NL-14-027, LTR-PAFM-13-115-NP, Rev. 0, Technical Justification to Support Extended Volumetric Examination Interval for Indian Point Generating Station Unit 3 Reactor Vessel Inlet Nozzle to Safe End Dissimilar Metal Welds.
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| Site: | Indian Point  |
| Issue date: | 11/30/2013 |
| From: | Glunt N Westinghouse |
| To: | Office of Nuclear Reactor Regulation |
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Westinghouse Non-Proprietary Class 3 LTR-PAFM-13-115-NP Revision 0 Technical Justification to Support Extended Volumetric Examination Interval for Indian Point Generating Station Unit 3 Reactor Vessel Inlet Nozzle to Safe End Dissimilar Metal Welds November 2013 Author: N.L. Glunt*, Piping Analysis and Fracture Mechanics Verifier: A. Udyawar*, Piping Analysis and Fracture Mechanics Approved: S. A. Swamy*, Manager, Piping Analysis and Fracture Mechanics
- Electronicallyapproved records are authenticatedin the Electronic Document Management System.
© 2013 Westinghouse Electric Company LLC All Rights Reserved OeWestinghouse
LTR-PAFM-13-115-NP Rev. 0 FOREWORD This document contains Westinghouse Electric Company LLC proprietary information and data which has been identified by brackets. Coding (a~c,e) associated with the brackets sets forth the basis on which the information is considered proprietary. These codes are listed with their meanings in WCAP-7211 Revision 5 (November 2005),
"Proprietary Information and Intellectual Property Management Policies and Procedures."
The proprietary information and data contained in this report were obtained at considerable Westinghouse expense and its release could seriously affect our competitive position. This information is to be withheld from public disclosure in accordance with the Rules of Practice 10CFR2.390 and the information presented herein is to be safeguarded in accordance with 10CFR2.903. Withholding of this information does not adversely affect the public interest.
This information has been provided for your internal use only and should not be released to persons or organizations outside the Directorate of Regulation and the ACRS without the express written approval of Westinghouse Electric Company LLC. Should it become necessary to release this information to such persons as part of the review procedure, please contact Westinghouse Electric Company LLC, which will make the necessary arrangements required to protect the Company's proprietary interests.
The proprietary information in the brackets has been deleted in this report, the deleted information is provided in the proprietary version of this report (LTR-PAFM-13-115-P Revision 0).
Page 2 of 20
LTR-PAFM-13-115-NP Rev. 0 1.0 Introduction Service induced cracking of the nickel-base alloy components and weldments have been occurring more and more frequently in recent years, resulting in the need to repair and/or replace these components. Such cracking and leakage have been observed in the reactor vessel upper and bottom head penetration nozzles as well as the dissimilar metal butt welds of the pressurizer and reactor vessel nozzles exposed to the high reactor coolant temperatures. These Pressurized Water Reactor (PWR) power plant field experiences and the potential for Primary Water Stress Corrosion Cracking (PWSCC) require reassessment of the examination frequency as well as the overall examination strategy for nickel-base alloy components and weldments. Code Case N-770-1 (Reference 1) provides the visual and volumetric inspection guidelines for the primary system piping dissimilar metal (DM) butt welds to augment the current inspection requirements.
In accordance with Code Case N-770-1 guidelines, volumetric examinations are required for the unmitigated dissimilar metal butt welds at the Reactor Vessel (RV) inlet nozzles every second inspection period not exceeding 7 years. A volumetric examination was performed for the Indian Point Unit 3 reactor vessel inlet nozzle to safe end dissimilar metal butt welds during the Spring 2009 Re-Fueling Outage (RFO) and no relevant indications were detected.
The next required volumetric examination for the Reactor Vessel inlet nozzle DM welds will be during the Spring 2015 RFO in accordance with Code Case N-770-1. This evaluation will determine the impact of performing the volumetric examination during the Spring 2019 RFO. The time interval between the previous examination in Spring 2009 RFO and the planned examination in Spring 2019 RFO is 10 years rather than the 7 years allowed by Code Case N-770-1. Therefore, a relief request will be submitted to the Nuclear Regulatory Commission (NRC) seeking relaxation from the ASME Code Case N-770-1 examination requirement to be able to defer the volumetric examination from the Spring 2015 RFO to the Spring 2019 RFO. The technical justification to support this relief request is developed in this report based on a flaw tolerance analysis.
The objective of the flaw tolerance analysis is to determine the largest undetected axial and circumferential flaw size that could be left behind in service and remain acceptable for a service life of 10 years from the Spring 2009 RFO to the Spring 2019 RFO. The maximum undetected flaw size will then be compared to the flaw size which could have been reasonably missed during the Spring 2009 inlet nozzle DM weld examination based on the current inspection detection capability.
The following sections provide a discussion of the methodology, geometry, loading and the flaw tolerance analyses performed to develop the technical justification for deviating from the volumetric examination requirements of Code Case N-770-1.
Page 3 of 20
LTR-PAFM-13-115-NP Rev. 0 2.0 Methodology In order to support the technical justification for deferring the volumetric examination from the Spring 2015 RFO to the Spring 2019 RFO, it is necessary to demonstrate the structural integrity of the RV inlet nozzle DM welds subjected to the PWSCC crack growth mechanism. To demonstrate the structural integrity of the DM welds, it is essential to determine the maximum allowable undetected flaw size that would be acceptable in the DM welds for the duration from the Spring 2009 RFO to the Spring 2019 RFO. This maximum allowable flaw size would be the largest flaw size that could go undetected during the Spring 2009 RFO examination of the RV inlet nozzle dissimilar metal weld and be acceptable for a service life of 10 years from the Spring 2009 RFO to the Spring 2019 RFO. The maximum allowable undetected flaw size for a given plant operation duration can be determined by subtracting the PWSCC crack growth for that plant operation duration from the maximum allowable end-of-evaluation period flaw size, which is determined in accordance with ASME Code Section XI (Reference 2).
To determine the maximum allowable end-of-evaluation period flaw sizes and the crack tip stress intensity factors used for the PWSCC analysis, it is necessary to establish the stresses, crack geometry and the material properties at the locations of interest. The applicable loadings which must be considered consist of piping reaction loads acting at the dissimilar metal weld regions and the welding residual stresses which exist in the region of interest.
The latest piping loads at the reactor vessel inlet nozzle DM weld locations are based on the Indian Point Unit 3 Stretch Power Uprate and Snubber Elimination Program, which are documented in WCAP-16212-P (Reference 3a) and WCAP-8228 (Reference 3b).
In addition to the piping loads, the effects of welding residual stresses are also considered. For PWSCC, the crack growth model for the dissimilar metal weld material is based on that given in MRP-115 for Alloy 182 weld material (Reference 4). The nozzle geometry and piping loads used in the fracture mechanics analysis are shown in Section 3.0. A discussion of the plant specific welding residual stress distributions used for the dissimilar metal welds is provided in Section 4.0. The determination of the maximum allowable end-of-evaluation period flaw sizes is discussed in Section 5.0.
The maximum allowable undetected flaw size will be determined based on the crack growth due to PWSCC growth mechanism at the RV inlet nozzle dissimilar metal weld.
The PWSCC crack growth is calculated based on the normal operating temperature and the crack tip stress intensity factors resulting from the normal operating steady state piping loads and welding residual stresses as discussed in Section 6.0. Section 7.0 provides the crack growth curves used in developing the technical justification to deviate from the Code Case N-770-1 guidelines by deferring the volumetric inspection of the RV inlet nozzle DM welds from Spring 2015 to the Spring 2019 RFO.
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LTR-PAFM-13-115-NP Rev. 0 3.0 Nozzle Geometry and Loads The dissimilar metal weld geometry for the Indian Point Unit 3 Reactor Vessel inlet nozzle is based on the nozzle detail drawings (Reference 5). The RV inlet nozzle geometry and normal operating temperature are summarized in Table 3-1.
The piping reaction loads at the RV inlet nozzle DM weld locations are based on the Indian Point Unit 3 structural evaluations (Reference 3) and are summarized in Table 3-
- 2. These loads are used in determining the maximum allowable end-of-evaluation period flaw sizes and the PWSCC crack growth.
Table 3-1 Indian Point Unit 3 Reactor Vessel Inlet Nozzle Geometry and Normal Operating Temperature Dimension Inlet Nozzle Outside Diameter (in.) 32.5 Inside Diameter (in.) 27.5 Thickness (in.) 2.50 RV Inlet Nozzle Normal Operating Temperature = 541.5 0 F Table 3-2 Indian Point Unit 3 Reactor Vessel Inlet Nozzle Piping Loads Forces Moments (kips) (in-kips)
Fx Mx My Mz (Axial) (Torsion) (Bending) (Bending)
Deadweight -0.25 -188.6 -60.9 -413.23 Normal Operating Thermal 16.58 3165.5 -2274.75 -5571.29 OBE (Operational Basis Earthquake) 78.67 63.37 4606.82 755.16 DBE (Design Basis Earthquake) 84.59 68.14 4953.58 812.00 Max LOCA (Loss of Coolant Accident) 1577 3404 13188 3978 Page 5 of 20
LTR-PAFM-13-115-NP Rev. 0 4.0 Dissimilar Metal Weld Residual Stress Distribution The welding residual stresses used in the PWSCC crack growth analysis are determined from the finite element stress analysis (Reference 6) based on the Indian Point Unit 2 Reactor Vessel inlet nozzle DM weld specific configuration. Based on a review of the Indian Point Unit 2 and Unit 3 Reactor Vessel inlet nozzle DM weld specific configurations, the finite element stress analysis in Reference 6 is also applicable to Indian Point Unit 3. Figure 4-1 shows a sketch of the Indian Point Unit 3 inlet nozzle DM weld configuration, which matches the Indian Point Unit 2 inlet nozzle DM weld configuration used in Reference 6. The finite element analysis (FEA) in Reference 6 is based on a two-dimensional axisymmetric model of the inlet nozzle dissimilar metal weld region. The FEA model geometry includes a portion of the low alloy steel nozzle, the stainless steel safe end, a portion of the stainless steel piping, the DM weld attaching the nozzle to the safe end, and the stainless steel weld attaching the safe end to the piping.
The FEA also assumes a 3600 inside surface weld repair with a repair depth of 50%
through the dissimilar metal weld thickness, which is consistent with MRP-287 guidance (Reference 7). The following fabrication sequence was simulated in the finite element residual stress analysis and matches the information provided in the reactor vessel nozzle details drawings (Reference 5):
- The inlet nozzle was welded to the safe end ring forging using an Alloy 182 weld.
The outer and inner diameters of the dissimilar metal weld were machined to finished size.
" An assumed 50% inside surface weld repair 3600 around the circumference was conservatively simulated in the Alloy 182 weld, which is consistent with MRP-287 (Reference 7).
" The welded configuration was then raised to a temperature of 1100°F to simulate post-weld heat treatment.
" Shop hydrostatic test was then performed at a pressure of 3110 psig and a temperature of 300°F
" The safe end was then machined for the piping side weld preparation.
- The machined safe end was welded to a long segment of stainless steel piping using a stainless steel weld.
" A plant hydrostatic test was performed at 2485 psig pressure with a temperature of 300 0 F.
- After the plant hydrostatic test, normal operating temperature and pressure was uniformly applied three times to consider any shakedown effects, after which the model was set to normal operating conditions.
Based on the FEA model, residual stresses at three different cuts (centerline path on the DM weld, and two paths along the fusion lines of the DM weld) from the DM weld were obtained. The hoop welding residual stresses based on the limiting centerline cut along the DM weld are used in the analysis and shown in Figure 4-2. The axial residual Page 6 of 20
LTR-PAFM-13-115-NP Rev. 0 stresses do not vary significantly for the three cuts along the DM weld. The axial residual stresses that would result in the most limiting PWSCC crack growth are used in the analysis and shown in Figure 4-2, which is the stress profile along the centerline path of the DM weld.
AIf 5i e . U.]
or 4it g*1L i-2-----
-??2_
ýw
. 7
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i Nt) g'1KA.PROX.
- ~~~~~1 z
- ",,. 0 MACHIMM70~~c THIS
+., +1 OF. 'JJL0.
I TH S OIA. BE FOQE.
OF VJ-LJDS Figure 4-1: Indian Point Unit 3 Reactor Vessel Inlet Nozzle DM Weld Configuration Page 7 of 20
LTR-PAFM-13-115-NP Rev. 0 a,c,e Figure 4-2: Reactor Vessel Inlet Nozzle DM Weld Through-Wall Residual Stress Profiles with Post-Weld Heat Treatment and 50% Inside Surface Weld Repair Page 8 of 20
LTR-PAFM-13-115-NP Rev. 0 5.0 Maximum Allowable End-of-Evaluation Period Flaw Size Determination In order to develop the technical justification to defer the volumetric examination of the RV inlet nozzle DM welds from the Spring 2015 RFO to the Spring 2019 RFO, the first step is the determination of the maximum allowable end-of-evaluation period flaw sizes.
The maximum allowable end-of-evaluation period flaw size is the size to which an indication is allowed to grow to until the next inspection or evaluation period. This particular flaw size is determined based on the piping loads, geometry and the material properties of the component. The evaluation guidelines and procedures for calculating the maximum allowable end-of-evaluation period flaw sizes are described in paragraph IWB-3640 and Appendix C of the ASME Section XI code (Reference 2).
Rapid, nonductile failure is possible for ferritic materials at low temperatures, but is not applicable to the nickel-base alloy material. In nickel-base alloy material, the higher ductility leads to two possible modes of failure, plastic collapse or unstable ductile tearing. The second mechanism can occur when the applied J integral exceeds the J1, fracture toughness, and some stable tearing occurs prior to failure. If this mode of failure is dominant, then the load-carrying capacity is less than that predicted by the plastic collapse mechanism. The maximum allowable end-of-evaluation period flaw sizes of paragraph IWB-3640 for the high toughness materials are determined based on the assumption that plastic collapse would be achieved and would be the dominant mode of failure. However, due to the reduced toughness of the dissimilar metal welds, it is possible that crack extension and unstable ductile tearing could occur and be the dominant mode of failure. To account for this effect, penalty factors called "Z factors" were developed in ASME Code Section Xl, which are to be multiplied by the loadings at these welds. In the current analysis for Indian Point Unit 3, Z factors based on Reference 8 are used in the analysis to provide a more representative approximation of the effects of the dissimilar metal welds. The use of Z factors in effect reduces the maximum allowable end-of-evaluation period flaw sizes for flux welds and thus has been incorporated directly into the evaluation performed in accordance with the procedure and acceptance criteria given in IWB-3640 and Appendix C of ASME Code Section X1. It should be noted that the maximum allowable end-of-evaluation period flaw sizes are limited to only 75% of the wall thickness in accordance with the requirements of ASME Section XI paragraph IWB-3640 (Reference 2).
The maximum allowable end-of-evaluation period flaw sizes determined for both axial and circumferential flaws have incorporated the relevant material properties, pipe loadings and geometry. Loadings under normal, upset, emergency and faulted conditions are considered in conjunction with the applicable safety factors for the corresponding service conditions required in the ASME Section XI Code. For circumferential flaws, axial stress due to the pressure, deadweight, thermal expansion, seismic and LOCA loads are considered in the evaluation. As for the axial flaws, hoop stress resulting from pressure loading is used. The RV inlet nozzle piping loads (Table 3-2) at the DM weld locations for Indian Point Unit 3 are based on the Indian Point Unit 3 Stretch Power Uprate and Snubber Elimination Program (Reference 3).
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LTR-PAFM-13-115-NP Rev. 0 The maximum allowable end-of-evaluation period flaw sizes for the axial and circumferential flaws at the RV inlet nozzle DM welds are provided in Table 5-1. The maximum allowable end-of-evaluation period flaw size was calculated for the axial flaw with an assumed aspect ratio (flaw length/flaw depth) of 2. The aspect ratio of 2 is reasonable because the axial flaw growth due to PWSCC is limited to the width of the DM weld configuration. For the circumferential flaw, a conservative aspect ratio of 10 is used.
Table 5-1 Maximum End-of-Evaluation Period Allowable Flaw Sizes (Flaw Depth/Wall Thickness Ratio - a/t)
Axial Flaw Circumferential Flaw (Aspect Ratio = 2) (Aspect Ratio = 10) 0.75 0.75 Page 10 of 20
LTR-PAFM-13-115-NP Rev. 0 6.0 PWSCC Crack Growth Analysis A PWSCC crack growth analysis was performed to determine the maximum allowable undetected flaw size that would be acceptable based on ASME Section Xl acceptance criteria (Reference 2) for the duration from the Spring 2009 RFO to the Spring 2019 RFO. The maximum allowable undetected flaw size for the given plant operation duration is determined by subtracting the crack growth due to PWSCC for the specific plant operation duration from the maximum allowable end-of-evaluation period flaw size shown in Table 5-1.
Crack growth due to PWSCC is calculated for both axial and circumferential flaws using the normal operating condition steady-state stresses. For axial flaws, the stresses included pressure and residual stresses, while for circumferential flaws, the stresses considered are pressure, 100% power normal thermal expansion, deadweight and residual stresses. The input required for the crack growth analysis is basically the information necessary to calculate the crack tip stress intensity factor (KI), which depends on the geometry of the crack, its surrounding structure and the applied stresses. The geometry and loadings for the nozzles of interest are discussed in Section 3.0 and the applicable residual stresses used are discussed in Section 4.0. Once K, is calculated, stress corrosion crack growth can be calculated using the applicable crack growth rate for the nickel-base alloy material (Alloy 182) from MRP-1 15 (Reference 4).
For all inside surface flaws, the governing crack growth mechanism for the RV inlet nozzle is PWSCC.
Using the applicable stresses at the dissimilar metal welds, the crack tip stress intensity factors can be determined based on the stress intensity factor expressions from API-579 (Reference 9). The through-wall stress distribution profile is represented by a 4 th order polynomial:
,,(a)j= (3 +a 1(a) +G2 (a 2+'73C(a) 3+ (4 (aD where:
0 0, C7, 02, 03, and a4 are the stress profile curve fitting coefficients, a is the distance from the wall surface where the crack initiates; t is the wall thickness; and a is the stress perpendicular to the plane of the crack.
The stress intensity factor calculations for semi-elliptical inside surface axial and circumferential flaws are expressed in the general form as follows:
K = Gj (a/c, a/t, t/R, 01) aj j=0 Page 11 of 20
LTR-PAFM-13-1 15-NP Rev. 0 where:
a: Crack Depth c: Half Crack Length Along Surface t: Thickness of Cylinder R: Inside Radius
$D: Angular Position of a Point on the Crack Front Gj: Gj is influence coefficient for jf stress distribution on crack surface (i.e.,
Go, G1, G2, G3, G4).
Q: The shape factor of an elliptical crack is approximated by:
Q = I + 1.464(a/c)1 65 for a/c < 1 or Q = 1 + 1.464(c/a)1 65 for a/c > 1.
The influence coefficients at various points on the crack front can be obtained by using an interpolation method. Once the crack tip stress intensity factors are determined, PWSCC crack growth calculations can be performed using the crack growth rate below with the applicable normal operating temperature.
The PWSCC crack growth rate used in the crack growth analysis is based on the EPRI recommended crack growth curve for Alloy 182 material (Reference 4):
d- =exp 1
-- - T-'1r1 a(K)1
[t R Ti Tref]
where:
= Crack growth rate in m/sec (in/hr) dt Qg = Thermal activation energy for crack growth = 130 kJ/mole (31.0 kcal/mole)
R Universal gas constant = 8.314 x 10-3 kJ/mole-K (1.103 x 10-3 kcal/mole-°R)
T = Absolute operating temperature at the location of crack, K (OR)
Tref = Absolute reference temperature used to normalize data = 598.15 K (1076.67 0 R) a = Crack growth amplitude
= 1.50 x 10-12 at 325 0C (2.47 x 10-7 at 617 0 F)
Exponent = 1.6 K = Crack tip stress intensity factor MPa*/m (ksiin)
The normal operating temperature used in the crack growth analysis is 541.5 0 F at the RV inlet nozzle. It should be noted that the fatigue crack growth mechanism is not considered in the crack growth analysis as it is considered to be small when compared to that due to the PWSCC crack growth mechanism at the reactor vessel inlet nozzle for the duration of interest. This is demonstrated by the low fatigue usage factor of 0.049 for Page 12 of 20
LTR-PAFM-13-115-NP Rev. 0 the location of interest at the inlet nozzle documented on Page 4-2 of the Stretch Power Uprate program reactor vessel analytical report WCAP-16209-P (Reference 10).
Therefore, it is not necessary to consider fatigue crack growth in the evaluation.
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LTR-PAFM-13-115-NP Rev. 0 7.0 Technical Justification for Deferring the Volumetric Examination from Spring 2015 RFO to Spring 2019 RFO In accordance with ASME Code Case N-770-1 (Reference 1), the volumetric examination interval for the unmitigated reactor vessel inlet nozzle to safe end dissimilar metal welds must not exceed 7 years. A relief request will be submitted to the NRC seeking relaxation from the ASME Code Case N-770-1 requirement in order to defer the volumetric examination of the reactor vessel inlet nozzle to safe end dissimilar metal welds from the Spring 2015 RFO to the Spring 2019 RFO. Since no relevant indications were detected in the inlet nozzle DM welds during the Spring 2009 RFO, technical justification can be developed to support deferring the volumetric examination from the Spring 2015 RFO to the Spring 2019 RFO by calculating the maximum undetected flaw size that could be left behind in service and remain acceptable for a service life of 10 years from the Spring 2009 RFO to the Spring 2019 RFO. This maximum undetected flaw size can then be compared to the flaw size which could have been reasonably missed during the Spring 2009 RFO inlet nozzle DM weld examination.
The maximum allowable undetected flaw depth is determined by subtracting the PWSCC crack growth for a plant operation duration of 10 years from the maximum allowable end-of-evaluation period flaw depth shown in Table 5-1. The end-of-evaluation period flaw depth is calculated based on the guidelines given in paragraph IWB-3640 and Appendix C of ASME Section Xl Code (Reference 2). The PWSCC crack growth at the Alloy 182 weld is calculated based on the normal operating condition, piping loads, and the welding residual stresses at the DM weld as well as the crack growth model in MRP-115 (Reference 4). The maximum allowable undetected flaw depth was calculated for an axial flaw with an assumed aspect ratio of 2. An aspect ratio of 2 is reasonable for the axial flaw due to the DM weld configuration since any PWSCC axial flaw growth is limited to the width of the weld. For the circumferential flaw, a conservative aspect ratio of 10 is used in the crack growth analysis.
The hoop welding residual stresses used in the PWSCC crack growth analysis for the axial flaw is shown in Figure 4-2. The PWSCC crack growth analysis of the circumferential flaw considered two cases: normal operating piping loads with axial residual stresses (shown in Figure 4-2) and normal operating piping loads without residual stresses in order to obtain the most limiting crack growth results since a portion of the residual stress profile is compressive. The PWSCC crack growth result for the circumferential flaw was found to be more limiting for the case without residual stresses (pressure, deadweight, normal operating thermal loads) than the case with residual stresses (pressure, deadweight, normal operating thermal loads, residual stresses).
The PWSCC crack growth curves and the maximum allowable undetected flaw sizes for an axial flaw and a circumferential flaw are shown in Figures 7-1 and 7-2 respectively.
The horizontal axis displays service life in Effective Full Power Years (EFPY), and the vertical axis shows the flaw depth to wall thickness ratio (a/t). The maximum allowable end-of-evaluation period flaw sizes are also shown in these figures for the respective flaw configurations. Based on the crack growth results from Figures 7-1 and 7-2, the maximum allowable undetected flaw sizes for the axial and circumferential flaws are tabulated in Table 7-1.
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LTR-PAFM-13-115-NP Rev. 0 Table 7-1 Maximum Allowable Undetected Flaw Sizes Axial Flaw Circumferential Flaw (Aspect Ratio = 2) (Aspect Ratio = 10)
Maximum Allowable Undetected Flaw Size (a/t)
Flaw Depth (inches) 0.95 1.025 Flaw Length (inches) 1.90 10.25 The maximum allowable undetected flaw sizes shown in Table 7-1 are the largest undetected axial and circumferential flaw sizes that could be left behind in service and remain acceptable for a service life of 10 years from the Spring 2009 RFO to the Spring 2019 RFO. In accordance with the detection and sizing requirements in Supplement 10 of ASME Section XI Appendix VIII (Reference 2) pertaining to the qualification of inspection procedures, the minimum required detectable flaw depth is 10% of the wall thickness. As a result, based on the current inspection detection capability, these maximum allowable undetected flaw sizes are larger than the flaw sizes that could have been reasonably missed during the last Spring 2009 RFO volumetric examination of the RV inlet nozzle DM welds.
Therefore, deferring the volumetric examination for the RV inlet nozzle DM welds from the 7 years allowed by Code Case N-770-1 to 10 years is technically justified. This is because the maximum allowable undetected flaw sizes that have been shown to be acceptable for a service life of 10 years from the Spring 2009 RFO to the Spring 2019 RFO in accordance with the ASME Section Xl IWB-3640 acceptance criteria are larger than the flaw sizes that might have been reasonably missed during the Spring 2009 RFO.
Page 15 of 20
LTR-PAFM-13-115-NP Rev. 0 0.8 SME Maximum Allowable End-Of-Evaluation Period Flaw Size Ar---L -'-'---, ** *--- '--'-*- * -
0.7
_ _i _ _ I _ _ _ _ I __ __ _
0.6
,__ ___ - I r 0.5
- 0. Maimu Aloal _Fa et 3]_i ,/__ __
.2
__ _ _ _ I. __ - - 1 __ 1 i I 0.2 0
0.1 i__
-4,
-t-,-.
~ ~ 11EFPY I ---. ,-
0 10 Spring 2009 20 Spring 2019 30 40 RFO RFO Service Life (EFPY)
Figure 7-1: PWSCC Crack Growth Curve for Inlet Nozzle Axial Flaw (DM weld), Aspect Ratio = 2 Page 16 of 20
LTR-PAFM-13-115-NP Rev. 0 0.8 /
. .- .. ~ ASME Maximum Allowable End-Of-Evaluation Period Flaw Size I I ~
0.6 __ __..
__ I ____ !__
t Maximum Allowable Undetected Flaw Depth (a/t) 0.41 0.4 - ~ 1 _ {- _
0.3 __ " _ I __ ______
-i-- ___ I _
.. I a.3 0.1 -
~I1EP 0 10 20 Spring2009 30 Spring2019 40 RFO RFO Service Life (EFPY)
Figure 7-2: PWSCC Crack Growth Curve for Inlet Nozzle Circumferential Flaw (DM weld), Aspect Ratio = 10 Page 17 of 20
LTR-PAFM-13-115-NP Rev. 0 8.0 Summary and Conclusions A volumetric examination of the reactor vessel inlet nozzle to safe end dissimilar metal butt welds was performed at Indian Point Unit 3 during the Spring 2009 RFO and there were no relevant indications in the inlet nozzle DM welds. The next required volumetric examination will be during the Spring 2015 RFO in accordance with Code Case N-770-1.
However, the volumetric examination will be deferred to the Spring 2019 RFO for the Indian Point Unit 3 Reactor Vessel inlet nozzle DM welds. Since the time interval between the previous examination in Spring 2009 RFO and the planned examination in Spring 2019 RFO exceeds 7 years, which deviates from the Code Case N-770-1 inspection interval requirements, a relief request will be submitted to the Nuclear Regulatory Commission (NRC) seeking relaxation from the ASME Code Case N-770-1 examination requirement to defer the volumetric examination of the inlet nozzle DM welds.
This letter report provides technical justification to support the relaxation request by performing a flaw tolerance analysis to determine the largest undetected axial and circumferential flaws that could be left behind in service and remain acceptable for a service life of 10 years from the Spring 2009 RFO to the Spring 2019 RFO. The maximum undetected flaw size can then be compared to the flaw size which could have been reasonably missed during the Spring 2009 RFO inlet nozzle DM weld examination.
Based on the PWSCC crack growth analysis results from Section 7.0, the maximum allowable undetected flaw sizes for the reactor vessel inlet nozzle DM welds are tabulated in Table 8-1. These allowable undetected axial and circumferential flaw sizes have been shown to be acceptable in accordance with the ASME Section Xl IWB-3640 acceptance criteria through the Spring 2019 RFO taking into account of potential PWSCC crack growth since the last volumetric examination during the Spring 2009 RFO. In accordance with the detection and sizing requirements in Supplement 10 of ASME Section XI Appendix VIII pertaining to the qualification of inspection procedures, the minimum required detectable flaw depth is 10% of the wall thickness. Therefore, based on the current inspection detection capability, these maximum allowable undetected flaw sizes are larger than the flaw sizes that could have been reasonably missed during the last volumetric examination of the RV inlet nozzle DM welds in Spring 2009 RFO. As a result, deferring the volumetric examination for the RV inlet nozzle DM welds from 7 years allowed by Code Case N-770-1 to 10 years is technically justified.
This is because the maximum allowable undetected flaw sizes that have been shown to be acceptable for a service life of 10 years from the Spring 2009 RFO to Spring 2019 RFO in accordance with the ASME Section Xl IWB-3640 acceptance criteria are larger than the flaw sizes that might have been reasonably missed during the Spring 2009 RFO.
Page 18 of 20
LTR-PAFM-13-115-NP Rev. 0 Table 8-1 Maximum Allowable Undetected Flaw Sizes Axial Flaw Circumferential Flaw (Aspect Ratio = 2) (Aspect Ratio = 10)
Maximum Allowable Undetected Flaw Size (a/t)
Flaw Depth (inches) 0.95 1.025 Flaw Length (inches) 1.90 10.25 Page 19 of 20
LTR-PAFM-13-115-NP Rev. 0 9.0 References
- 1. ASME Code Case N-770-1,Section XI Division 1. "Alternative Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated with UNS N06082 or UNS W86182 Weld Filler Material With or Without Application of Listed Mitigation Activities." Approval Date December 25, 2009.
- 2. Rules for Inservice Inspection of Nuclear Power Plant Components, ASME Boiler &
Pressure Vessel Code, Section Xl, 2001 Edition through 2003 Addenda.
- 3. a. WCAP-16212-P, Rev. 0, "Indian Point Nuclear Generating Unit No. 3 Stretch Power Uprate NSSS and BOP Licensing Report," June 2004.
- b. WCAP-8228, Vol. 1, Rev. 1, "Structural Evaluation of Reactor Coolant Loop/Support System for Indian Point Nuclear Generation Station Unit No. 3," April, 1997.
- 4. Materials Reliability Program: Crack Growth Rates for Evaluating Primary Water Stress Corrosion Cracking (PWSCC) of Alloy 82, 182, and 132 Welds (MRP-115),
EPRI, Palo Alto, CA: 2004. 1006696.
- 5. Drawings for Indian Point Unit 3 RV inlet nozzle:
- a. Combustion Engineering, Inc. Drawing E-234-045-1, Rev. 1, "Nozzle Details for Westinghouse Electric Corp. 173" I. D. Reactor Vessel."
- b. Combustion Engineering, Inc. Drawing E-234-044-9, Rev. 9 "Pressure Vessel Final Machining for Westinghouse Electric Corp. 173" I. D. Reactor Vessel."
- c. Combustion Engineering, Inc. Drawing E-234-064-3, Rev. 3, "Material Identification Vessel for Westinghouse Electric Corp. 173" I. D. Reactor Vessel."
- 6. Dominion Engineering Calculation C-8828-00-01, Rev. 0. "Welding Residual Stress Calculation for Indian Point Unit 2 RPV Inlet Nozzle."
- 7. Materials Reliability Program: Primary Water Stress Corrosion Cracking (PWSCC)
Flaw Evaluation Guidance (MRP-287). EPRI, Palo Alto, CA: 2010. 1021023.
- 8. Materials Reliability Program: Advanced FEA Evaluation of Growth of Postulated Circumferential PWSCC Flaws in Pressurizer Nozzle Dissimilar Metal Welds (MRP-216, Rev. 1): Evaluations Specific to Nine Subject Plants. EPRI, Palo Alto, CA: 2007. 1015400.
"Fitness-For-Service," June 2007.
- 10. WCAP-16209-P Rev. 0, (Addendum to CENC-1 122), "Analytical Report for Indian Point Reactor Vessel Unit No. 3" (Stretch Power Uprate Reactor Vessel Evaluation), Addendum to Combustion Engineering Inc. Report No. CENC-1122, January 2004.
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