reduce the in-service inspection coverage of the nozzle-to-shell-weld and the nozzle blend radii of the recirculation nozzle (N2) to 25% per ASME Code Case N-702 for the extended period of operation.
11.0 REFERENCES
- 1. BWRVIP Report, BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP-05), Electric Power Research Institute TR-105697, September 1995.
- 2. VIPER, Vessel Inspection Program Evaluation for Reliability, Version 1.2 (1/5/98), Structural Integrity Associates.
File No.: 1101463.301 Page 8 of 11 Revision: 2 F0306-01R1
- 4. SI Calculation EPRI-180-303, Deterministic Crack Growth Calculation for BWR Nozzle-to-Shell-welds and nozzle blend radii region, Revision 0.
- 5. Document NX8290-13, General Plan Shows Vessel ID (17-2 or 206 Inches) and Vessel Wall Thickness (5-1/5) and Cladding (1/8), SI File Number 1101463.203.
- 6. SI Calculation 1000720.301, Finite Element Stress Analysis of Monticello RPV Recirculation Inlet Nozzle, Revison 0.
- 7. DIT 19181-01 MNGP Recirculation Inlet Nozzle-to-shell-welds VIPER Analysis, QF-0545 (FP-E-MOD-11) Revision 3, SI File 1101463.201.
- 8. SI Calculation W-EPRI-180-302, Evaluation of inspection on the probability of failure for BWR nozzle-to-shell-welds and nozzle blend radii region, Revision 0.
- 10. Document DBD-B.1.1, Design Bases Document for Reactor and Vessel Assembly DBD B.1.1, Revision C, SI File Number 1101463.203.
- 11. BWRVIP Report, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii, 1003557, October 2002, SI File Number BWRVIP-01-308.
- 12. 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
- 13. Document B.01.01-06, Revision 14, Operations Manual Section: Reactor and Vessel Assembly B.01.01-06 Figures, SI File Number 1101463.204.
- 14. Private Communication, P. M. Besuner (Failure Analysis Associates) to P. C. Riccardella, Three Dimensional Stress Intensity Factor Magnification Constant for Radial Feedwater Nozzle Cracks, June 1976.
- 15. NRC, Final Safety Evaluation of the BWR Vessel and Internals, Project BWRVIP-05 Report, TAC # M93925, July 1998.
- 16. DIT 19181-02 MNGP Recirculation Inlet Nozzle-to-shell-welds VIPER Analysis, QF-0545 (FP-E-MOD-11) Revision 3, SI File Number 1101463.201.
- 17. CB&I Drawing No. 7, Revision 9, 12"Ø Nozzle MK. N2 A/K 17-2 I.D. x 63-2 Ins. Heads Nuclear Reactor, Monticello Document No. NX-8920-90, SI File No. 1000720.201.
- 18. BWRVIP Report, Evaluation of Chemistry Data for BWR Vessel Nozzle Forging Materials, 1014995, May 2007, SI File Number BWRVIP-01-373.
- 19. DIT 19181-03 MNGP Recirculation Inlet Nozzle to Shell Welds VIPER Analysis, QF-0545 (FP-E-MOD-11) Revision 3, SI File Number 1101463.206.
- 20. 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.
- 21. 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.
File No.: 1101463.301 Page 9 of 11 Revision: 2 F0306-01R1
Table 1: Monticello Weld Chemistry Shell Inner Thickness/Path BWR Dia Length Initial Plant (in) (in) %Cu %Ni RTndt(F)
Monticello N2 Nozzle-to-shell-weld 206.4 5.0625 0.1 0.99 -65 [8]
Monticello N2 Blend Radii 206.4 9.4845 0.18 0.86 40 [9]
Note: %Cu and %Ni were obtained from Reference 7. Initial RTndt were obtained from References 8 and 9 Nozzle-to-shell-weld and Blend radii respectively.
Table 2: Probability of Failure Results Summary for Nozzle Blend Radii (a) Normal Operation Nozzle Blend Radii at 25%
In-Service Inspection for period of Extended Operation Total PoF at Year 60 7.00x10-6 Average PoF/year for 60 years 1.17x10-7 (b) LTOP Event Nozzle Blend Radii at 25%
In-Service Inspection for period of Extended Operation Total Conditional PoF at Year 60 6.29x10-2 Average Conditional PoF/year for 60 years 1.05x10-3 Average PoF/year for 60 years 1.05x10-6
-3 Note: Probability of LTOP event: 1x10 per year File No.: 1101463.301 Page 10 of 11 Revision: 2 F0306-01R1
Table 3: Probability of Failure Results Summary for Nozzle-to-Shell Weld (a) Normal Operation Nozzle-to-Shell Weld at 25%
In-Service Inspection for period of Extended Operation Total PoF at Year 60 1.35x10-4 Average PoF/year for 60 years 2.25x10-6 (b) LTOP Event Nozzle-to-Shell Weld at 25%
In-Service Inspection for period of Extended Operation Total Conditional PoF at Year 60 1.01x10-4 Average Conditional PoF/year for 60 years 1.69x10-6 Average PoF/year for 60 years 1.69x10-9
-3 Note: Probability of LTOP event: 1x10 per year File No.: 1101463.301 Page 11 of 11 Revision: 2 F0306-01R1
APPENDIX A COMPUTER FILES File No.: 1101463.301 Page A-1 of A-2 Revision: 2 F0306-01R1
File Name Description RID2I25.INP VIPERNZ input file for 25% inspection coverage at nozzle blend radii.
RIC2I25.INP VIPERNZ input file for 25% inspection coverage at nozzle-to-shell-weld.
RID2I25.OUT VIPERNZ output file for 25% inspection coverage at nozzle blend radii.
RIC2I25.OUT VIPERNZ output file for 25% inspection coverage at nozzle-to-shell-weld.
VIPERNOZ1P3.EXE VIPERNZ executable program File No.: 1101463.301 Page A-2 of A-2 Revision: 2 F0306-01R1