SVP-02-102, Supplemental Request for Approval of Pipe Flaw Evaluation and Response to Request for Additional Information

Supplemental Request for Approval of Pipe Flaw Evaluation and Response to Request for Additional Information

ML023380611
Person / Time
Site:	Quad Cities
Issue date:	11/25/2002
From:	Tulon T Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
SVP-02-102
Download: ML023380611 (26)

Text

ExekI n.

Exelon Generation Company, LLC wwwexeloncorp corn Nuclear Quad Cities Nuclear Power Station 22710 206th Avenue North Cordova, IL 61242-9740 SVP-02-102 November 25, 2002 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Quad Cities Nuclear Power Station, Unit 1 Facility Operating License Nos. DPR-29 NRC Docket Nos. 50-254

Subject:

Supplemental Request for Ap-proval of Pipe Flaw Evaluation and Response to Request for Additional Information

References:

(1) Letter from T. J. Tulon (Exelon Generation Company, LLC) to U. S. NRC, "Quad Cities Nuclear Power Station Plans to Inspect and/or Weld Overlay Repair Welds 02BS-F4, 02AS-$4, and 02AD-F1 2," dated January 31, 2001 (2) Letter from T. J. Tulon (Exelon Generation Company, LLC) to U. S. NRC, "Change of Commitment to Inspect and/or Weld Overlay Repair Welds 02BS-F4, 02AS-S4 and 02AD-F1 2," dated August 21, 2002 (3) Letter from T. J. Tulon (Exelon Generation Company, LLC) to U. S. NRC, "Request for Approval of Pipe Flaw Evaluation," dated November 13, 2002 During the Fall 2000 Unit 1 refueling outage (i.e., Q1 R1 6), Exelon Generation Company, LLC (EGC) planned to implement a full structural weld overlay repair of the 02BS-F4 weld in the Reactor Recirculation System piping at Quad Cities Nuclear Power Station, Unit 1. Due to high dose rates, the weld overlay repair was only partially completed. In Reference 1, EGC committed to complete the weld overlay repair of the 02BS-F4 weld, during the Unit 1 refueling outage scheduled for Fall 2002 (i.e., Q1 R17). Subsequently, in Reference 2, EGC described a change of commitment due to anticipated high dose rates during Q1 R1 7 and concerns with performing a chemical decontamination of the Reactor Recirculation System.

Specifically, in lieu of completing the weld overlay of the 02BS-F4 weld during Q1 R17, EGC committed to perform a Performance Demonstration Initiative (PDI) qualified Ultrasonic Test (UT) of the partial weld overlay repair to demonstrate the weld overlay and outer 50% of the original base metal is free from Intergrannular Stress Corrosion Cracking defects. EGC also

November 25, 2002 U. S. Nuclear Regulatory Commission Page 2 committed to request NRC approval of a fracture mechanics evaluation to demonstrate continued structural integrity through at least the next cycle of operation, provided that the examination results are consistent to the previous examination.

On November 7, 2002, the PDI qualified UT of the partial weld overlay repair was performed.

The results of the examination are consistent with the previous examination with the exception that the outer 50% of the pipe wall versus the outer 25% of the pipe wall (including the weld overlay) was examined with a PDI qualified manual technique. There was no change in the examination results (i.e., no additional crack growth was observed on weld 02BS-F4 since the previous inspection).

In the Reference 3, EGC requested NRC approval of the pipe flaw evaluation that provides that technical basis for concluding that continued structural integrity will be maintained at least through the next cycle of Unit 1 operation (i.e., Cycle 18). On November 22, 2002, a telephone conference between the NRC and EGC was held to discuss NRC questions regarding EGC's request for approval of the pipe flaw evaluation. Attachment A to this letter provides responses to the NRC's questions, as discussed in the telephone conference.

During development of responses to the NRC's questions, EGC identified a minor discrepancy in the flaw evaluation rel5ort submitted to the NRC as an Attachment to Reference 3. This error was discussed with'the NRC during the November 22, 2002,',

conference call. Specifically, an error in the calculation of the term in Section 5.1.1 was, identified, which resulted in a slight impact to the calculated Limit Load values and allowable hours of operation. However, the error did not impact the conclusions of the report (i.e.j,

structural integrity will be maintained at least through the next cycle of Unit 1 operation).

EGC has entered this condition into the Corrective Action Program for evaluation.

Attachment B of this letter provides a revised flaw evaluation report, with the error corrected.

Based on information provided to the NRC in Reference 3 and during the conference call, the NRC provided verbal approval of the pipe flaw evaluation in accordance with Generic Letter 88-01. If you have any questions or require additional information, please contact Mr. Wally Beck at (309) 227-2800.

Respectfully, T

othyJ.Tulon ite Vice President Quad Cities Nuclear Power Station

November 25, 2002 U. S. Nuclear Regulatory Commission Page 3 Attachments:

Attachment A:

Attachment B:

Response to Request for Additional Information General Electric Nuclear Energy Report No.

GE-NE-0000-0009-4533-01-Ri, "Quad Cities Unit 1 Evaluation of the Indication in Weld 02BS-F4 in the 28 Inch Recirculation Piping,"

Revision 1, dated November 2002 cc:

Regional Administrator-NRC Region III NRC Senior Resident Inspector - Quad Cities Nuclear Power Station I

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Attachment A RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION NRC Question 1 Provide details about the proposed PDI method in support of your assertion that you can detect flaws in the outer 50% of the original base metal of the piping. Describe the performance demonstration used to qualify the technique and personnel. Discuss the volume interrogated and whether examination was performed in four orthogonal directions. If single-sided examination was used, discuss how this was qualified.

Response to Question 1 1.a.

Provide details about the proposed PDI method in support of your assertion that you can detect flaws in the outer 50% of the original base metal of the piping.

The partial weld overlay (WOL) of the 02BS-F4 weld, was manually examined in the Fall 2002 refueling outage (i.e., Q1 R17) using a Performance Demonstration Initiative (PDI) qualified procedure (i.e., PDI UT-B) and personnel. PDI UT-8 is qualified to examine a maximum depth of 1.42" from the outer diameter (OD) surface. The total depth that must be interrogated for this partial WOL material (average thickness of 0.22")

and the outer 50% of the original pipe wall thickness (0.62") is 0.84". The required thickness of 0.84" for WOL 02BS-F4 is well within the demonstrated depth capability of PDI UT-8 to detect flaws. Therefore, EGC concludes that if flaws exist in the required volume in the outer 50% of the original pipe wall, they can be can be'reliably detected using PDI UT-8.

1.b.

Describe the performance demonstration used to qualify the technique and personnel.

The technique/procedure (i.e., PDI UT-8) has been successfully demonstrated in accordance with approved PDI protocols and requirements. This is the PDI approved generic procedure to examine full structural WOLs.

Similarly, WOL 02BS-F4 examination was performed by personnel who are currently qualified to conduct PDI WOL examination in accordance with approved PDI protocols and requirements.

1.c.

Discuss the volume interrogated and whether examination was performed in four orthogonal directions. If single-sided examination was used, discuss how this was qualified.

02BS-F4 is a 28" pipe-to-pipe weld located in the Reactor Recirculation System. As such, there is access to both sides of the weld/weld overlay to allow examination in the four orthogonal directions. A single-sided examination procedure was not required for this examination.

The required inservice inspection (ISI) examination volume specific to this WOL is defined by adding a 0.5" distance to the toes of the original butt weld. Assuming a 37.5 degree bevel for the original butt weld, the resulting volume is conservatively defined by a rectangle with a length of approximately 2.9" and a width of 0.85". The length of this rectangle is centered about the original weld centerline, and includes a 0.5" distance

Attachment A from the toe of the original butt weld. The width of this rectangle is the depth measured from the OD surface of the WOL.

In order to completely examine the required volume described above, a minimum distance of approximately 3" from the original weld centerline is required to allow examination of the entire volume by a 60 degree angle beam, as required by PDI UT-8.

The WOL was originally designed as a full structural WOL with sufficient WOL width to allow PSI and ISI examinations in accordance with the PDI technique and procedure.

The measured WOL width is 8.4N centered about the original weld centerline. This results in approximately 4.2" width available on each side of the original weld centerline, which exceeds the minimum required distance to support a 60-degree angle beam examination.

NRC Question 2 Usually, the calculated Appendix C allowable flaw size is quite different from the Section Xl limit of 0.75t. Yet, you reported almost the same values in Table 2. Please elaborate on that.

Response to Question 2 EGC has identified a minor error in the flaw evaluation report submitted to the NRC as an Attachment to Reference 3. This error was discussed with the NRC during a telephone conference call between the NRC and EGC on November 22, 2002. Specifically, an error in the calculation of the 3 term in Section 5.1.1 was identified, which resulted in a slight impact to the calculated Limit Load values and allowable hours of operation. However, the error did not impact the conclusions of the report (i.e., structural integrity will be maintained at least through the next cycle of Unit 1 operation). Attachment B of this letter provides a revised flaw evaluation report, with the error corrected. EGC has'entered this condition into the Corrective Action Program for evaluation.

Attachment B General Electric Nuclear Energy Report No.

GE-NE-0000-0009-4533-01-Ri, "Quad Cities Unit 1 Evaluation of the Indication in Weld 02BS-F4 in the 28 Inch Recirculation Piping,"

Revision 1, dated November 2002

GE Nuclear Energy REACTOR SERVICES GE Nuclear Energy 175 Curtner Avenue, San Jose, CA 95125 GE-NE-0000-0009-4533-01-R1 Rev. 1 DRF 0000-0009-4533 Class II November 2002 Quad Cities Unit 1 Evaluation of the Indication in Weld 02BS-F4 in the 28 inch Recirculation Piping November 2002 Prepared for Exelon Prepared by GE Nuclear Energy

GE-NE-OOOO-0009-4533-O1-RI Quad Cities Unit 1 Evaluation of the Indication in Weld 02BS-F4 in the 28 inch Recirculation Piping November 2002 Prepared by:

D.V om Engineer Structural Mecranics and Materials Approved by:

2, R. M. Horn, Manager Hardware Design, Reactor Services GE Nuclear Energy

GE-NE-OOOO-0009-4533-O1-RI IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT Please Read Carefully The only undertakings of the General Electric Company (GE) respecting information in this document are contained in the contract between Exelon and GE, QlR17 Contingency Weld Overlay Services, effective 10/11/02, as amended to the date of transmittal of this document, and nothing contained in this document shall be construed as changing the contract. The use of this information by anyone other than Exelon, or for any purpose other than that for which it is furnished by GE, is not authorized; and with respect to any unauthorized use, GE makes no representation or warranty, express or implied, and assumes no liability as to the completeness, accuracy, or usefulness of the information contained in this document, or that its use may not infringe privately owned rights.

Copyright, General Electric Company, 2002.

i GE Nuclear Energy

GE-NE-O000-0009-4533-OI-RI TABLE OF CONTENTS

1.

EXECUTIVE

SUMMARY

I

2.

BACKGROUND 3

3.

WATER CHEMISTRY 5

4.

IGSCC INDICATION PARAMETERS 5

5.

STRUCTURAL ANALYSIS AND CRACK GROWTH PREDICTION 6

5.1.

Structural Analysis 6

5.1.1.

Limit Load Analysis Methodology 6

5.1.2.

Applied Piping Stresses 8

5.2.

Crack Growth Evaluation 9

6.

ANALYTICAL RESULTS 9

7.

CONCLUSIONS 10

8.

REFERENCES 11 ii GE Nuclear Energy

GE-NE-OOOO-0009-4533-O1-R1 LIST OF TABLES TABLE 1: APPLIED STRESSES AT THE 02BS-F4 WELD.................................................... 12 TABLE 2: TABULATION OF ASME SECTION XI AND APPENDIX C ALLOWABLE FLAW SIZES.....................................................................................................

12 TABLE 3:

SUMMARY

OF CRACK GROWTH RATES AND ALLOWABLE HOURS OF OPERATION.....................................................................................................

12 LIST OF FIGURES FIGURE 1: AVERAGE MONTHLY CONDUCTIVITY DURING THE LAST CYCLE.................. 13 FIGURE 2: ECP HISTORY DURING THE LAST CYCLE....................................................... 14 FIGURE 3: STRESS DISTRIBUTION IN A CRACKED PIPE AT THE POINT OF COLLAPSE.......... 15 iii GE Nuclear Energy

GE-NE-OOOO-0009-4533-OI-RI 1.Executive Summary During the Fall 1998 outage at Quad Cities Unit 1, a flaw indication was identified in weld 02BS-F4 in the 28-inch recirculation piping. Earlier inspections in 1989 and 1996 had identified UT reflectors at the same location, but they were interpreted as root geometry.

Since the weld had been subjected to stress improvement, it appears that the indication existed since 1989 and there was no appreciable growth since then. The evaluation of the indication in 1998 showed that operation for two cycles was justified, but Exelon Generating Company (EGC) decided to implement a full structural weld overlay repair in 2000.

However, because of high dose rate, only three weld layers (0.22 in. thickness) were applied and it was decided that the plant was to return to service with a partial thickness overlay.

The NRC approved deferral of the overlay completion to the next cycle. The weld was classified as a category F weld per Generic Letter 88-01 (cracked with inadequate repair) and shown to be acceptable for continued operation for at least one more cycle.

EGC decided to request a change in their commitment to complete the weld overlay in 2002

[7]. The change of commitment is due to (1) anticipated high dose rates during Q1R17, and (2) results of a review of BWRVIP-75 "BWR Vessel and Internals Project, Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedules". In lieu of repair, EGC used a Performance Demonstration Initiative (PDI) qualified Ultrasonic Testing (UT) of the partial weld overlay repair to demonstrate the weld overlay and outer 50% of the original base metal is free from Intergranular Stress Corrosion Cracking (IGSCC) defects. This varies from the previous practice where only the outer 25% of the original pipe base metal is examined when a full structural weld overlay is in place. The examination results confirmed that there was no cracking in the weld overlay and the outer 50% of the original pipe. Continued operation was evaluated assuming that there was a 3600 full circumferential crack with depth equal to 50% of the original pipe wall thickness. This report describes the results of the fracture mechanics evaluation. The analysis was performed using methods consistent with the ASME Code and NRC approved BWRVIP criteria. It is shown that continued operation can be justified for at least two more cycles (48 months) and the required ASME Code structural I

GE Nuclear Energy

GE-NE-OOOO-0009-4533-O1-R1 margins will be maintained. The fact that the weld has had multiple mitigations implemented - HWC operation with NobleChemTM, stress improvement, and weld overlay and the apparent absence of appreciable growth of the indication since 1989 adds further confidence on the justification for continued operation. As described in Ref. 7, the optimum time for completing the weld repair will be in the next scheduled outage in January 2005 (Q1R18) when the repair can be performed in a lower dose environment. This would be well within the acceptable period of continued operation as defined in this report.

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GE-NE-0000-0009-4533-01-R1

2.Background

During the Fall 1998 outage at Quad Cities Unit 1, a flaw indication was identified in weld 02BS-F4 in the 28-inch recirculation piping. UT reflectors were found in the 1989 and 1996 inspections but were evaluated as root geometry. The 1998 inspection was performed by PDI qualified examiners using EPRI qualified procedures. The 1998 inspection determined that there was an Intergrannular Stress Corrosion Cracking (IGSCC) indication 0.25 in. deep and 27 inches long in the weld region.

As part of the mitigation program to reduce IGSCC susceptibility, the weld was subjected to induction heat stress improvement (IHSI) in 1984. The IHSI treatment was intended to eliminate the tensile weld residual stress pattern and produce a compressive residual stress pattern at the inside diameter surfaces of the girth welds. Fully effective IHSI produces compressive stresses up to the inner 50% of the pipe wall. Even if the IHSI were partially effective, the as-welded residual stresses would have been considerably reduced.

Hydrogen Water Chemistry (HWC) was implemented in 1990 and the plant has been operating with NobleChemTM since April 1999. Considering the fact that the weld has two mitigation measures in place, the likelihood of a service induced crack initiating during the last 10 years is low. This suggests that the weld had an existing indication since 1989 and the root geometry attributed by the 1989 and 1996 inspections was most likely the IGSCC indication identified in 1998. The indication was evaluated using ASME Code Section XI, IWB-3640 and Appendix C [1] procedures using crack growth rates for normal water chemistry based on NUREG-0313, Rev. 2 [2] and shown to be acceptable for continued operation for two 24 month operating cycles (until October 2002).

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GE-NE-OOOO-0009-4533-O1-R1 Nevertheless, it was decided that a weld overlay should be performed on weld 02BS-F4 during the October 2000 outage so that a permanent repair would be in place. The weld overlay was designed to be a full structural overlay that meets ASME Code Case 504-1 requirements. During the welding process the personnel exposure dose rate was found to be

,extremely high. If the overlay were to be completed as planned, the resulting personnel exposure would have been unacceptable. The weld overlay application was stopped after three layers (approximately 0.22 in.). A flaw evaluation was performed in 2000 which provided justification for continued operation until Fall of 2002.

EGC revised their commitment to the NRC regarding the schedule for completion of the full structural overlay on the subject weld [7]. Rather than completing the overlay during the upcoming outage (QI R17), the weld was to remain classified as a category F weld (cracked with inadequate repair) [8], and would be re-examined with a manual PDI UT technique.

The examination would interrogate the outer 50% of the original pipe thickness. EPRI/PDI reviewed the suggested PDI procedure and basis for its qualification and determined that the procedure could effectively examine the 02BS-F4 weld overlay material and the outer 50%

of the original base metal [7].

During the Fall 2002 outage, the subject weld was re-examined using the PDI UT technique discussed in Reference 7. The examination results were consistent with the previous examination and showed that the outer 50% of the original base material and the weld overlay material were free of defects [9].

This report provides the justification for continued operation for two, two year cycles considering a fully circumferential flaw of depth equal to 50% of the nominal pipe thickness.

4 GE Nuclear Energy

GE-NE-OOOO-0009-4533-O1-R1 3.Water Chemistry Quad Cities Unit Ihas been operating with HWC since 1990 and NobleChemTM was implemented in April 1999. Figure 1 shows the conductivity data for the last cycle. It is seen that the overall conductivity ranges from 0.08 to 0.15 gtS/cm over the last cycle. The ECP history during the last cycle is shown in Figure 2. The measured ECP is well below the

-230 mV SHE threshold for crack arrest. IGSCC crack growth is expected to be negligible under these conditions.

The HWC availability was 97.9% for 2001 and 95.5% for 2002 to date. It is expected that HWC availability during the coming cycle will be comparable. Based on the low conductivity and ECP, it is reasonable to expect extremely low (near zero, <1 0-7 in/hr) growth rates during the coming cycle.

4. IGSCC Indication Parameters This flaw evaluation is performed using an assumed set of indication parameters. The indication previously identified in weld 02BS-F4 is assumed to be fully circumferential with a depth of 50% of the nominal pipe thickness, 0.62 inch. The indication initial size is repeated below:

Depth:

0.5*t =

0.62 in Length:

27cR

=

88 in The subject weld was re-examined during the Q1R17 outage using an approved PDI UT technique. The examination results demonstrated that the previously identified indication in the subject weld is bounded by the assumed indication discussed above [9].

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GE-NE-O000-0009-4533-O1-Ri

5. Structural Analysis and Crack Growth Prediction This section describes both the analytical methods used to perform the structural analysis and the crack growth model used to predict the end of cycle crack depth.

5.1. Structural Analysis The pipe weld with the indication can be evaluated using the procedures of ASME Code Section XI, Appendix C and the acceptance criteria of IWB-3640 [1]. The evaluation is based on the limit load failure mechanism. The limit load formulas are used to determine the allowable flaw size. The allowable number of hours of operation are calculated considering crack growth. A 90% HWC availability is assumed. The number of hours per cycle were determined using a 2 year operating cycle and a 100% capacity factor as shown below:

Hours per cycle = 2 years/cycle

• 365.25 days/year

• 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s/day

• 1 = 17532 hours 5.1.1. Limit Load Analysis Methodology The limit load method used in the analysis is consistent with the procedures outlined in Appendix C of Section XI of the ASME Code [1]. A brief description of the method is provided next.

Consider a fully circumferential crack of length, I = 2nrR and constant depth, d. In order to determine the flaw parameters at which limit load is achieved, it is necessary to apply the equations of equilibrium assuming that the cracked section behaves like a plastic hinge. For this condition, the assumed stress state at the cracked section is as shown in Figure 3, where the maximum stress is the flow stress of the material, af. Equilibrium of longitudinal forces and moments about the original neutral axis gives the following equations:

6 GE Nuclear Energy

GE-NE-OOOO-0009-4533-O1-RI 1 _ _ _

( 1 )

t af) d 2-t Pbl 2 c~f.(2

_d )-(2)p

Where, t

pipe thickness in inches d

the flaw depth in inches P3 = angle that defines the location of the neutral axis in radians Pm = Primary membrane stress in psi Pb1 = Failure bending stress in psi The safety factor, SF, is then incorporated as follows:

PbI > Z.SF. Pm+ Pb + Le P-(3)

Pm and Pb are the primary membrane and bending stresses, respectively. P, is the secondary stress and includes stresses from all displacement-controlled loadings such as thermal expansion and dynamic anchor motion. Pe is applicable for flux welds only. All three quantities are calculated from the analysis of applied loading. The safety factor is 2.77 for normal/upset conditions and 1.39 for emergency/faulted conditions. The Z factor is discussed next.

Z Factor The test data considered by the ASME Code in developing the flaw evaluation procedure (Appendix C,Section XI) indicated that the welds produced by a process that did not use a flux had fracture toughness as good or better than the base metal. However, flux welds had lower toughness. To account for the reduced toughness of the flux welds (as compared to non-flux welds) the Section XI procedures prescribe a penalty factor, called a 'Z' factor. The examples of flux welds are submerged arc welds (SAW) and shielded metal arc welds (SMAW). Gas metal-arc welds (GMAW) and gas tungsten-arc welds (GTAW) are examples of non-flux welds. Figure IWB-3641-1 of Reference 1 may be used to define the weld-base 7

GE Nuclear Energy

GE-NE-OOOO-0009-4533-O1-R1 metal interface. The expressions for the value of the Z factor in Appendix C of Section XI are given as follows:

Z =

1.15 [1 + 0.013(OD-4)] for SMAW (4)

=

1.30 [1 + 0.010(OD-4)] for SAW Where OD is the nominal pipe size (NPS) in inches. Except for the root pass, the 02BS-F4 weld was completed using a SMAW process [3]. Therefore, the Z factor in the evaluation was calculated using the expression for SMAW.

5.1.2. Applied Piping Stresses The applied piping stresses were extracted from the 2000 flaw evaluation report [3] and the internal pressure is obtained from the Extended Power Uprate (EPU) design specification [5].

For completeness, the discussion provided in Reference 3 regarding the applied piping stresses is repeated here.

The applied piping stresses were calculated from the reported axial and bending loads at the subject weld. Table 1 shows the calculated values of the stresses for various load cases. The three thermal load cases are the following:

1. System at 5460F, except between MO-1-0202-6A and -6B and from MO-1001-50 to penetration X-12 which is at 135'F

2. System at 546°F, except between MO-1-0202-6A and -6B and all of the line 1-1025 20"-A to penetration X-12 which is at 135°F

3. System at 340'F, except between MO-1-0202-6A and -6B which is at 135*F.

8 GE Nuclear Energy

GE-NE-O000-0009-4533-O1-R1 For the purposes of cross-section area and section modulus calculations, the pipe OD was 28 inches, and the pipe wall thickness was 1.24 inches. The stresses due to weight and seismic were treated as primary stresses (Pm and Pb) and those due to thermal load cases as secondary (Pe).

5.2. Crack Growth Evaluation The crack growth rate approved by the BWRVIP-14 NRC SER [4] for effective HWC is used in this evaluation. Although the crack growth rates provided in BWRVIP-14 [6] are intended for shroud materials, they are consistent with the crack growth rates calculated using the NUREG-0313, Rev. 2 methodology [2]. Considering the agreement between the NUREG-0313 and BWRVIP-14 crack growth rates, it is acceptable to use the BWRVIP-14 value for this flaw evaluation.

Reference 4 approves the use of a factor of improvement (FOI) equal to 2 on the through wall crack growth rate in shroud materials of 2.2 xl 0.5 in/hr for NWC if effective HWC can be demonstrated. Considering the FOI, the effective crack growth rate for 90% effective HWC is 1.1xl0 5 in/hr.

6. Analytical Results The allowable flaw size used to determine the allowable hours of operation is the limiting value between the ASME Code Section XI allowable depth (0.75t) [1], and the Appendix C flaw size predicted using the limit load formulas discussed above. Table 2 summarizes the two allowable flaw sizes. Table 3 summarizes the crack growth rate used and the allowable hours of operation.

9 GE Nuclear Energy

GE-NE-OOOO-0009-4533-O1-R1

7. Conclusions This report presents the fracture mechanics evaluation results for an assumed indication of 3600 length and 0.5t depth (0.62"). The PDI UT results of the subject weld obtained during the Fall 2002 outage demonstrated that the outer 50% of the original base material and the weld overlay material were free of defects [9]. These results are consistent with the assumed flaw parameters used in this evaluation. The ECP and Conductivity data contained in this report demonstrate that the water chemistry during the last cycle satisfied the requirements for effective HWC. It is expected that the water chemistry during the next cycles will remain consistent with the previous cycle; therefore, the crack growth rate used in this evaluation is acceptable. The fracture mechanics evaluation was conducted using the procedures of Appendix C and Paragraph IWB-3640, in ASME Section XI [1]. Considering the weld overlay thickness, at least two, two year fuel cycles of operation may be justified using the NRC approved [4] crack growth rate of 1.1xl 0 5 in/hr for effective HWC.

10 GE Nuclear Energy

GE-NE-OOOO-0009-4533-O1-R1

8. References

[1]

ASME Boiler & Pressure Vessel Code,Section XI, 1995 Edition, through and including 1996 Addenda.

[2]

"Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping," NUREG-0313, Revision 2, January 1998.

[3]

Stark, Randy. "Evaluation of the Indication in Weld 02BS-F4 in the 28 inch Recirculation Piping", GE-NE. San Jose, CA. October 2000. GE-NE-B13-02064-00

18.

[4]

Strosnider, Jack R. "Final Safety Evaluation of Proprietary Report TR 105873 "BWR Vessel and Internals Project, Evaluation of Crack Growth in BWR Stainless Steel Internals (BWRVIP-14)" (TAC NO. M94975)". USNRC. Washington D.C.

December 3, 1999.99-496.

[5]

"Reactor Vessel - Power Uprate Certified Design Specification for Quad Cities 1 &

2", GE-NE, San Jose, CA. May 2000. (26A5588, Rev. 0).

[6]

"Evaluation of Crack Growth in BWR Stainless Steel RPV Internals". BWRVIP-14.

EPRI. Palo Alto, CA. EPRI TR-105873. March 1996.

[7]

Tulon, Timothy J. "Change of Commitment to Inspect and/or Weld Overlay Repair Welds 02BS-F$, 02AS-$4 and 02AD-F12". SVP-02-072. August 21, 2002. Exelon.

[8]

"Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedules".

BWRVIP-75. EPRI. Palo Alto, CA. EPRI TR-1 13932. October 1999.

[9]

"GENE Examination Report", Q I R17-007. GE-NE. November 2002.

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\\ V -OOOO-0009-4533-01-R1 Table 1: Applied stresses at the 02PS-F4 Weld Weld 02BS-F4 Load Stresses (ksi)

Membrane Bending Weight 0.154 0.175 Thermal 1 0.036 3.065 Thermal 21 0.061 3.014 Thermal 3 0.001 1.905 Seismic OBE 0.032 0.193 Seismic DBE 0.063 0.387 Pressure 4.154 0

Table 2: Tabulation of ASME Section X1 and Annendix C ASME Code Limit allowable tlawv sizes.

Limit Load Allowable Flaw Depth Ratio (d/t) 0.75 0.709 Allowable Flaw Depth, in 1.095 1.035 (Considering WOL)

The nominal pipe thickness is 1.24 in The partial weld overlay thickness is 0.22 in Table 3: Summary of crack growth rates and allowable hours of operation.

CCrack Growth Rate, Allowable Hours of Crack Growth Model in/hr Operation BWRVIP-14 with HWC 1.1 x 10-,

37740

1. One cycle of operation is 17,532 hours0.00616 days <br />0.148 hours <br />8.796296e-4 weeks <br />2.02426e-4 months <br />.

12 1.

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GE-NE-OOOO-0009-4533-O1-R1 Figure 3: Stress Distribution in a Cracked Pipe at the Point of Collapse 15 GE Nuclear Energy