ML101260060

ASME Code Section XI Flaw Evaluation of Dissimilar Metal Weld Flaws Identified by Ultrasonic Testing During Unit 1 2010 Refueling Outage

ML101260060
Person / Time
Site:	Calvert Cliffs
Issue date:	05/03/2010
From:	John Stanley Calvert Cliffs, Constellation Energy Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML101260060 (24)

Text

Calvert Cliffs Nuclear Power Plant 1650 Calvert Cliffs Parkway Lusby, Maryland 20657 CENG a joint venture of A% Constellation wEnergy-0T D

CALVERT CLIFFS NUCLEAR POWER PLANT May 3, 2010 U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION:

Document Control Desk

SUBJECT:

Calvert Cliffs Nuclear Power Plant Unit No. 1; Docket No. 50-317 ASME Code Section XI Flaw Evaluation of Dissimilar Metal Weld Flaws Identified by Ultrasonic Testing During Unit 1 2010 Refueling Outage During the Unit 1 2010 refueling outage, Calvert Cliffs Nuclear Power Plant, LLC, conducted ultrasonic examinations on Reactor Coolant System pressure boundary dissimilar metal welds.

The ultrasonic examination of pressurizer safety relief valve (IRV-201), 4 inch nominal pipe size nozzle to safe end, weld number 4-SR-1006-l, detected two embedded indications that were sized using ultrasonic examination techniques and found to exceed the acceptance standards of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, 2004 Edition no addenda Subsection IWB, Sub-article IWB-3500.

An ASME Boiler and Pressure Vessel Code,Section XI, Sub-article IWB 3600 analytical evaluation was conducted to evaluate the flaws.

The analytical evaluation concluded that the flaws are acceptable to be left in the as-found condition for the remainder of Unit 1 licensed life. As required by ASME Boiler and Pressure Vessel Code,Section XI, 2004 Edition no Addenda, Sub-section IWB, Sub-paragraph IWB 3134(b), the analytical evaluation is to be provided to the regulatory authority and is provided in Attachment (1).

X0V\\7

Document Control Desk May 3, 2010 Page 2 Should you have questions regarding this matter, please contact Mr. Douglas E. Lauver at (410) 495-5219.

JJS/KLG/bjd

Attachment:

(1).

Pressurizer Relief Valve IRV-201 Nozzle Flaw Evaluation for Remainder of Licensed Life (CA073 51) cc:

D. V. Pickett, NRC S. J. Collins, NRC Resident Inspector, NRC S. Gray, DNR

ATTACHMENT (1)

'3 PRESSURIZER RELIEF VALVE 1RV-201 NOZZLE FLAW EVALUATION FOR REMAINDER OF LICENSED LIFE (CA07351)

Calvert Cliffs Nuclear Power Plant, LLC May 3, 2010

Structural Integrity Associates, Inc.

File No.: 0801014.372 CALCULATION PACKAGE Project No.: 0801014 Quality Program: Z Nuclear D Commercial PROJECT NAME:

CCNPP Alloy 600 DMW NDE and Contingency Overlay Repair CONTRACT NO.:

Constellation Energy Purchase Order No. 428505, Rev 1.

CLIENT:

PLANT:

Constellation Energy

'Calvert Cli~ffs Nuclear Power Plant, Unit I CALCULATION TITLE:.

Pressurizer Relief Valve I RV-20i Nozzle Flaw. Evaluation for Remainder of L[iensed Life Document Affected Project Manager Preparer(s) &

Revision Pages Revision Description Approval Checker(s)

RPSignature & Date Signatures & Date 0I

- 21 Initial Issue Michael Isle 03/23/10 D. V. Sommerville 03/23/10 C. Fourcade 03/23/10 Page 1 of 21 F0306-01 RO

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Table of Contents 1.0 2.0 3,0 4.0 5.0 6.0 7.0 8.0 O B JE C T IV E....................................

F.................

3 M E T H O D O LO G Y..................................................................................................

3 A SSU M PT IO N S......................................................................................................

4 D E SIG N IN PU T S.....................................................................................................

7 CALCULATIONS..............................................

10 RESULTS OF ANALYSIS.......................................

20 CONCLUSIONS AND DISCUSSIONS...............................................................

20 R EFER EN C E S.......................................................................................................

20 List of Tables Table 1: Summary of RV Nozzle Circumferential Indications.

8 Table 2: Summary of RV Load Conditions and Cycles.......................................................

9 Table 3: Piping Forces and Moments for RV Flaw Evaluation...................... I.................... 10 Table 4: Summary of Load Conditions Evaluated for RV Flaw Evaluation.......................

12 Table 5: FCG Calculation Sum m ary..................................................................................

17 Table 6: Allowable Flaw Sizes for RV Circumferential Indication..................................

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1.0 OBJECTIVE This calculation package documents an American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI,IWB-3600 [1b]fflaw evaluation of the two circumferential indications identified in the Calvert Cliffs Nuclear Power!Plant (CCNPP) Unit I Pressurizer Relief Valve (RV) I RV-201 Nozzle. The calculation is performed to assess whether the indications in the RV nozzle are acceptable, as-is, for the remainder of the plant licensed life (30 years).

2.0 METHODOLOGY This section presents the methodology used to perform the analysis.

I.

Flaw Characterization The two circumferential indications identified in Reference [2] are bounded by a single elliptical circumferential flaw. Both indications in the Unit 1 RV Nozzle dissimilar metal weld (DMW) are categorized as subsurface flaws.

Il.

Load Conditions andCycles The load conditions and number of applicable. cycles given in Reference,[3, pg; 9-10] and Reference

[4, Sht. 10-12] are used'to define the loadconditions and cycles applicable for this flaw evaluation, Ill.

Load Magnitudes

1. Piping Reaction Loads: The primary load piping forces and moments summarized in Reference

[5, Table 3].are used for this evaluation, These loads include seismic and dynamic loads. The applicable secondary loadpiping forces and moments are obtained from Reference [3, pg. 44].

2. Internal Pressure: The internal pressure applicable for each load condition is obtained from Reference [3, pgs. 9, 10, 27, 28, 31, 34]

3. Radial (through-wall) Thermal Gradients: The Pressurizer fluid temperatures given in Reference [3, pgs. 9, 10, 27, 28, 31, 34] are used to define the thermal transient conditions for which a through-wall thermal gradient would exist in the RV Nozzle. This through-wall gradient will contribute a load on the flaw; therefore, it is considered in the flaw evaluation.

IV.

Flaw Evaluation I. Flaw Evaluation Process: The methods for piping flaw evaluations given in Section XI, Non-mandatory Appendix C [Ib] are used for this analysis.

2. Flaw Models: Where appropriate, guidance for linear elastic fracture mechanics (LEFM) subsurface flaw models and radial thermal gradient LEFM models is taken from Section X1, Non-mandatory Appendices A and G [I b], respectively.

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3. Crack Growth Mechanisms: Recognizing that the indications evaluated in this analysis are subsurface flaws, they are not exposed to the pressurized water reactor (PWR) coolant environment and are consequently not.susceptible to pressurized water reactor stress corrosion crack (PWRSCC) growth. Fatigue Crack Growth (FCG) is the only applicable crack growth mechanism for the subsurface flaws.Section XI [1 b] does not currently contain FCG correlations for Alloy 82/182 materials; therefore, the air FCG curves presented in NUREG/CR-6721 [6] are used, for thisevaluation.

4. Crack Growth Analysis: Recognizing that.it~is-not appropriate to evaluate, the incremental growth for30 years of cycles atonce, the crack growth analysis Will be performed in an iterative manner where two years of cycles.:will be considered for eachanalysis iteration. The incremental crack growth for each two year period will be used to update the crack size prior to.

performing the next analysis iteration. This iterative process,will be repeated until 30 years of load cycles have been evaluated.

3.0 ASSUMPTIONS Multiple conservative assumptions are used in order to perform a bounding analysis and to simplify the calculation process. It is intended that there be significant and overlapping methodological conservatisms in the analysis documented in this calculation package. The following assumptions are utilized for this analysis.

1.

Flaw Characterization

1. A single, bounding, subsurface, circumferential flaw is assumed for this analysis. The length and depth of the flaw is defined such that it bounds the two circumferential indications reported in Reference [2]. The flaw size is defined by using the deepest dimension of the two flaws and

'the shallowest dimension of the two flaws for the flaw depth, 2a, and using a length, I, which bounds the two flaws. This simplifying assumption is conservative in that it takes no credit for separation between the two flaws or forthe effect that-the smaller sizes associated with the individual, flaws would have on the calculated range of Mode I stress intensity factor (AK1) used for the FCG calculations. The circumferential flaw size used for this analysis is:

2a.= 0.423 inches.

L = 1.500 inches.

II.

Materials I,

The weld butter and weld material joining the RV nozzle and safe end are assumed to be Alloy 82/182 material.

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IMl.

Load Conditions and Cycles I. All load conditions and associated cycles identified in References [3, 4] are bounded by four load conditions considered for this analysis.

a. Load Condition I will bound the pressurizer heatup/cooldown, leak test, and hydrotest events.

b. Load Condition 1I will represent theReactor Trip event.

c. Load Condition Ill will represent the PORV/SRV, OBE, Plant Loading/Unloading, and 10% Step Increase/Decrease events.

d. Load Condition IV will represent the Normal Pressure Variations event.

2. The number of load cycles assumed to occur in each:two yearoperating period is. 2/4.0 of the total number of cycles defined in References [3, 4]. This assumption assumes that all transient cycles are evenly distributed throughout the 40 year design life of the component; therefore, 2/40 of the total cycles areassumed to-occur during each 2 year operating cycle.

Iv.

Load Magnitudes I. The range of pressure for all conditions represented by Load Condition I and 1I is conservatively assumed to be the full hydrotest pressure range of 3125 psig.

2, The range of pressure for all conditions represented by Load Condition III and IV is conservatively assumed to be 200 psig.

3. The temperature gradient in the RV Nozzle for the hydtrotest condition is assumed to be bounded by the 200 "F/hr limit specified for the Cooldown Transient in Reference [3, pg. 27].

4. The radial thermal gradient for all thermal transients represented by Load Condition I and III is assumed to be the 200 *F/hr limit identified in Reference [3, pg. 27] for the Heatup/Cooldown event. This assumption is bounding for all other thermal transients represented by this load condition. The majority of the thermal transients applicable to the RV Nozzle exhibit minor temperature variations which are on the order of 30-40 'F over a. period of time which would yield a thermal transient rateof 50-100 "F/hr; therefore, the value assumed for this load condition is conservative.

5. The radial thermal gradient for the Reactor Trip event, represented by Load Condition.I,, is assumed to be 2900 "F/hr, which is based upon converting the 40 "F temperature drop which occurs over approximately 50 seconds [3, pg. 341 into a ramp rate per hour.

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6. Thetemperature variation for the normal pressure variation event is listedin Reference

[3, pg., 10] as +/- 7 "F; this thermal variation is negligible; :therefore, no radial thermal gradient will be considered for the FCG calculations performed for Load Condition IV.

7., The primary load piping forces and moments applicable for Load Condition I, 1, and HII are taken as the bounding Level C values as summarized in Reference [5, Table 3],

8. The secondary load piping forces and moments applicable for Load Condition I, I, and III are taken as the bounding values of the THERMAL I and THERMAL2 load conditions combined using absolute sum with the seismic anchor motion (SAM) forces and moments, for both RV Nozzles, given in Reference [3, pg. 44].

9. There are no primary load piping forces and moments applicable to Load Condition IV since this load case represents only the normal variation in pressure and temperature during normal operation of the pressurizer. All attached piping loads contributed by seismic and dynamic events are considered in Load Condition I, 1I, and 1i11 Also note that the deadweight piping forces and moments are a mean load which do not affect the range of stress intensity factor, AK1, which is considered for FCG.

10. There are no secondary load piping forces and moments applicable to Load Condition IV since this load case represents only the normal variation in pressure and temperature during normal operation of the pressurizer. All attached piping loads contributed by seismic; dynamic, and thermal events are:considered in Load Condition 1;1I, and IlL V.

Flaw Evaluation

1. The DMW is.conservatively assumed to be applied with a shielded metal arc welding (SMAW) process. This assumption will require consideration of secondary stresses in the allowable flaw size assessment; therefore, the Elastic-Plastic flaw evaluation methods of C-6000 flb] are used for the flaw evaluation. This assumption is conservative in that it assumes the material has lower toughness than if a non-flux weld process were used.

2. All loads on the flaw are assumed to be membrane loads; therefore, the membrane stress correction factor for the Appendix A subsurface flaw model [I b] will be applied to both the membrane and bending loads. This assumption is conservative since the bending stress correction factor is always less than the membrane stress correction factor; therefore, this analysis takes no credit for the lower K, that would be predicted by more precisely distinguishing between membrane and bending stresses.

3. The contribution of the radial thermal gradient to the range of Mode I stress intensity factor considered for FCG is conservatively determined using the radial temperature gradient KIT solution given in Section Xl, Non-Mandatory Appendix 0, Paragraph G-2214.3 [lb]. This simplifying assumption is Very conservative since the K1 solution contained in Appendix G considers a 1/4/4 thickness, infinitely longsurface crack in a flat plate. The actual flaw is shallow, File No.: 0801014.372 Page 6 of 21 Revision: 0 F0306-01

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short, and subsurface; therefore, the actual K, anticipated for this flaw is bounded by the Appendix G model used..

4.

The FCG calculation is performed assuming an R-ratio approaching 1.0. This conservative assumption bounds the possible effect of weld residual stress on the mean stresses considered for fatigue. Further, this assumption takes no credit for Mechanical Stress Improvement (MSIP).

5. The FCG calculation is performed assuming a metal temperature of 650 "F for all transients.

This conservative assumption bounds the effectof temperature on the predicted FCG.

4.0 DESIGN INPUTS The following design inputs are used for this analysis:

1.

Geometry: RV Nozzle and DMW geometry is obtained from Reference [7].

a. The nozzle inside diameter (ID) is given as ID 3.875 inches (no credit for butter)

Note: This dimension is taken at the ID of the nozzle base material; there/bre, no credit is taken for the weld butter.

b. The nozzle outside diameter (OD) is given as OD =

6.063 inches.

HI.

Materials: The RV Nozzle, safe end, and DMW materials are obtained from Reference [7].

a.

RV Nozzle material:

SA-508 Class 2

b.

Nozzle butter and weld material:

ENiCrFe-3 (Assumed)

c.

Safe-end material:

SA-182 Type 316 III.

MaterialProperties: The following material properties are obtained from Reference [la] for Alloy 82/182 assuming material properties of the weld metal are equivalent to N06600, SB-166:

a.

Yield Strength, ry.:

29.7 ksi at 650 F

b.

Ultimate Strength, cru:

80.0ksi at 650 'F

c.

Design Stress Intensity, S,,:

23.3 ksi at 650 T File No.: 0801014.372 Revision: 0 Page 7 of 21 F0306-01

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IV, Indication Dimensions: RV Nozzle DMW indications are obtained from Reference [2] and summarized in Table I below.

Table 1: Summary of RV Nozzle Circumferential Indications.

Start End Indication Position(O, Position(),

Length(2" Depth 1(3),

Depth 2(,

2a, in in in in in in

1. 1 0.30 0.60 0.30 0.610 0.780 0.17

1. 2 17.8 0.80 1.5 0.357 0.711 0.354 Bounding 17.8 0.80 1.5 0.357 0.78 0.423 Flaw I.

2.

3.

4.

I his is the linear position oi-me starting pointmot the mnoicaton, measuredcuOckwise arouna ti pipe, starting from top dead center.

The length. dimensions aie taken from Reference [21.

The depth from the pipe OD at~which the uppermost portion of the flaw is detected.

Thedepth from the pipe OD at which the Iowermost (deepest) portion of the flaw is detected.

e cJrcumnerence:oi ei File No.: 0801014.372 Revision: 0 Page 8 of 21 F0306-01

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V.

Load Conditions and Cycles: The applicabtle load conditions and numbers of cycles for Level A, B, C, and test events are obtained from References [3,.4] and are summarized in Table 2 below.

Although. Level C events are not required for FCG calculations' they are considered here for conservatism.

Table 2: Summary of RV Load Conditions and Cycles.,

Load(s)

Pressure, Temperature, Number Reference psi*

T of Cycles PORV/SRV 2485 668 1000

[4, Sht. 11]

OBE 2485 653 650

[4, Sht. 11 ]

Normal Pressure Variation 2235 +/- 100 653 +/- 7 106

[3, pg. 10]

Plant I-Ieatup/CooldownM 1

0-2235 70-653 500

[3, pg. 9, 27]

Plant Loading/Unloading(2) 2200-2300 640-660 15,000

[3, pg. 9, 28]

10% Step Load increase/Decrease(3) 2150-2350 650-660 2,000

[3, pg. 9, 3 1]

Reactor Trip 4) 1700-2300 620-660 400

[3, pg. 9, 34]

Hydrostatic Test 0-3125 100-400 10

[3, pg. 9]

Leak Test 0-2500 100-400 320

[3, pg..9]

I.

2.

3.

4.

5.

i~evlew o1 t~elertinf.

L.', pg. L/j snows tie nHatup.LtransLienti llmi *i *,uiMu ri and tll U

IlIl 1.UolUollw traLns et s*

limited to 200 "F/hr.

The pressure and temperature range is estimated from the plots provided in Reference [3, pg. 28].

The pressure and temperaturerange is estimated from the plots provided inReference [3, pg. 31].

The pressure and temperature range is estimated from theplots provided in Reference [3, pg. 34] for steam.

Level D load conditions are not required for fatigue crack growth.

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VI.

Piping Forces and Moments: The bounding Level A, B, C, and D forces and moments for the two RV Nozzles at CCNP Unit I are obtained from Reference [5, Table 3] and are summarized in Table 3 below. The bounding secondary load condition forces and moments are obtained from Reference

[3, pg. 441 by taking the absolute sum of the seismic anchor motion (SAM) and bounding thermal load case and are summarized in Table 3 below.

Table 3: Piping Forces and Moments for RV Flaw Evaluation.

Condition Fx, V

.)

MRS) lbs Lbs lbs in-lbs in-lbs in-lbs in-lbs Level A")

1,735 156 83 11099 588 17,83.1 17,875 Level BO()

3,542 2,823 2,162 20,527 14,597 40,671 47,839 Level C(3) 4,446 4,156 3,202 30,236 21,604 52,089 63,986 Level D(3) 4,500 4,166 3,218 30,732 22,110 52,742 64,923 Secondary(4'5 )

1748 378 525 9,093 30,564 31,444 44,784 I.

The FY component is defined as the force component acting axially along the RV Nozzle [3, 5).

2.

All moments are combined using Square Root of the Sum of the Squares (SRSS).

3.

These piping forces and moments reflect only the primary loads.

4.

As identified in Section 3(IV)(8), the component forces and moments are taken, individually, from the maximum of the Thermal I and Thermal 2 load cases for both nozzles listed in Reference [3, pg. 441 and combined using absolute sum with the SAM forces and moments. Consistent with Reference.[5], the forces and moments are assumed to act at the safe-end to piping weld; therefore, the transverse moments (M,, Mj) are increased to account for the additional moment contributed by the shear forces (F,, F.) acting over the 5 inch distance between the safe end to pipe weld and thenozzle to safe-end weld.

5. The moments provided in Reference [3, pg. 44] are listed in ft-lbs; whereas, they are presented in in-lbs in this table.

5.0 CALCULATIONS Theanalysis was performed usingan Excel spreadsheet to facilitate the iterative nature of the calculations.

To clearly communicate the methodology used for this calculation, the first iteration is summarized in'the steps that follow. The results of all iterations are summarized in Section 6.0 below.

1. Definition of Load Conditions:

As. discussed inSection 3.0, Assumptions 111.1 and Ill.2,the loads identified in Table 2 are segregated into four Load Conditions as shown below:

Load Condition 1:

• Plant Ileatup/Cooldown (n=500) o Hydrostatic Trest (n= 10)

• Leak Test (n=320)

Load Condition fi:

• Reactor Trip (n=400)

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Load Condition III:

• PORV/SRV (n= 1000)

• OBE (n=650) o Plant Loading/Unloading (n=l 5,000)

• 10% Step Load Increase/Decrease (n=2,000)

Load Condition IV:

• Normal Pressure Variation (n= I E6)

The number of cycles considered for each two year interval, for Load Condition 1, is:

_ (500 + 10 + 320)2 = 42 cycles 40 The number of cyclesconsidered for each two year interval, fortLoad Condition 1I, is:

Nil 2 y

="(400) 2 = 20 cycles N-1 40 The number of cycles considered for each two year interval, for Load Condition [Il, is:

N111 years (1000 + 650 + 15000 + 2000) 2-933 2 1000 4ycles 40 The number of cycles considered for each two year interval, for Load Condition IV, is:

N 2,,

-(E6).

2 = 50,000 cycles 40 t<9 rage ii 01 ~I File No.: 0801014.372 Revision: 0 t-age I I of z I F0306-01

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Table 4 summarizes the.pressure range, thermal transientucondition, and piping forces and moments applied for each Load Condition..

Table 4: Summary of LoadConditions Evaluated for RV:Flaw Evaluation.

(2)

Pressure Thermal Condition

Range, Transient Cycles lbs in-lbs Tsin Condition 1 4,156 + 378 = 4,534 63,986+44,784=108,770 0-3125 200 "F/hr 42 Condition 11 4,156 + 378 = 4,534 63,986+44,784=108,770 0-3125 2900 "F/hrt 3' 20 Condition II1 4,156 + 378 = 4,534 63,986+44,784= 108,770 200 200 *F/hr 1000 Condition IV 0

0 200 0

50,000 I.

2, 3.

The FY component is defined as the sum of the bounding Level C and bounding thermal piping force.

The MsRss component is defined as the sum of the bounding Level C and bounding secondary SRSS moments.

As identified in Section 3(IV)(5), the conservative ramp rate used for Load Condition II is calculated by converting the temperature transient defined for the pressurizer fluid for the Reactor Trip event [3, pg. 34] into an equivalent ramp rate in "F/hr, This is calculated as follows: 40 "F/50 sec. (3600 sec/hr) = 2880 F/hr. 2900 "/hr is conservatively used.

1I.

Calculation of Applied Stresses:

a. The pipe area available to react to an axial load is:

A =

dr(

, jd

= { 6.063-*8 IM 17;.07 in-

b. The pipe section modulus is:

d "4-d 4 7r6.03'

- 3.875' JiL d,,

3 6.063

=18.22 in'

c. The Level C load combination is used for this evaluation since review of the bounding Level C forces and moments combined with the conservative assumptions for internal pressure range and thermal gradient make this the bounding load condition. The structural factor for Level C events is 1.8 for membrane loads and 1.6 for bending loads. A SF = 1.8 is conservatively used throughout the calculation.

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d. The membrane stress at the crack section resulting from internal pressure and primary+ secondary

. piping loads is:

L.C.I:

1:

W.

Ii L.C. !:

or.., =SF P,, 2 i

.2 d,,2 - d d, 2 P.

d2 2

d1 -di d,, 2 -di 2 Fyl (

.3.8752 4534=

+-l 1i8

3125,

+

34364 psi 6.0632

-3.8752 17.07)

+F

=1.8(3127.

38752

+ 4

=4364 psi A

• 6.0632 -187 17.07 F

3.'8752 4534

-`ý~~~~

~

~

~ )=

.(0._

2'--727 psi A-J' 820 6.0.632 -3.875 17.07)=

L.C. II1:

ICIII = SF(t LC, IV:

- C SF( p.

1. 200.

3.8752 249 psi do2 -d, )

6.063' - 3.875 )

e. The bending stress at the crack section resulting from the primary + secondary piping loads is:

L.C. 1, 11, 111:

(M S

(.8108770 5F =F 1 8( 1821= 10746 psi z18.22 L.C.LIV:

7bSF(ýý

=j (L82 0;0ps Ill.

Calculation of AK:

a. The AKI contribution from the through-wall thermal gradient considered for each Load Condition is calculated using the method ofG-2214.3 [Ib]:

Ki,= sF. (0.95.

• 3

.CR.5 )kshfin Where:

CR is the cooldown rate, "F/hr t

is the wall thickness, excluding butter, inches Note that the structural factor is applied to the KI solution above.

Thus, L.C. I:

L.C. I!:

L.C. IIl:

L.C. IV:

Kit 0.43.kvsiVn Kit 0.43 kysi-./i KI,,

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b. The AKI from the loading considered for each Load Condition is calculated using the embedded flaw solution contained in A-33 10 [1 b]:

x,~

A + a,., A4,

+

M).*

Where:

am is the membrane stress, psi Mm is the membrane stress correction factor, unitless orb is the bending stress, psi Mb is the bending stress correction factor, unitless a

is the flaw half depth, inches Q

is the shape parameter given below Q, I+ 4.593 Where::

I is the flawlength, inches qy is the plastic zone size correction factor given below q o,, M, + Ch.M )]2 q

93's 6

Where:

Crys is the yield strength of the material, psi Note that the plastic zone size correction factor and shape factor will be different for each load condition since the stresses resulting from the applied loads are different. For this analysis, the plastic zone size correction factor for Load Condition I will be conservatively used for all load conditions.

Using Figure A-33 10-I with a 2a/t=0.423/1.09=0.388, and a 2e/t=2(0.0265)/l.09=0.049, the membrane stress correction factor is, Mm=.1 5. Note that e is the flaw eccentricity defined consistent with Figure A-3300-1 [1 b].

Recall thatboth the membraneand:bending stress will be conservatively treated as a membrane stress; therefore, L.C. 1:

K, =13,36 ksifin L.C.1I:

K, = 13.36 ksi'7in L.C, III:

K, = 10,.14 k.yivi-in L.C. IV:

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Since the loads are a range of pressure and piping loads from zero load to maximum load, the Ki is the AKI to be considered for FCC..

c. Finally, the total AKI for each load condition is the sum of the contributions from the piping and internal pressure contribution and the~through-wall thermal gradient contribution:

L.C. 1:

AK, = 13.36 + 0.43 = 13.79 ksiVn IL.C. II:

AK/ = 13.36 + 6.17 = 19.53 ksii-n L.C. III:

AK, = 10.14+0.43 = 10.57 ksiV4in L.C. IV:

AKI = 0.22 + 0.00 = 0.22 ksi-jin IV, Calculation of FCG:

a. The FCG equation for Alloy 600 materials provided in Reference [6] is used for this evaluation:

(da)w

-'C,.oo (i -0.82.)

-(&k)'

rn/cycle Where:

CA00 =4,835. 107'4 +1.622. l0-'6 (T)-1.49. I10-18(T2 )+ 4.355. i0o-2 (T3)

T is the temperature inside the pipe, "C (taken as max. during transient)

R is the R-ratio AK is the range of stress. intensity factor, Mpa-m0`5

b.

CA600 will be evaluated at 650 "F (343 'C) for all Load Conditions.

CA600 = 4.835.10-"4 +1.622. 10 16(343)-l.49.I 0-(343')+4.355.10-2(3433)

CA600 = 1.044. 10-13

c. The AKI calculated for all load conditions must be converted to Mpa-m 0 5:

L.C. I:

AK, = 1.099.13.79 = 15.15 MpaVjm L.C. 11:

AK1 = 1.099. i9.53 = 21.46 Mpa-fmi L.C. IIl:

AK, = 1.099.10.57 =1.62 MpaV.r-m L.C. IV:

AK, = 1.099-0.22 = 0.24 Mpa.[/-

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d. The incremental crack growth per cycle is calculated, assuming R-ratio approaching. O

,0:

L.C. I:

d-

=3.14.- lO-mr/,cycle L.C.11:

dNi alp 3.0 i/yl fda m]

L.C. 111:

(

i

.3.Q10~ rn/cycle L.C.

II:

4

.6-cce WNair L.C. IV:

d )

= 1.35:.10-" mr/cycle

e.

The incremental crack growth per cycle is converted to in/cycle:

L.C. 1:

rda

-*.=l24'l0-in/cycle

'~dN ji L.C. 11:

(d 5.16-10-5 in/cycle

~.dN air L.C. Ill:

4.17 1.06 in/cycle (da L.C.IV:

da 5.32 10-"~ in/ cycle

f. The FCG crack growth per tip calculated for the. conservative load conditions defined for this flaw evaluation is:

Aa = 42(1.24. 10-)+ 20(5.16. 0-')+ 1000(4.17.10-4)+ 50000(5.32. 10-")

Aa = 0.0005 in + 0.001 in + 0.0042.in + 2.7.10-4 in =0.0057 in

g. The flaw size after the first two years of crack growth is:

1 = 1.5 + 2(0.0057)= 1.511 in 2a = 0.423 + 2(0.0057) = 0.434 in V.

Summary of FCG Calculations:

Table 5 summarizes the FCG calculations for all evaluation intervals.

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Structural integrity Associates, Inc.

Table 5: FCG Calculation Summary.

Crak Sze>... ______________________________

I.8M Clciaton

______________Calculation______

a, C.

. Ci L

Ak"

k.

1 Akwk AK 1 1 A

Akwv.

Akj"av AK,,*,f, AKIUjaj-AK1.:w,,*XI..L:M, "In n

Qi Ks sKs14nh' Ksl-in0" Ksi-ln°,

K 8 s1 4 M

Ul4l-n Ksi-inoij Ksi-in"s MPa-m' MPa-m°'P MPa-m*

MPa-m0 s 0

0,423 1.S 1,15 0.0571 L1242 13.36 0.43 13.36 6.17 10.14 0.43 0.22 0

15.15 21.46 11.62 0.24 2

0.434 1.511 1.15 0.0571 1.13 13.32 0.43 13.32 6.17 10.12 0.43 0.22 0

15.12 21.42 11.59 0.24 4

0.446 1.523 1,15 0.0571 1.1357 13.29 0.43 13.29 6.17 10.09 0,43 0.22 0

15.08 21.39 11.56 0.24 6

0457 1;534 1.15 0,0571 1.1414 13.26 0.43 13.26 6.17 10.07 0.43

.0.22 0

15.04 21.35 11.54 0.24 8

0.468 1.S45.

L.5 00571 1.1470 13.23 0.43 13.23 6.17 10,04 0.43 0.22 0

15.01 21.32 11.51 0.24 10 0.479 L556 1.15 0.0571 1.1525 13.19 0.43 13.19 6.17 10.02 0.43 0.22 0

14.97 21.28 11.48 0.24 12 0a490 L567

".150 0S71.1.1579 13.16 0.43 13.16 6.17 9.99 0,43 0.22 0

14.94 2:.25 1.-46 0.24 14 0.501 1.578 1.25 0.0674 1.1529 14.34 0.43 14.34 6.17 10,89 0:43 0.24 0

16.23 22-54 12.44 0.26 16 0,516 1.593 1.25 0.0674 1.1603 14.29 0.43 14.29 6.17 10.85 0,43 0.24 0

16.18 22.49 12.40 0.26 18 0.530 1.607 1.25 0.0674 1.1675 14.25 0.43 14.25 6.17 10`82

.0.43 0.23 0

16.13 22.44 12.36 0.26 20 0.545 1,622 1.25 0.0674 1.1746 14.21 0.43 14.21 6.17

.10.79 0.43 0.23 0

16.08 22.39 12.33 0.26 22 0.559 1.636 1.25 0.0674 11816 14.16 0.43 14.16 6.17 10,75 0.43 023 0

16.04 22.35 12.29 0.26.

24 0.574 1.651 125 0.0674 1.1885 14.12 0.43 14.12 6,17 10.72 0.43 0.23 0

15.99 22.30 12.26 0.26 26 0.588 L665 1,25 0.0674 1.1952 14.08 0.43 14.08 6,17 10.69 0.43 0.23 0

15.95 22.26 12-22 0.26 0.602 1.679

1.

0.0674 1.2018 14.04 0.43 14.04 6.17 10.66 0.43 0.23 0

15.91 22.21 12.19 0.25 30 0,615 1.692 K(G.alclidtlon 4~~~

a/d a da/dn..

da/dn~wn v,

N, N1 A,,

A ~

AN

'~

R,.

N, MI N11 N~v rm/cycle

.n/ccfde m/cyde

. ni/cyde in in in in in 1.04428E-13 1

3.14312E-07 1.3094-06 L05803-07 L35017E-14 42 20 1000 5000 0.00 0.00103 0.00417 2.7E-08 0.0057 1.04426E-13 1

3.11118E-07 1.3E-06 1.04738E-07 1.33601E-14 42 20 1000 5(000 0.0006 0.00102 0.00412 2.6E-08 0,0057 1.044281-13 1

3.0815D6-07 1.290836-06 1.01703E.07 L322256-14 42 20 1000 50(4 0.0005 0.00102 0.00408 2.6-08 00056 1.044286-13 1

3.04991E-07 1.28189*-06 1.02696E-07 1`3086M-14 42 20 1008 50010 0.0(85 0.00101 0.00404 2.6t-08 0.0056 1.04428&-13 1

3.02032E-07 1.273186-06 1.01716E-07 1295840-14 42 20 1000 500D 0.005 0.001 0.004 2.6E-08 0.0055 1.04428E-13 1

29919JE-07 1.26468E-06 1.00762E-07 1.283i7E-14 42 20 1000 50000 0.0005 0.001 0.00397 15E-08 0.0055 1,04428E-13 1

2.96407-07 1.25639E-06 9.98334E-08 1i27084E-14

.42 20 1000 50M 0.0005 0.00099 0.00393 2.5E-08 0.0054

• .04428E-13 1

4.16459E-07 1.60039E-06 1.39815E-07 1.86467&14

42.

20 1000 S0)0 0.0007 0.00126 O.OSS 3,6E-08 0.0075

-104428E-13 1

4.112146-07 1.58586E-06 1.3807E-07 1.78127E-14.

42 20 1000

.50000 0.0007 0.00125 0.00544 3.5E-08 0.0074 1.04428E-13 1

4,0616E-07 1.57181E-06 1.363886-07 1.75872E-14

42.

20 1000 50000 0.0007 0.00124 0.00537 3.5-08 0.0073 1.044286-13 1

4.01286E-07 1.55824E-06 1.347676-07 1.73698E-14 42 20 10 5M000 0.0007 0.00323 0.00531 3.46-08 0.0072 1.044286-13 1

3.965636-07 1.5451E-06 1,33202E-07 1,71601E-14 42 20 1000 50000 0.0007 0.00122 0.00524 34E-068 0.0071 1.04428E-13 1

3.92043E-07 1.532396-06 1.31691E-07 1.69577E-14 42 20 1000 50000 0.0006 0.00121 0.00518 3.3-08 0`0070 1.04428E-13 1

3.876596-07 1.52009E-06 1.302328-07 1.67623E-14 42 20 "1000 5000 0.0006 0.0012 0.00513 3.3E-08 0.0070 1.044286-13 1

3,83422E-07 1..50176-06 1.288228-07 1.657358-14 42 20 1000 50O0 0.0006 0.00119 0.00507 3.3E-080.0069 Cumulative:

630 300 15000 750000 0.00881 0.01685 0.07056 4.5607 0.096 File No.: 0801014.372 Revision: 0 Page 17 of 21 F0306-01

Structural Integrity Associates, Inc.

VI.

Allowable Flaw Size Determination:

a.

Since a flux weldprocess was conservatively assumed for this evaluation, a Z~factor is calculated using the equations in C-6330 [Ib]:

Z = 1.3[! + 0.01O(NPS - 4)] =.3[I + 0.0110(6 -4)]= 11326 Where 6 inches is used for the nominal pipe size since the OD of the nozzle is nominally 6 inches,

b.

A stress ratio for comparison to the Level A, B, C, D, and pure membrane stress tabular allowable flaw sizes, given in Tables C-53 10-1 through C-5310-5, is conservatively calculated without reducing the expansion stresses by the structural factor and conservatively using the Level C stresses for Level A and Level B. Also note that the Level C stress is approximately I ksi lower than the Level b stress, based upon calculations using the forces and moments in Reference [5], and the structural factor for Level C is larger than for Level D; therefore, the conservative Level C stress calculated for this evaluation is considered bounding for the Level D stresses:

SR = 1326 (4364+10746) =0.365 29700.r+80000)

c. Table 6 summarizes the allowable flaw sizes for Level.A, Level B, Level C, Level D, and pure membrane stress, obtained from Tables.C-53 10-4 through C"53 10-5, considering the conservative method and assumptions used for this flaw evaluation. Also shown in Table,6 is the end of evaluation periodflaw-depth to wall thickness ratio, 2a/t. It can be seen that the end of evaluation period flaw size meets the Appendix CUallowable flaw size requirements~for all conditions.

Table 6: Allowable Flaw Sizes for RV Circumferential Indication.

End of Allowable Evaluation Condition l/htd SR Table 2/t Prd 2a/t Period 2a/t Level A 0.089 0.365 C-5310-1 0.75 0.565 Level B 0.089 0.365 C-5310-2 0.75 0.565 Level C 0.089 0.365 C-5310-3 0.75 0.565 Level D 0.089 0.365 C-5310-4 0.75 0.565 Puire Membrane 0.089 0.365 C-5310-5 0.75 0.565 Stress File No.: 0801014.372 Revision: 0 Page 1.8of21 F0306-01

Structural Integrity Associates, Inc.

VII.

Proximity Check Per IWA-3300 [1b] a subsurface flaw must be characterized as a surface flaw if S < 0.4a, where S is the minimum distance between the flaw and the free surface and a is the half depthof the flaw. The minimum distance between the reported indications, shown in Table 1, and the free surface, excluding the butter, prior to crack growth, is equal to 1.09-0.357-0.423-0.3,i0 inches. The distance between the free surface andecrack tip is reduced to 0.31-0.096 = 0.214 inches at the end of the evaluation period. This gives" S

_0.214 Sa/2).

0.21/

= 069 > 0.4 ; thus,:the flaw remains a subsurface flaw.

VIII.

Primary Stress Check Per IWB-3610(d)(2), the primary stress limits of NB-3000 must be maintained considering the local reduction in the pressure retaining area of the pipe caused by presence of the crack.

a. Treating the end of evaluation interval crack as a rectangular flaw of dimensions 1.692" x 0.615" gives a reduction in tensile area of the pipe wall of 1.04 in2; therefore, the pipe area available to support a membrane load becomes 17.07-1.04.= 16.03 in2. The allowable stress limit for primary membrane stress is Sm=2 3.3 ksi. It is recognized that presence of a circumferential indication does notreduce the load carrying capacity of the pipe in the hoop direction; rather, it affects the load carrying capacity of the pipe in the axial direction. Further, for cylindrical components, the axial, hoop, and. radial component stresses are the principal stresses; therefore, it can be shown that the primary membrane stress intensity limits remain satisfied at the flawed location by demonstrating that the pseudo-stress intensity calculated using the difference between the radial and axial stresses remains less than the primary membrane stress-intensity limit, By conservatively using the bounding Level A, B, C, and D loads fromReference [5], it can.be shown that the pseudo-stress intensity calculatedconsidering the affect of the flawed cross-seciion..is, always less than:

=(2z.gd2 +

(P

.*3.875' 4534 3125 1P' =+P 4I

-I_=I 125..

+,

1+

=4144 psi

~4.A A

t2)K 4-16;03 16.03) 2 Therefore, Pm< S,, in the cracked condition.

b. Conservatively assuming the flaw occurs at the azimuth of the largest bending stress, the effect of the flaw on the available section modulus can be determined by subtracting the contribution of the crack area to the uncracked pipe section modulus as shown below:

d).c1 L64d,=1 6.063 File No.: 0801014.372 Page 19 of 21 Revision: 0 F0306-01

Structural Integrity Associates, Inc.

Z = 16.11 in' Where:

1 is the area moment of inertia, in.4 d,

is the pipe outer diameter, in di is the. pipe inner diameter, in 2a is the crack depth, in I

is the crack length c

is the distance between the centroid of the crack and the centroid of the uncracked pipe, in. Since the crack eccentricity is very close to-zero this distance is assumed as the pipe mid-wall location.

The allowable stress limit for primary membrane plus bending stress is, I,5Sm1=34.95 ksi. By conservatively using.the bounding Level A, B., C, and D loads from Reference [5], it is shown that Prn+Pb is always less than:

P. +I~(~

P=(64923).

+4144 =8174 psi

" )

\\16.11J Therefore, Pm + Pb < 1.5Sm in the cracked condition.

6.0 RESULTS OF ANALYSIS The results of the allowable flaw size calculation considering FCG for 30 years are presented in '[able 6.

The end of evaluation interval flaw size satisfies the Section X1, lWB-3600 allowable flaw size requirements.

A proximity check was performed to confirm that the flaw will remain a subsurface flaw at the end of the evaluation period.

A primary stress check was performed to confirm that the NB-3200 primary stress limits remain satisfied.

7.0 CONCLUSION

S AND DISCUSSIONS A conservativ e flaw evaluation of the two circumferential indications identified in the CCNP Unit I Pressurizer RV Nozzle was performed using the methods of IWB-3600 [I b] considering an operating period to the end of current licensed life (30 years). Significant conservatisms were:applied throughout the flaw evaluation process to ensure. thata bounding evaluation was performed.. The indications satisfy all applicable Section Xl IWB-3600 acceptance criteria; therefore, the indications areaCcePtabl!e to be left in the' as-found condition for the remainder of the licensed plantlife (assumed to be: 30 years).

8.0 REFERENCES

I. American Society of Mechanical Engineers Boiler and Pressure Vessel Code:

a. Section 11, Part D, 2004 Edition without Addenda.

b. Section Xl, 2004 Edition without Addenda.

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Structural Integrity Associates, Inc.

2. Phased Array Ultrasonic Examination Record, Data Sheet No. CC10-IU-067, "

SUMMARY

OF WELD OVERLAY ULTRASONIC EXAMINATIONS - Calvert Cliffs Nuclear Power Plant Unit 1," SI Report No. 0801014.402.

3. CCNPP Design Spec. No. 8067-31-4, Rev. 12, "Project Specification for a Pressurizer Assembly for Calvert Cliffs I and 2," Dominion Engineering, February 14, 2006, SI File No. 0801014.212.

4. CCNPP Calc. No. CA01457, Rev. 1, "CCNPP Unit I Pressurizer Relief Valve I-RV-201 Piping Analysis," SI File No. 0801014.231.

5. Padmala, A.," Full Structural Weld Overlay Sizing for the Four-inch Pressurizer Safety/Relief Valve Nozzles," Rev. 0, SI File No. 0801014.370, CCNPP Calc. No. CA07281, Rev. 0.

6. NUREG/CR-672 1, "Effects of Alloy Chemistry, Cold Work, and Water Chemistry on Corrosion Fatigue and Stress Corrosion Cracking of Nickel Alloys and Welds," U.S. Nuclear Regulatory Commission (Argonne National Laboratory), March 2001.

7. Combustion Engineering Inc., "Instruction Manual Pressurizer," Calvert Cliffs Station, C.E. Book No.

72367, S1 File No. 0801014.213.

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