RS-09-165, Request for Approval of Updated Pipe Flaw Evaluation Report

Request for Approval of Updated Pipe Flaw Evaluation Report

ML093420189
Person / Time
Site:	Dresden
Issue date:	12/07/2009
From:	Benyak D Exelon Generation Co, Exelon Nuclear
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RS-09-165, TAC M93862
Download: ML093420189 (33)

Text

Fxelon (,cric~ration www exeloncorp corn 4300 Winfii,ld Road Nuclear December 7,2009 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Dresden Nuclear Power Station, Unit 2 Renewed Facility Operating License No. DPR-19 NRC Docket No. 50-237

Subject:

Request for Approval of Updated Pipe Flaw Evaluation

References:

(1)

NRC Generic Letter 88-01, "NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping," dated January 25, 1988 (2)

Letter from B. Rybak (Commonwealth Edison Company) to U.S. NRC, "Dresden Nuclear Power Station Unit 2 Recirculation System and Reactor Head Flaw Indication Evaluations," dated October 16, 1995 (3)

Letter from J. Stang (U.S. NRC) to D. Farrar (Commonwealth Edison Company), "Dresden, Unit 2, Flaw Evaluations (TAC No. M93862),"

dated February 15, 1996 In accordance with Generic Letter (GL) 88-01, "NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping," dated January 25, 1988 (i.e., Reference I), Exelon Generation Company, LLC (EGC) is submitting an updated pipe flaw evaluation for a weld in the Reactor Recirculation (RR) system piping at Dresden Nuclear Power Station (DNPS), Unit 2 that EGC proposes to leave as-is without repair. Two flaw indications located at the weld do not meet the acceptance standards of the American Society of Mechanical Engineers (ASME)

Boiler and Pressure Vessel Code (Code)Section XI, 1995 Edition with 1996 Addenda, which is the DNPS Unit 2 Section XI code of record for the current fourth inspection interval, for continued operation without repairlreplacement or flaw evaluation.

An NRC Region Ill inspector raised questions, during the recent DNPS Unit 2 refueling outage (i.e., D2R21), regarding this weld, which contains two previously known circumferential flaw indications. The inspector's questions centered on the fact that neither the original flaw evaluation (Reference 2) nor the NRC safety evaluation (Reference 3) discussed the examination limitations. During a subsequent telephone call, it was agreed to update our flaw evaluation to the current ASME Code edition and addenda and to clearly identify and justify the examination limitations.

December 7,2009 U. S. Nuclear Regulatory Commission Page 2 The weld inspections were conducted in accordance with Reference 1 and Boiling Water Reactor Vessel and lnternals Project (BWRVIP) Report 75-A, "Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedules," dated October 2005. The UT examination used Performance Demonstration Initiative (PDI) qualified personnel, equipment, and procedures. The circumferential flaw indications are located in weld PS2-TEEl202-4B, which is a Category "F" weld, on the tee side of a 28-inch diameter "A" Recirculation Pump tee to valve weld. There are no axial flaw indications detected in this weld. The tee material is SA403 WP304 wrought material; the valve material is SA-358 CF8M cast stainless steel material. The length of the flaw indications are identified in the attachment to this letter and are unchanged from the previous examination which also used PDI qualified personnel, equipment, and procedures. In fact, the two indications have not changed since first identified in 1995 (see Reference 2) and, therefore, sample expansion is not necessary.

An evaluation of the circumferential flaw indications assuming a conservative estimate of crack growth using linear elastic fracture mechanics has been performed by Structural Integrity Associates, Inc. and is attached to this letter. The evaluation was performed using the methodology and acceptance criteria specified in ASME Code,Section XI, 1995 Edition with 1996 Addenda, subarticle IWB-3640, "Evaluation Procedures and Acceptance Criteria for Austenitic Piping," and the guidance of NUREG-031 3, Revision 2, "Technical Report on Material Selection and Process Guidelines for BWR Coolant Pressure Boundary Piping."

This flaw evaluation considered a conservative flaw size and expected growth rates assuming hydrogen water chemistry and plant chemistry parameters. It demonstrates that substantial structural margin exists for more than one operating cycle (i.e., 24 months duration) considering hydrogen water chemistry, since the acceptance criteria of subarticle IWB-3640 are met.

In accordance with the ultrasonic examination report, included as Appendix B to the attached flaw evaluation, there were no examination limitations on the tee side of the weld. However, given the valve configuration and cast material, the required examination volume on the valve side could not be examined. Delta ferrite measurements were taken on the valve body and the results demonstrate an average delta ferrite content of 14.6 FN. Given the high ferrite content, the valve body can be considered resistant to intergranular stress corrosion cracking in the boiling water reactor environment. Using a conservative treatment of residual stress, flaw shape and length and crack growth correlations, the evaluation concluded that, at the end of the evaluation period, a 360 degree flaw would still be below the IWB-3641 allowable flaw depth of about 60 percent.

Therefore, EGC has concluded that the flaw indications are acceptable as-is for continued operation through the next operating cycle. Weld PS2-TEEl202-48 classification will remain as Category "F" (i.e., cracked weld with inadequate or no repair) and will require inspection each subsequent refueling outage.

There is no proprietary information contained in either this letter or its attachment.

EGC requests NRC review and approval of the attached flaw evaluation report by February 25, 2010.

December 7,2009 U. S. Nuclear Regulatory Commission Page 3 Should you have questions concerning this submittal, please contact Mr. Timothy Byam at (630) 657-2804.

Darin Benyak I

Director - Licensing and Regulatory Affairs Exelon Generation Company, LLC

Attachment:

Structural Integrity Associates, Inc. Report, "Crack Growth Analysis for Weld PS2-TEEl202-46," File No.: 0901 309.303

ATTACHMENT Structural Integrity Associates, Inc. Report, "Crack Growth Analysis for Weld PS2-TEEJ202-4B" File No.: 0901 309.303

CALCULATION PACKAGE Dresden Unit 2 Recirculation Line Flaw Evaluation Requisition 109492 1 Exelon GSA 441 728 CALCULATION TITLE:

Revision Description Independent Verifier Page 1 of 8 F0306-01 RO

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Table of Contents

1.0 INTRODUCTION

/OB JECTIVE 3

2.0 METHODOLOGY............................................................................................................. -3 3.0 DESIGN INPUTS................................................................................................................ 4 3.1 Materials and Geometry.......................................................................................... 4 3.2 Loading 5

4.0 ASSUMPTIONS 5

5.0 CRACK GROWTH CALCULATIONS 6

5.1 Fatigue Crack Growth (FCG)..................................................................................

6 5.2 Intergranular Stress Corrosion Cracking (IGSCC)..................................................

6 6.0 RESULTS AND CONCLUSIONS 7

7.0 REFERENCES

.................................................................................................................... 7 APPENDIX A FATIGUE CRACK GROWTH PC-CRACK OUTPUT FILES A-1 APPENDIX B GE-HITACHI INSPECTION REPORT D2R2 1-029 DATED 1 1/10/09 B-1 APPENDIX C DRESDEN TRANSMITTAL OF DESIGN INPUT (TODI) 09.045.

"HYDROGEN WATER CHEMISTRY DATA. " DATED NOVEMBER 30. 2009.... C-1 APPENDIX D DRESDEN TRANSMITTAL OF DESIGN INPUT (TODI) 09.048. "VALVE BODY MATERIAL PROPERTIES FOR VALVE 2.0202.4B. " DATED NOVEMBER

23. 2009......................................................................................................................... D-1 File No..090130 9.303 Revision: 0 Page 2 of 8

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1.O INTRODUCTION/OBJECTIVE The purpose of this calculation is to update the evaluation of two flaw indications observed in Weld PS2-TEEl202-4B, located in the Dresden Generating Station Unit 2 Reactor Recirculation System piping, to account for changes in its input data that have occurred since it was originally performed in 1995 [2]. An ultrasonic examination of the weld was performed during outage D2R2 1, which identified two indications requiring evaluation [5]. These indications were also identified during the previous inspections. Also, Dresden's committed Edition of ASME Boiler & Pressure Vessel Code Section XI, which governs the evaluation has changed, and the piping analysis of the Reactor Recirculation System has been revised [I]. The evaluation that follows incorporates these changes to the input data. The evaluation applies linear elastic fracture mechanics (LEFM) to establish a conservative estimate of the growth of the flaw. Both fatigue crack growth (FCG) and Intergranular Stress Corrosion Cracking (IGSCC) are considered.

2.0 METHODOLOGY The allowable end-of-evaluation period flaw depth to thickness ratio (dt) was taken from Table IWB-3641-1, which is derived by the methodology of Appendix C of ASME Code,Section XI [3].

The geometry and applied stresses for Weld PS2-TEEl202-4B are summarized below.

A conservative estimate of crack growth is computed using linear elastic fracture mechanics (LEFM) techniques. Both fatigue crack growth (FCG) and Intergranular Stress Corrosion Cracking (IGSCC) are considered. The time for the observed flaw to grow to the Section XI IWB-3641-1 allowable depth is determined.

The fatigue crack growth (FCG) analysis was performed using the LEFM analysis option in pc-CRACK for WindowsTM software [4]. This software includes options for the evaluation of FCG, and allows for defining load cases, material properties, crack models, and the selection of the applicable crack growth law. FCC results are summarized in Section 5.1. The IGSCC growth calculation follows in Section 5.2.

Appendix A contains pc-CRACK output of the FCG growth analysis.

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3.0 DESIGN INPUTS 3.1 Materials and Geometry The dimensions of Weld PS2-TEEl202-4B are described below. These are taken from the previous evaluation [2].

Tee Material (from [2] page 8):

SA403 WP 304 Austenitic Stainless Steel Valve Material (From Appendix D)

SA-358 CF8M Cast stainless steel Allowable Stress Intensity Sm (typical): 16.95 ksi Yield Strength (typical):

30 ksi Outside diameter:

OD = 28 inches Wall Thickness (from [2] page 18):

t = 1.2 inches Configuration:

Tee-to-valve Yield and ultimate strength values are not used in the fatigue crack growth calculation below. The yield strength above is taken as a conservative upper bound on residual stress, as noted under assumption 2 below. A copy of the examination report is enclosed in Appendix B. The NDE inspection report [5]

identifies two reportable indications, circumferentially oriented on the tee side of the weld. Reported dimensions are: Indication 1: length = 1.7 inches, depth = 0.36 inches; Indicaton 2: length = 2.8 inches, depth = 0.27 inches. The inspection report states that there is no change in this data from previous examinations. Per the inspection report [5], indication 1 is located from 16.2 to 17.9 inches azimuthally, and indication 2 is located from 68.6 to 71.4 inches azimuthally. The minimum separation is therefore about 32 inches.

This corresponds to about 37% of circumference. These individual indications are sufficiently far apart to be evaluated separately.

Combined Flaw Length to Pipe Circumference Ratio:

0.05 Table IWB 3641.1 Allowable Depth to Wall Thickness (dt) Ratio:

0.75 Measured Flaw Depth to Wall Thickness (dt) Ratio:

0.3 (Note

this is the ratio of flaw depth to nominal wall thickness of 1.2 inches)

Because of component geometry, the inspections were single-sided on the tee side of the weld. No limitation of inspection of the flaw locations was noted in the report on the tee side of the weld [5]. The cast stainless steel valve body on the valve side of the weld is highly resistant to IGSCC. There has been no known IGSCC in cast austenitic stainless steel components in the BWR. Measurements of delta ferrite content in this valve body were taken by Exelon, and the results (included in Appendix D of this calculation) demonstrate an average delta ferrite content of 14.6 FN. The NRC has supported the fact that cast austenitic stainless steels are resistant to IGSCC in the BWR and has noted that CF3 and CF8 valve and pump bodies, welded to resistant materials, are considered resistant to IGSCC (NUREG-03 13 Revision 2). In addition, the NRC has supported a carbon ferrite concept by allowing low carbon cast components such as CF3 with sufficient ferrite (7.5 FN), to be considered resistant to IGSCC in the File No.:0901309.303 Page 4 of 8 Revision: 0

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BWR environment. As ferrite number increases, higher carbon material is considered resistant as well, as discussed in [9].

3.2 Loading Loads considered in this evaluation are taken from [I] and [2]. These affect only the fatigue crack growth portion of the present calculation, since a constant (K-independent) IGSCC crack growth correlation is used as described below. Stresses considered in the fatigue crack growth calculation include pressure, deadweight (not cyclic, but conservatively included with pressure in equation 8 calculations in [I]), and thermal cycling, taken from [2]. Appendix D of [1] reports stresses at the top ten locations for each load combination (equation). The highest value reported for equation 8 (design pressure plus dead weight) at any location is 8970 psi. This value will be conservatively used in the fatigue crack growth calculation. The thermal transient stress is reported in [2, page 201 as 7752 psi.

The sum of these two values (=8970+7752= 16722 psi) will be used as the maximum of the stress cycle for fatigue crack growth calculation.

Weld residual stresses are steady state secondary stresses. Since a K-independent crack growth correlation is used in the IGSCC crack growth calculation as described below, residual stresses do not affect the IGSCC growth calculation in this evaluation. Because they are steady state stresses, residual stresses only affect fatigue crack growth as a mean stress. Therefore, in lieu of using a residual stress distribution, a constant stress of 30 ksi (comparable to material yield stress) will conservatively be used as a mean stress in the fatigue crack growth calculation.

Seismic stresses, either OBE or SSE are not generally included in fatigue crack growth calculations, because the number of expected cycles is so small over the life of the plant. Also, there is an extremely low probability of an operating basis earthquake (OBE) seismic event occurring in the two year evaluation period, and the number of fatigue cycles per event is small. Therefore, the use of the primary stress values from [2] as described above is appropriate.

4.0 ASSUMPTIONS Basic assumptions for the analysis are listed below:

1. Weld PS2-TEE/202-4B in the Dresden recirculation system does not experience a large number of cycles. Most stress cycles in this system are due to plant start-up/shutdown cycles.

2. Residual stress is modeled as a constant tensile stress of 30 ksi (comparable to material yield stress) for the purpose of fatigue crack growth. Residual stresses are steady state secondary stresses and therefore only act as a mean stress in regards to fatigue crack growth.

3. The flaw indications are modeled as extending 360 degrees circumferentially around the pipe in the LEFM crack growth calculations.

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4. Hydrogen water chemistry performance as summarized in Appendix C is assumed to be representative of the performance over the next operating cycle.

5.0 CRACK GROWTH CALCULATIONS 5.1 Fatigue Crack Growth (FCG)

FCG is calculated using the loading described in Section 3.2 of this calculation. In order to eliminate flaw length changes from consideration, the initial flaw was conservatively assumed to extend 360 degrees around the circumference. The minimum stress is taken as a constant value of 30 ksi, conservatively representing weld residual stress, which is present when the plant is shutdown. The maximum stress intensity K,,

is the sum of the pressure, thermal, and deadweight stresses and the (constant) residual stress. K,,, corresponds to residual stress only. This stress intensity range represents the range seen from shutdown to startup. The FCG after 1000 applied cycles is calculated using pc-CRACK software using the crack growth law for austenitic material in air, as described by Figure C-3210-1 in Section XI Appendix C of the ASME Code [3]. (Note that ASME Section XI does not yet include fatigue crack growth laws for austenitic stainless steel in a BWR water environment.) A value of 1000 cycles was chosen as a conservative estimate that more than doubles the total number of startup and shutdown cycles the plant will likely experience over its lifetime Appendix A contains output files for the fatigue crack growth calculation. Over the assumed 1000 cycles including primary and thermal stresses as described above, the flaw is predicted to grow from 0.36 inches deep to 0.3687 inches deep, or a growth of 0.0087 inches.

5.2 Intergranular Stress Corrosion Cracking (IGSCC)

PVP2007-26618 [6] provides stress corrosion crack growth laws for austenitic stainless steel materials in the BWR environment, considering both normal water chemistry and hydrogen water chemistry. This paper provides the technical basis of crack growth correlations used in BWRVIP-14A. Appendix C of this calculation presents a historical summary of hydrogen water chemistry present in the Dresden Unit Reactor Recirculation System. Based on the data in Appendix C, Weld PS2-TEEl202-4B is mitigated by effective hydrogen water chemistry. On that basis, the SCC crack growth law for hydrogen water chemistry provided by Reference [6] equation 6 is justified.

The SCC growth law is taken from Reference [6], Equation 6:

Note that this correlation is K-independent (stress independent as well). Using this correlation, the larger of the observed indications, which is reported as 0.36 inches deep, would grow by 0.193 inches in two years (17520 hours). The flaw depth would increase from 0.36 inches to 0.553 inches, which is 46% of the nominal wall thickness as compared to the allowable depth of 75% for the actual reported File No.:0901309.303 Revision: 0 Page 6 of 8 F0306-01

@ Strucfml lnfegrity Associates, Inc, combined flaw length of 4.5 inches. The allowable depth to thickness ratio (for primary stress ratio of 0.6) from Table IWB 3641.I for a flaw that extends 360' around the circumference of the pipe is 63% of wall thickness. Assuming that the combined flaw grows from its current length of 4.5 inches to 360' of pipe circumference (88 inches in length), the reported flaw is predicted to grow to 63% of wall thickness in 3.9 years. Using the IGSCC crack growth correlation along with the FCG results from above, and assuming that the flaw grows to 360°, a starting depth of 0.52 inches would be required for the flaw to grow to 63% of pipe wall thickness over the next two year inspection interval. On that basis, the current combined flaw could have been approximately 40% deeper than the currently reported depth and still not grow beyond the IWB 3641.1 allowable depth over the next operating cycle.

6.0 RESULTS AND CONCLUSIONS The present analysis shows that over the next operating cycle, the deepest observed indication would conservatively be expected to grow from the current 0.36 inches depth to a maximum depth of 0.57 inches due to the combination of IGSCC and fatigue crack growth. This end of evaluation period depth represents 47.5% of the wall thickness, as compared to allowable depths per IWB-3641 of 75% for the as-measured flaw length and about 60% if the flaw were to grow to 360° in circumference. Thus, the final flaw size will remain acceptable by the criteria of ASME Section XI IWB-3640, well beyond the end of the next two year inspection interval. This prediction is quite conservative because of the very conservative treatment of residual stress, flaw shape and length applied stresses considered, and crack growth correlation used [5].

7.0 REFERENCES

I. Exelon Analysis DRE-98-0017, Reactor Recirculation Piping Stress Analysis, Revision 001 A, May, 2004

2. Commonwealth Edison Calculation NED-P-MSD-085, Flaw Evaluation for PDIA-Dl4 and PS2-TEEl202-4B Welds in Dresden Unit 2 Recirculation System, Revision 0, 101 1995.

3. ASME Boiler & Pressure Vessel Code,Section XI, 1995 Edition, with 1996 Addenda

4. pc-CRACK for Windows, Version 3.1-98348, Structural Integrity Associates, 1998.

5. GE-Hitachi Inspection Report D2R2 1-029 Dated 1 111 0109. Included as Appendix B to this calculation.

6. ASME Paper PVP2007-26618, Technical Basis for BWRVIP Stainless Steel Crack Growth Correlations in BWRs, 2007

7. Dresden Transmittal of Design Input (TODI)09-045, "Hydrogen Water Chemistry Data,"

dated November 30,2009. Included as Appendix C to this calculation.

8. Dresden Transmittal of Design Input (TODI)09-048, "Valve Body Material Properties for Valve 2-0202-4B," dated November 23, 2009. Included as Appendix D to this calculation.

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9. EPRI Report NP-7103-D, "Justification for Extended Weld-Overlay Design Life", January 199 1, prepared by Structural Integrity Associates.

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APPENDIX A FATIGUE CRACK GROWTH PC-CRACK OUTPUT FILES File No.: 0901309-303 Page A-1 of A-7 Revision: 0

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tm pc-CRACK for Windows Version 3.1-98348 (C) Copyright '84 - '98 Structural Integrity Associates, Inc.

33 15 Almaden Expressway, Suite 24 San Jose, CA 951 18-1557 Voice: 408-978-8200 Fax: 408-978-8964 E-mail: pccrack@structint.com Linear Elastic Fracture Mechanics Date: Mon Nov 16 06:47:35 2009 Input Data and Results File: FCGTEE.LFM

Title:

0901309.303 Fatigue crack growth for weld PS2-Tee-2024B Load Cases:

Stress Coefficients Case ID CO Cl C2 C3 Type Unit 10 0

0 0 Coeff

---

Through Wall Stresses for Load Cases With Stress Coeff-------

Wall Case Depth Unit File No.: 0901309-303 Page A-2 of A-7 Revision: 0

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Crack Model: Circumferential Crack in Cylinder (t/R=O. 1)

Crack Parameters:

Wall thickness:

1.2000 Max. crack size:

0.8000 Stress Intensity Factor....................

Crack Case Size Unit File No.: 0901309-303 Page A-3 of A-7 Revision: 0

Slmtml Inlegrify Associates, lnc, Crack Growth Laws:

Law ID: XI-Air Model: ASME Section XI - austenitic stainless steel in air environment da/dN = C

• 10°F

• S

• dKA3.3 where S = 1.0 for R < 0

= 1.O+ 1.8

• R forO<R<0.79

= -43.5 + 57.97

• R for 0.79 < R < 1 F = code specified function of temperature dK = Kmax - Kmin R = Krnin 1 Kmax where:

C

• 10°F = 1.1017e-010 is for the currently selected units of:

force: kip length: inch temperature: 70.0000 Fahrenheit Material Fracture Toughness KIc:

Material ID: K 200 Depth KIc File No.: 0901 309-303 Page A-4 of A-7 Revision: 0

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Initial crack size=

0.3600 Max. crack size=

0.8000 Number of blocks=

1 Print increment of block=

1 Cycles Calc. Print Crk. Gnv.

Mat.

Subblock

/Time incre. incre. Law Klc operating 1000 10 10 XI-Air K 200 Kmax Kmin Subblock Case ID Scale Factor Case ID Scale Factor operating Unit 3.0000 Unit 3.0000 Unit 1.6722 Crack growth results:

Total Subblock Cycles Cycles DaDn

/Time /Time Kmax Kmin DeltaK R DaDt Da a dthk Block:

I 10 6.69e+001 4.29e+001 2.39e+001 0.64 8.45e-006 8.45e-005 0.3601 0.30 20 6.69e+001 4.30e+001 2.39e+001 0.64 8.4%-006 8.45e-005 0.3602 0.30 30 6.69e+001 4.30e+001 2.39e+001 0.64 8.46e-006 8.46e-005 0.3603 0.30 40 6.69e+001 4.30e+001 2.40e+001 0.64 8.46e-006 8.46e-005 0.3603 0.30 50 6.69e+001 4.30e+001 2.40e+001 0.64 8.47e-006 8.47e-005 0.3604 0.30 60 6.70e+001 4.30e+001 2.40e+001 0.64 8.48e-006 8.48e-005 0.3605 0.30 70 6.70e+001 4.30e+001 2.40e+001 0.64 8.48e-006 8.48e-005 0.3606 0.30 80 6.70e+001 4.30e+001 2.40e+001 0.64 8.49e-006 8.49e-005 0.3607 0.30 90 6.70e+001 4.30e+001 2.40e+001 0.64 8.49e-006 8.49e-005 0.3608 0.30 100 6.70e+001 4.30e+001 2.40e+001 0.64 8.50e-006 8.50e-005 0.3608 0.30 110 6.70e+001 4.30e+001 2.40e+001 0.64 8.50e-006 8.50e-005 0.3609 0.30 120 6.70e+001 4.30e+001 2.40e+001 0.64 8.51e-006 8.51e-005 0.361 0.30 130 6.71e+001 4.31e+001 2.40e+001 0.64 8.52e-006 8.52e-005 0.361 1 0.30 File No.: 0901 309-303 Page A-5 of A-7 Revision: 0

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APPENDIX B GE-HITACHI INSPECTION REPORT D2R21-029 DATED 11/10/09 File No.: 0901309-303 Page B-1 of B-8 Revision: 0

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2/t /6)202B-28/FSZ-TEE/2024B bzF.ti-b t l Augmented Exam, XGSCC Category F Uompnent: 'I'ee-Valve 28" Dla. 1,82" Xam. Cal Block: DRE-PDI-304-01 Procedure: GE-PDI-UT-Z,Rev* 4 DRfZ 07-04 / a?-29; CE-PDI-UT4,Rev.2 Manual Exam Scan as s h m in fiw. The absolute minimum scan incf udm the wefd pius 1/4".

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Previous indications observed.

Any questions or problem call S. Snyder @ 4262 or 3. Erichn @ 4276 ASAP

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APPENDIX C DRESDEN TRANSMITTAL OF DESIGN INPUT (TODI)09-045, "HYDROGEN WATER CHEMISTRY DATA," DATED NOVEMBER 30,2009 File No.: 090130-303 Page C-1 of C-1 Revision: 0

Slmctidal lnfegrity Associates, Inc.

ation: Dresden Unit 2 Source Docments:

f. Exelon lnspectbn Relief Data 2009, Oresden Summary dated 10-31-09,

2. EPRI BWRViA Radiolysls Wel V. 2.0, War Ratio vs Feedwater Hydrogen Concentration Output Chart Dated r 1-22-09.

DistriMian: H. L. Gustin fShucturaf lntsgrtty Associates), C. Podczerwinski, T. Loch, J. Strrnec, D. Malauskas, P. Mankoo File No.: 090130-303 Revision: 0 Page C-2 of C-2

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File No.: 090130-303 Page C-3 of C-3 Revision: 0 EXELON TRAPISMTTTAL OF DESIGN IMFQRABATYON TODI NO.09-045 Revision 4 of 3.

According to the s a f q evaluatbn dated January 30,2001, on the BWRVIP-62 report, the crack growth rate of 1.fx 10' inhr can onfy be: used for units having effective hydrogen water chemistry (HWC) and acceptabte noble metal chemical application (NMCA). The criteria tot acceptable NMCA:

ff ) hydrogen vs. oxygen mdar ratio ts 4:1 and above, (2) mmitortng program is required, and (3) is available 90 percent of the time, The following data supports critena f -3 above:

Dresden Nuclear Power StationTkDNPS) Unit 2 has been operating under Hydrogen Water Chemfstry

( W C ) since 1983. NottleChem was implemented dunng the fall outage in 1999, and reappi'catin of lWobleChmrM 'HS performed in ttre fall of 2003. DNPS Unrt 2 has been under a NAflCA pmgram stme the fati of f 999 whtch supseed~s the HWC program.

1. Mofar Ratio:

The cycle average measured hydrogen to oxygen mokr ratb over the last two DNPS Unit 2 cycles was 4953 for Cycle 20 and 289:1 tor Cycle 21 at the measured location agarnst a goal of > 4:l measured. The measured location ts a sample, which is representative of the reurculation loop.

As calculated by the EPRl Wifing Water Reactor Vessel and tnternals Assessment JBWRVIA) Radidysb Model, hydrogen is injected tnto the feedwater at rt rate necessary to achieve a hydrogen to oxygen molar ratlo of 2 3:1 m the Upper Downcomer (DC-UP) which corresponck; to a molar ratio of approxtmatefy 3.5:f in the recirculation pipe (RC-END). Chart I below provtdes the output of a recent DNPS Untt 2 BWRVIA model.

2. Monitoring Program:

DNPS has an Bffective monitwing program for determining catalyst loading and pradicttng reapplication.

3. Hydrogen Availability:

• Hydrogen avatiabtlfty over the last two cycles has been bewen 98.3% and 99.5% at temperatures above ;?OO"F.

r Barnng major fransiants, it is expected that hydrogen avttikbtlity will be above 98% witti a goal of at least lbOO/o during the coining cycle.

Other mitigation parameters over the last two cycfes were as fottaws:

Electrochemicaf Comron Potentiat (ECP):

• Etectrochemical Corrosmn Potential (ECP) over the last two DNPS Unit 2 cycles ranged from -491 to -476 mV (SHE) with a gml of.c -230 mV (SHE). ECP is current& monitored in the Mobb Metfais Iv90nrtoring Skid (NMMS), whrch is located on a sKte kmp off Qf the reactor water cteanup system.

I

@ structural lnfeprity Associates, lnc.

File No.: 090 130-303 Page C-4 of C-4 Revision: 0 EXELON TRAPISMrrrAt OF DESIGN iNFORMATlON 1

TODI NO, @?-045, R e v a m Page.a of 3, Reactor dissolved oxygen at 0.3 ppb, wrth an inspection relief goal of < 5 ppb Average Reactor condlretkrify betw8n 0.074 and 0.095 pSlm, wrth an inspection relief goal of c. 0.3 @/cm r Trend of Pt/Rh foading is projected to have sufficient noble chemhfs for cycle 22, meettrig the inspaion relief goEcl of ~0.15 p@cm2. On-bne Noble Chern ( O M ) is planned for appficatm to DIVPS Unrt 2 ~n 2010 to malntain the WRh loading*

R is current& expected that Cycle 21 chemistfy can be rautinety achieved, bantng mqor transients.

PravklBd the rsack?r water chemistry can be maintained in Cy& 22 at comparabk values to those in cycle 21, the Hydrogen water ~ h e r n ~ s ~ y ti w e

Chemistry ( w ~ / ~ o b w h ~ ~

crack Q

~

M I

from BWRVIP-14 are vafid and can be utilized.

Chart 1: EPRl BWRViA V2.0, H2:02 Molar Ratio vs FW Hydrogen Concsnrration Dresden Unit 2 BP-km3c MP-IR*

Dc-W*

DC-LO*

RC-END+-

Jp-WT+-

W-TOP+

LP-BOTc

? 5 t

0 5 5 (I

4 43 i *:

1 2 4 a

• 6 1

0 5 0

0 D M 0 1 0 t.3 9 1 0 15 0 3 Fsedwaor W (ppra)

@ Structural lnfegrify Associates, Inc.

APPENDIX D DRESDEN TRANSMITTAL OF DESIGN INPUT (TODI)09-048, "VALVE BODY MATERIAL PROPERTIES FOR VALVE 2-0202-4B," DATED NOVEMBER 23,2009 File No.: 0901309-303 Page D-1 of D-2 Revision: 0

@ Sfru~fwaI lntegrily Associates, Inc.

Ferrite Numtler rnfsasurements weti3 taken at 12 lacatkms cm the Outside surface d the body of RscircttI~M System Valve 02024 orr 11/20-.

The measwad values ranged f m

13 to 17. The average vaiue wa8 The valve M y

materid b

-A351 Grade CFBM. The C a b n Content of the wta? h d y material was reported to be Se0.08% by the hkat~rl&l CMWimUbn Strat. referencad Wowr, File No.: 0901309-303 Page D-2 of D-2 Revision: 0