ML090641017

License Amendment Request Updating Leak-Before-Break Evaluation

ML090641017
Person / Time
Site:	Catawba
Issue date:	02/26/2009
From:	Pitesa J Duke Energy Carolinas, Duke Energy Corp
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML090641017 (19)

Text

Duke DUKE ENERGY CAROLINAS, LLC Catawba Nuclear Station Energye 4800 Concord Road Carolinas York, SC 29745 February 26, 2009 U.S. Nuclear Regulatory Commission Attention:

Document Control Desk Washington, DC 20555-0001

Subject:

Duke Energy Carolinas, LLC (Duke)

Catawba Nuclear Station Unit 2, Docket Number 50-414 License Amendment Request Updating Leak-Before-Break Evaluation

Reference:

Letter from Duke to NRC, same subject, dated November 20, 2008 The reference letter submitted a License Amendment Request (LAR) to update the Leak-Before-Break (LBB) evaluation for Catawba Nuclear Station Unit 2.

This LAR was submitted in conjunction with Duke's proposal to apply Full Structural Weld Overlays (FSWOLs) to the reactor vessel hot leg nozzle-to-safe end welds in the upcoming Spring 2009 refueling outage.

On February 13, 2009, the NRC electronically transmitted 14 Requests for Additional Information (RAIs).

These RAIs were discussed in a telephone conference call between Duke and the NRC on February 19, 2009.

The purpose of this letter is to respond to these RAIs.

The attachment to this letter contains.

our RAI responses.

The format of the attachment is to restate each RAI question, followed by our response.

The original supporting regulatory and environmental analyses contained in the reference letter are unchanged as a result of these RAI responses.

Pursuant to 10 CFR 50.91, a copy of this letter and its attachment has been forwarded to the appropriate State of South Carolina official.

There are no regulatory commitments contained herein.

www. duke-energy. com

U.S. Nuclear Regulatory Commission Page 2 February 26, 2009 If you have any questions or need additional information on this matter, please contact L.J. Rudy at (803) 701-3084.

Very truly yours, John W. Pitesa Attachment

U.S. Nuclear Regulatory Commission Page 3 February 26, 2009 John W. Pitesa affirms that he is the person who subscribed his name to the foregoing statement, and that all the matters and facts set forth herein are true and correct to the best of his knowledge.

John W. Pitesa, Station Manager Subscribed and sworn to me:

LU t"

Notaj Public v

Date My commission expires:

Date SEAL

U.S. Nuclear Regulatory Commission Page 4 February 26, 2009 xc (with attachment):

L.A. Reyes U.S. Nuclear Regulatory Commission Regional Administrator, Region II Atlanta Federal Center 61 Forsyth St.,

SW, Suite 23T85

Atlanta, GA 30303 A.T. Sabisch Senior Resident Inspector (CNS)

U.S. Nuclear Regulatory Commission Catawba Nuclear Station J.H. Thompson (addressee only)

NRC Project Manager (CNS)

U.S. Nuclear Regulatory Commission One White Flint North, Mail Stop 8-G9A 11555 Rockville Pike Rockville, MD 20852-2738 S.E. Jenkins Section Manager Division of Waste Management South Carolina Department of Health and Environmental Control 2600 Bull St.

Columbia, SC 29201

U.S. Nuclear Regulatory Commission Page 5 February 26, 2009 bxc (with attachment):

R.D. Hart (CN01RC)

L.J. Rudy (CN01RC)

W.O.

Callaway (CN03SE)

T.L. Bradley (CN04MD)

M.R.

Robinson (EC07C)

J.M.

Shuping (EC07C)

D.H. Llewellyn (EC09E)

R.L. Gill, Jr.

(EC050)

NCMPA-1 NCEMC PMPA Document Control File 801.01 RGC File ELL-EC050

February 19, 2009

Subject:

Delegation of Signature, NRC Correspondence Nuclear System Directive (NSD) 227, Communicating with the U.S. Nuclear Regulatory Commission, provides guidance on Executive Signature for written NRC correspondence in section 227.9.6. Specifically, it states the following:

Direct reports (Managers) to Vice Presidents at the Sites and General Office may sign submittals that require Oath or Affirmation provided such delegation of signature authority has been provided in writing by the above officers in advance.

This letter delegates the signature authority to the Station Manager for Catawba Nuclear Station correspondence that involves license applications, annual UFSAR updates, responses to requests for additional information, responses to Notices of Violations or Orders, other responses such as NRC Bulletins or Generic Letters and affidavits.

Ja RZis Vice President Catawba Nuclear Station xc:

Document Control File 801.01 RGC Date File

ATTACHMENT RESPONSE TO REQUESTS FOR ADDITIONAL INFORMATION (RAIs)

REQUEST FOR ADDITIONAL INFORMATION LEAK BEFORE BREAK ANALYSIS OF HOT LEG NOZZLE WELD OVERLAYS CATAWBA NUCLEAR STATION, UNIT 2 DUKE ENERGY DOCKET NUMBER NO.

50-414 By letter dated November 20,

2008, Duke Energy submitted for NRC review and approval a license amendment request (LAR).

This LAR was submitted to update the leak-before-break (LBB) evaluation as part of weld overlay installation on hot leg nozzles at Catawba Nuclear Station, Unit 2, during the spring 2009 refueling outage.

To complete its review of the Catawba LBB analysis, the staff requests the following information.,

Leak-before-Break Evaluation, to the November 20, 2008 submittal

1. Leakage Crack Size Analyses: In paragraph 5.d in Section 5.2 (Leakage Evaluation) it states: "Crack roughness is taken as 0.000197 inches for fatigue crack growth in materials other than the Alloy 600 weld.

There are no turning losses assumed for fatigue cracking."

For the Alloy 600 weld, paragraph 7 of Section 5.2 indicates that the crack morphology properties for the PWSCC-susceptible Alloy 82/182 material were taken from Reference 22 (Rudland, et. al.).

Paragraph 6 of Section 5.2 states that "For the weld with Alloy 82/182 material, the adverse effects of PWSCC crack morphology will be considered for the affected material; for other material, the crack morphology for fatigue cracking will be used."

NRC staff believes that the crack morphology properties from Rudland (Ref.

22) used for the PWSCC-susceptible Alloy 82/18.2 material' are appropriate, but the fatigue crack morphology properties quoted in paragraph S.d are significantly lower than values reported in NUREG/CR-6004 and. numerous prior Westinghouse LBB submittals.

The lower roughness and number of turns used in the analysis will result in a shorter postulated leakage crack size, hence, the margin between critical crack length and leakage crack length would be overstated with respect to an analysis where one used the NUREG/CR-6004 values.

(a) What is the justification for the use of the crack morphology parameters, e.g. roughness values and number of turns, used in this leak rate analysis?

(b) How were the PWSCC and fatigue 1

properties combined to yield a single set of composite crack morphology parameters?

Duke Response:

The crack morphology used in the evaluation was verified by early EPRI/Battelle work to justify the PICEP computer program

[Ref.B-6 of the Catawba Unit 2 weld overlay LBB evaluation report].

The values used provided a good mean prediction of leakage for a large number of tests.

The later work referenced in NUREG/CR-6004 was based on microscopic measurements to quantify flow path morphology and is the basis of a modified model that is reasonable for predicting leakage.

However, it has not been verified against similar fluid mechanics test data.

The uncertainties of the fluid mechanics leakage analysis were recognized by the initial developers of the LBB technology documented in NUREG-1061, Volume 3, where a factor of 10 on leakage was imposed to cover a multitude of uncertainties between the predictions and the plant detectable leakage rate.

In performing the LBB evaluation for the Catawba weld overlays, analysis was conducted to assure that the leakage prediction parameters and methodology would be conservative relative to the licensed LBB submittal that had been approved by the NRC staff, as reported in Section 2.0 of the Catawba Unit 2 weld overlay LBB evaluation report.

PICEP was used using the cold leg fluid and loads reported therein.

The predicted leakage with PICEP, using the crack morphology stated in the Catawba Unit 2 report was about 2/3 of that reported for the licensed LBB submittal.

Thus, it was judged to be conservative.

In addition, as requested by the staff in the phone call on February 19, 2009 regarding this Request for Additional Information, it was agreed to evaluate the corrosion fatigue crack morphology parameters (conservatively using the carbon steel parameters since no corrosion fatigue data were available for stainless steel) presented in NUREG/CR-6004:

Roughness from Table 3-4 of NUREG/CR-6004:

- Local Roughness = 347 x 10-6 inches

- Global Roughness = 1,595 x 10-6 inches Turns from Table 3-6 of NUREG/CR-6004:

- Number of turns per inch = 171 Flow Path Multiplier from Table 3-8 of NUREG/CR-6004:

KG+L = 1.06 KG = 1.017 2

The effect of the modified morphology including corrosion fatigue for the welds is shown in Tables 1 and 2.

The leakage analysis including corrosion fatigue parameters in place of the fatigue crack parameters used in the submitted LBB evaluation report indicates that the predicted leakage at the weld overlay locations remains above (with significant margin) the originally submitted LBB leakrate.

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Table 1:

Leakage for Weld 1 (DMW, Nozzle to SE)'

Case Crack 1/2 Critical Leakage2 Pressure Crack

@ 2000 F, Size, in GPM Minimum No 26.48 108.6 Thickness (96.3)

FSWOL, 0.875 in Maximum No 28.61 91.7 Thickness (80.3)

Yes 27.75 83.3

FSWOL, 1.35 in (75.0)

No Overlay No 21.40 112.8 (NA3)

Notes:

1. See Table 5-3 of the Catawba Unit 2 weld overlay LBB evaluation report

2.

Numbers in parentheses are leakage evaluations using corrosion fatigue morphology

3. Leakage unaffected since complete flow path is PWSCC Table 2:

Leakage for Weld 2 (SSW, SE to Pipe)'

Case Crack 1/2 Critical Leakage 2 Pressure Crack

@ 200°F, Size, in GPM Minimum No 21.08 176.0 Thickness (54.2)

FSWOL, 0.837 in Maximum No 24.74 209.5 Thickness (61.4)

FSWOL, 1.35 in No Overlay No 11.51 63.6 (22.7)

Notes:

1. See Table 5-4 of the Catawba Unit 2 weld overlay LBB evaluation report

2.

Numbers in parentheses are leakage evaluations using corrosion fatigue morphology Regarding the methodology for combining PWSCC and fatigue properties to yield a single set of composite crack morphology parameters (as requested in part b of the question),

the 4

methodology is described in Appendix B, Section B.3 of the Catawba Unit.2 weld overlay LBB evaluation report.

2. Margins: Tables 5-3 and 5-4.report leakage through a crack which is half of the critical crack length.

One interpretation of these tables is that the ratio between reported leakage (through half of the critical crack length) and leakage detection limit (with safety factor of 10) satisfies the margin on crack size of 2.0 from SRP 3.6.3 if the ratio is greater than 2.0.

In Table 5-3, these ratios are in range of -8.3 to -11.3 for the reactor vessel hot leg DMW location.

However, Tables 5-3 and 5-4 indicate that reported leak rates were calculated at 200'F (not 617'F which is the normal operating temperature).

At 200'F, the fluid inside the pipe will be single phase liquid and not two-phase as would be the case at operating temperature (617 0 F).

What is the justification for providing leak rates calculated at 200OF and not 617°F since the resultant leakage at 617'F will be significantly less than at 200'F such that margins would be less at 617'F?

Duke Response:

The leakage rates were calculated based on the pressure and temperature conditions of the hot leg at the operating pressure and temperature of 618°F and 2250 psia.

The fluid mechanics model of PICEP calculates the two phase flow mass flow rate.

The output from PICEP converts the mass flow rate (lb/sec) to a volumetric flow rate (gpm) using the density of water at 200 0 F, as reported.

3. Critical Crack Size Analyses: The added mass of the overlay may be substantial, especially for larger piping systems (hot leg).

What is the effect of the added mass of the overlay on the critical crack size caldulations. using normal and safe shutdown earthquake (SSE) loads?

Duke Response:

The effect of the weld overlay mass is insignificant with respect to the seismic analysis for the reactor coolant system (RCS),

which is a composite model of the reactor pressure vessel and internals, piping, reactor coolant pumps, steam generators and contained water.

An evaluation showed that the added mass is only about 2 percent of the total mass of the hot leg.

The seismic forces and moments in the RCS piping are mainly due to the interaction and relative movement between the steam 5

generators and the reactor pressure vessel, not due to the inertial forces of the piping excitation.

In addition, the weld overlay is applied on the nozzle of the vessel where it is rigidly attached.

The added stiffness of the piping at this location is not significant compared to the overall stiffness of the composite piping/structural model.

Based on the limited change in piping stiffness combined with the small relative increase in mass, the mass effects on the RCS seismic response and the critical crack size calculations are considered negligible.

4. Residual Stress Analyses: Figure 3-3 indicates that the hydro-test reduced the hoop residual stresses significantly (-40 ksi hoop stress reduction).

This is surprising for such a thick pipe and runs contrary to other analyses.

What were the boundary conditions for the hydro-test modeling, e.g.

pressure, and the sequence of when it was applied, e.g. post ID repair, after the post weld heat treatment, or post butt weld?

Duke Response:

This question was withdrawn based on the February 19, 2009 teleconference with the NRC staff.

5. Residual Stress Analyses: Axial shrinkage of the overlay can cause a tensile axial stress in the rest of the system when the weld overlay is in situ with the pipe system connected to the vessel and steam generator.

This shrinkage should result in slightly different residual stresses in. the WOL.

How was this accounted for in the analyses?

Duke Response:

A piping model of a typical Westinghouse PWR reactor coolant loop was developed based on the McGuire Unit 1 loop geometry.

The McGuire Unit 1 and Catawba Unit 2 RCS loop geometries are similar with only minor geometrical differences.

The model includes the hot leg, cold leg, crossover pipe, steam generator, reactor coolant pump and reactor pressure vessel outlet and inlet nozzles.

Steam generator and reactor pressure vessel support stiffnesses, locations, weights and centers of gravity were modeled.

The McGuire Unit 1 RCS loop is essentially the same as Catawba Unit 2, and the results are considered applicable.

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Weld overlay shrinkage and rotations were input to the model at the reactor pressure vessel hot leg nozzle weld overlay location.

This analysis was used, along with field and mockup weld overlay shrinkage data, to evaluate the effects on weld overlay shrinkage and rotation for large bore weld overlay applications.

These effects were demonstrated to produce acceptable stresses and displacements in the system, including at the weld overlay location.

Measured weld overlay shrinkages on large bore field readiness mockups performed for the Catawba Unit 2 weld overlay project were found to be smaller than evaluated.

6. Residual Stress Analyses: The PWHT is modeled via a creep analysis during a hold time which typically reduces stresses due to a creep relaxation effect.

What is the technical basis for the increased weld residual stresses from Figure 3-2 after the post weld heat treatment process which seems atypical to conventional understanding?

Duke Response:

This question was withdrawn based on the February 19, 2009 teleconference with the NRC staff.

7. Residual Stress Analyses: Numerous research studies demonstrate that the choice of an appropriate hardening law can make a large difference in residual stress model results.

What is the justification for using the kinematic hardening law and where was it applied, i.e. the original weld, repair, and overlay simulations?

Duke Response:

The residual stress analysis methodology and choice of material hardening law for large bore weld overlays is based on an analysis of residual stresses in the EPRI 36" diameter mockup1 '

and comparison of those results to strain gauge and X-ray diffraction measurements on the inside surface of that mockup.

Numerous analyses were performed on the mockup, using different strain hardening, weld heat input and heat transfer boundary conditions.

For these conditions, an optimum set of analysis assumptions was obtained to provide the best overall agreement with the measured residual stresses.

These assumptions include

'Ref. "Update on OWOLs", Presented by: Pete Riccardella at Industry Briefing to NRC on PWSCC Mitigation Research, Rockville, MD, July 16, 2008 7

a multi-linear isotropic (MISO) strain hardening assumption (not kinematic hardening).

The comparison with measured results under these assumptions indicates that the analysis reasonably predicts the hoop residual stress benefits and conservatively underpredicts the axial residual stress benefits of the overlay.

This same set of assumptions was used in the Catawba Unit 2 residual stress analyses.

8. Residual Stress Analyses: Water backing effects (cooling during the weld overlay laydown) and weld overlay sequencing may affect the weld residual stress model results.

How was water cooling and weld sequencing analyzed, or if not, what was the justification for not considering the effects of these conditions on the modeled residual stresses?

Duke Response:

During the weld overlay process, a heat transfer boundary condition of 100 Btu/hr-ft2' F at 70°F bulk temperature is assumed at the inside surface of the nozzle to simulate water backing.

Weld sequencing is modeled to closely simulate the expected field welding process, including temper bead and stainless steel buffer layers.

Controls are placed on field welding such that these sequence assumptions are closely adhered to and monitored.

9. Residual Stress Analyses: The submittal provided axi-symmetric analyses for the weld overlay which has been shown to produce conservative results at the inside diameter but not necessarily at the outside diameter.

What effect would a 3-dimensional analysis have on the conclusions of this submittal?

Duke Response:

3-D analyses were also performed of the EPRI 36" diameter mockup referred to under question 7 above.

The 3-D analysis yielded less underprediction in the axial stress predictions when compared to measurements, but also resulted in a significant overprediction of the hoop residual stress benefits of the overlay.

Based on the EPRI 36" diameter mockup studies, as well as other mockup studies and field experience (documented in MRP-169, Rev.

1), 2-D axisymmetric analyses are considered to produce reasonable and conservative conclusions relative to the residual stress benefits of weld overlays.

8

10. Residual Stress Analyses: Volumetric expansion may occur during cool down due to a potential phase change.

This expansion tends to reduce the magnitude of the tensile stresses in the affected area which tend to decrease the compressive stresses along the ID.

This is not considered important for the original weld because of the post weld heat treatment after buttering.

However, this phenomenon may be more important for a weld overlay interfacing with the ferritic nozzle.

Please address how the phase transformation plasticity effects are included in the weld overlay finite element model for the nozzle material.

Duke Response:

Volumetric expansion effects are not addressed in the residual stress analysis methodology.

However, there have been numerous experimental studies that have evaluated residual stress benefits of weld overlays, including (most recently) the EPRI 36" diameter mockup study referred to under question 7 above.

These studies, as well as extensive field experience with weld overlays in BWRs (including nozzles up to 28" in diameter) all confirm that weld overlays produce effective residual stress benefits, regardless of the potential effects of volumetric expansion.

11.

Margins: Page 3-9 states: "In this evaluation, plant design transients are utilized, assuming that the defined number of cycles for 40 years, multiplied by a factor 1.5 to conservatively define cycles for 60 years of operation...".

Also, page 4-4, top, states: "It is conservatively assumed that this weld was fabricated by flux welding with either submerge arc welding (SAW) or shielded metal arc welding (SMAW) and as such a Z factor has to be considered."

What is the justification for describing both of the above cases as "conservative"?

Duke Response:

The cycle assumption is considered conservative because the number of design cycles used in the crack growth evaluation is 1.5 times the number used in the extended 60 year plant life.

This increased number of design transients is prorated over the remaining life of the plant (38 years) to further establish a conservative occurrence rate for the crack growth evaluation.

Consistent with the requirements in SRP 3.6.3, the assumption of SMAW or SAW requires the use of a Z factor and consideration of 9

thermal expansion stresses.

This assumption is more conservative than assuming that the welds were created using a TIG process where SRP 3.6.3 would not require the Z factor or consideration of thermal expansion stresses for critical flaw sizing.

12.

In the middle of page 3-10, it is stated that the time it takes for an initial flaw of 75% of the original base metal thickness to reach 100% of the original base metal is reported.

Discuss the time for the initial flaw size to reach the final flaw size.

Discuss where the time period is reported.

Duke Response:

This question was withdrawn based on the February 19, 2009 teleconference with the NRC staff.

13.

Discuss the processes that were used to model the temper beads for the beginning of the weld overlay passes.

Duke Response:

As noted under question 8 above, the temper bead layers, as well as the stainless steel buffer layer, were specifically modeled in the weld overlay residual stress analyses, using weld heat input parameters (e.g., volts, amps and travel speed) provided by the weld process specification., License Evaluation, to the November 20, 2008 submittal

14.

Identify the differences in analysis method, assumptions, and parameters used between the original LBB calculations that the NRC approved and the current LBB calculations.

For each of the differences discuss its impact on various safety margins as specified in SRP 3.6.3.

Duke Response:

The methodology for evaluation of weld overlays is new and not addressed in SRP 3.6.3, so it is not possible to compare each difference related to the various safety margins.

However, the evaluation was intended to be conservative relative to all aspects of critical flaw sizing and leakage prediction using SRP 3.6.3 as guidance.

The following provides a comparison of the 10

Catawba Unit 2 weld overlay LBB evaluation with the original NRC approved LBB evaluation.

1. Location Evaluated

a. Original NRC approved LBB evaluation:

Moments and forces evaluated for all locations.

The LBB evaluation was based on the most critical location (in the crossover leg piping where Normal + SSE loads were the highest).

b. Current weld overlay evaluation:

The weld overlay applied at hot leg nozzle was evaluated.

To show effect on LBB, this specific location was evaluated for both the underlying dissimilar metal weld and safe end-to-piping weld.

==

c. Conclusion:==

The evaluation for the original analysis location is still applicable.

The revised evaluation was to specifically evaluate the effect of adding the weld overlay.

2. Piping Forces and Moments

a. Original NRC approved LBB evaluation:

The loads were based on the Westinghouse RCS loop analysis.

b. Current weld overlay evaluation:

Moments and forces were from the same analysis as for the original LBB evaluation.

==

c. Conclusion:==

No change.

3. Critical Flaw Sizing Methodology

a. Original NRC approved LBB evaluation:

Critical flaw sizes were determined based on both limit load and elastic plastic fracture mechanics (EPFM) methods considering thermally aged cast stainless steel.

b. Current weld overlay evaluation:

Evaluations were based on modified limit load methodology with Z factor to conservatively bound the original EPFM analysis.

Methods for considerations of weld overlay are new technology and are described in Appendix A of the Catawba Unit 2 weld overlay LBB evaluation report.

==

c. Conclusion:==

Details for evaluating critical flaw sizes as described in the Catawba Unit 2 weld overlay LBB evaluation report are consistent with the original evaluations and SRP 3.6.3 guidance.

4. Leakage Prediction Methodology

a. Original NRC approved LBB evaluation:

A Westinghouse proprietary model, with results documented in Section 2 of the Catawba Unit 2 weld overlay LBB evaluation 11

report, was used.

The specific crack morphology was not provided.

b. Current weld overlay evaluation:

PICEP was used, using crack morphology consistent with that used prior to recent NRC research.

==

c. Conclusion:==

Comparison to the results in the licensed LBB evaluation showed that the leakage presented for the critical LBB location was about 2/3 of that presented by Westinghouse.

Thus, the current evaluation is conservative.

5. Results of Evaluation

a. Original NRC approved LBB evaluation:

The original LBB submittal was approved based on an assessment of margins at the critical reactor coolant loop location related to applied loads, critical crack size and leak rate determinations in WCAP 10546.

Section 2.0 of the submitted report presents the results of the original LBB evaluation in terms of critical flaw size and leak rates.

b. Current weld overlay evaluation:

The thermal expansion stresses at the hot leg location where the weld overlay is installed are currently much greater than that for the critical crossover leg location.

The leakage predicted at the hot leg location is considerably greater than 10 gpm with or without the weld overlay.

==

c. Conclusion:==

The LBB evaluation shows that the weld overlay locations leakage margins are considerably higher than were approved in the licensed LBB evaluation.

This comparison shows that the margin between predicted leakage and the plant leakage detection system is greater than a factor of 80 for the weld overlay locations, as compared to the SRP 3.6.3 guideline of 10.

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