ML20078M218

Nonproprietary Fm Assessment of CR-3 Prz WP-15 Weld Flaw Until End-of-Life

ML20078M218
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Site:	Crystal River
Issue date:	01/25/1995
From:	Leslie Hill, Killian D BABCOCK & WILCOX CO.
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ML20078M208	List: 3F0295-03, Forwards Nonproprietary Version of B&W Nuclear Technologies Document 32-1236235-00, Fm Assessmnt of CR-3 Prz WP-15 Weld Flaw Until End-of-Life ML20078M218
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32-1236235, 32-1236235-00, NUDOCS 9502130276
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SUMMARY

SHEET (CSS) i DOCUMENT IDEFTIFIER 32-1236235-00 TITLE FM Assessment of CR-3 PRZ WP-15 Weld Flaw until End-of-Life PREPARED BY REVIEWED BY

,w L.T. Hill name D.E. Killian /24[.

' # $1GNATURE StoNATURE TITLE Engine II oxyg [ h [ TITLE Principal Engineer 1./[25/99 ,

cosi CENTER 4 20 REF. PAGE(S) 31 TM STATEMENT: REVIEWER !NDEPENDENCE ,

s bRPOSE AND

SUMMARY

or RtSULTS:

Purpose During 9R Section XI examinations of the CR-3 pressurizer, a flaw indication was detected in the WP 15 weld connecting the Surge nozzle to the lower head. The flaw was determined to exceed the flaw acceptance standards of ASME Section XI, IWB-3500.

The purpose of this acument is to determine the acceptability of the reported unacceptable indication per the IWB-3000 acceptance standards of the ASME Boiler and Pressure Vessel Code,Section XI. Through the understanding of stresses, material properties, and geometry acting on this body, a linear clastic fracture mech.mics analysis is performed in accordance with the ASME Boiler and Pressure Vessel Code,Section XI to ensure that the flaw indicatic, will not reach a critical length in the CR-3 design life.

Summary of Results Table 4-3 shows the results of this analysis. A small amount of fatigue crack growth is expected to occur over the design life of this component (-0.0026 in.) Furthermore, fractme toughness margins do not exceedV10 for Normal / Upset conditions and V2 for the Emergency / Faulted conditions. Also, the limit load margin was found to be acceptable for all ex;,ected loading conditions. Hence, the CR 3 WP-15 surge nozzle-to-lower head weld flaw indication has been found to be acceptable by IWB-3612 criteria for continued safe operation until end-of-life.

Note: This document is classified as BWNT Non-Proprietary.

THE 70LLOWING COMPUTER CODES MAVE BEEN USED IN THIS DOCUMENit CODE / VER$10N / REV c0DE / VER$10N / REV Tli!S DOCUMENT CONTAINS ASSUMPil0NS THAT Musi BE VERIFIED PRIOR TO UsE ON SAFETY RELATED WORK YES ( ) NO ( X)

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- B&W NUCLEAR TECHNOLOGIES 32-1236235-00 TABLE OF CONTENTS Section P.ags  !

1.0 Introduction 5 i i

2.0 Overview Description of Geometry and Flaw 6 2.1 Geometry and Material Involved 6 2.2 Flaw Geometry 6  ;

3.0 Resolution of Actual Flaw into a Simple Flaw to be Analyzed 9 3.1 Characterize Flaw in Accordance with IWB-3610' 9 ,

1 3.2 Flaw Shape and Location 10 3.3 Flaw Orientation 10 .

4.0 Fracture Mechanics Assessment 11 4.1 IWB 3612 Flaw Accepatance Criteria Based on Evaluation 11 4.2 Stress Intensity Factor 12 4.3 Stresses for Evaluation 13  ;

4.4 Limit Load Solution 14 4.5 Fatigue Crack Growth 15 t 4.6 Spectrum Analysis Required for Fatigue Crack Growth 16 4.7 Material Properties 25 4.7.1 Fracture Toughness 25 4.7.2 Material Strength 25 5.0 Procedure Used in Analysis 26 i v

6.0 Results and Conclusions 29 l

7.0 References 31  :

Appendix A - SORT.F - FORTRAN Program Listing used to Create Fatigue Spectrum History 32 -

Appendix B - BURIED.F - FORTRAN Program Listing Used to Evaluate Fatigue Crack Growth of the WP-15 Weld Flaw and to Evaluate the Required Fracture Toughness and Limit Load Margins 33  !

Appendix C - FORTRAN Programs Verification 34 Appendix D - Microfiche 37 )

Attachment 1 - Letter from R.B. Reynolds to E.E. Organ Showing Fracture Toughness I Data of WP-15 Weld -38' Attachment 2 - Internal BWNT Memo from K.B. Stuckey to L.T. Hill, dated 11/10/1994 39 Prepared by: L.T. Hill Date: 12/15/94 Reviewed by: D.E. Killian Date: 1/25/95 Page 3

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32-1236235-00 B&W NUCLEAR TECHNOLOGIES i

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LIST OF FIGURES  !

Eigilts East ,

2 7 2-1 Geometry of the Pressurizer Surge Nozzle Location and Size of Flaw Indication Detected by Ultrasonic Testing I

2-2 in CR-3's Pressurizer Surge Nozzle 8  !

3-1 Characterization and Proximity Rules for Analytical Evaluation of Components' 9 3-2 Simplified Flaw Geometr*/ Used for Analysis 10 4-1 Approximating Stress Distribution as Linear 13 ,

4-2 Plastic Collapse Solution Utilized in this Analysis' 14 4-3 Procedure Used in BURIED.F in Calculating Fatigue Crack Growth  :

of N/U Transients 27 l

4-4 Procedure Used in BURIED.F for Determination of Fracture Toughness i

and Limit Load Margins for .9/U and E/F Conditions (Based on flaw dimensions afte.e Fatigue Crack Growth) 28 t

LIST OF TABLES r-Tahlt fatt 4-1 Number of Normall Upset Fatigue Cycles that the Pressurizer Surge Nozzle Undergoes 17 [

4-2 Spectrum History For WP-15 Surge Nozzle-to-Lower Head of- l Pressurizer Weld 18 j 4-3 Results of Fracture Mechanics Assessment of the WP-15 Weld l Flaw Evaluation 29 Prepared by: L.T. Hill Date: 12/15/94 Reviewed by: D.E. Killian Date: 1/25/95 Page 4 )

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. B&W NUCLEAR TECHNOLOGIES 32-1236235-00 1.0 Introduction During 9R Section XI examinations of the CR-3 pressurizer, a flaw indication was detected in the WP-15 weld connecting the surge nozzle to the lower head. The flaw was determined to exceed the flaw acceptance standards of ASME Section XI, IWB-3500.

- A fatigue crack growth analysis, Ref. 8, was prepared to justify continued operation of the plant.

- Due to time constraints and lack of detailed stresses at the location of the flaw, the analysis was .

performed assuming the maximum stress ranges from any transient were applicable for all transient cycles. This very conservative approach resulted in a flaw acceptability of only one fuel cycle.

The purpose of this document is to determine the acceptability of the reported unacceptable indication per the IWB-3000 acceptance standards of the ASME Boiler and Pressure Vessel Code,Section XI'. '

2 Through the understanding of stresses', material properties , and geometry (Section 2.0) acting on this body, a linear elastic fracture wh=aics analysis will be performed in accordance with the ASME Boiler and Pressure Vessel Code,Section XI' to ensure that the flaw indication will not reach a critical length in the CR-3 design life.

The following is a summary of the analytical procedure undertaken:

(a) Characterize the flaw in accordance with IWB-3610'.

(b) Using A-2000', resolve the actual flaw into a simple flaw to be analyzed.  ;

(c) Determine stresses at the location of the observed flaw for normal & upset, emergency, and faulted conditions.

(d) Using A-4000', determine the necessary material properties.

(e) Calculate stress intensity factors for each condition in accordance with A-3000'. .i (f) Evaluate the flaw evaluation criteria of IWB-3600' to determine whether the observed flaw is acceptable for continued operation.

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Prepared by: L.T. Hill Date: 12/15/94  ;

Reviewed by: D.E. Killian Date: 1/25/95 Page5 ,

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B&W NUCLEAR TECHNOLOGIES 32-1236235-00 i

i 2.0 Overview Description of Geometry and Flaw ,

1 2.1 Geometry and MaterialInvolved ,

• Ihe geometry of the pressurizer surge nozzle is given in Reference 2 and is summarized in Figure 2-

1. The component consists of a A-508 Class 1 carbon steel (modified by ASME B&PV Code Case i 1332-4*), nozzle welded to the SA-516 Grade 70 carbon steel lower head. Both the nozzle and l pressurizer head are clad with stainless steel to prevent reactor coolant fluid from contacting the carbon steel base metal.

The initial weld was performed using the Semi-Automatic Gas Metal Arc welding process with fabricated McKay E70A1 flux cored weld wire (WP-15 Rev 4). Weld repairs were performed at various times using the Shielded Metal Arc welding process using E7015-Al electrodes (B&W manufactured) or E7018-Al electrodes (WP-15 Alt 1 Rev 3).5 Furthermore, post-weld heat treatment (PWHT) was performed at 1100*F to 1150"F for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> total.8 Hence, with this amount of stress relief, the effect of weld residual stresses was deemed to be negligible.

2.2 Flaw Geometry  !

Figure 2-2 shows the location and size of the indication found by ultrasonic testing.

The following section will demonstrate how this flaw is to be evaluated. ,

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Prepared by: L.T. Hill Date: 12/15/94 Reviewed by: D.E. Killian Date: 1/25/95 Page 6

Bd.W NUCLEAR TECHNOLOGIES 32-1236235-00

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Prepared by: L.T. Hill Date: 12/15/94 Reviewed by: D.E. Killian Date: 1/25/95 Page 7 l

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1. aauerue.s u.s.s Figure 2-2 Location and Site of Flaw Indication Detected by Ultrasonic Testing in CR-3's Pressurizer Surge Nozzle Prepared by: L.T. Hill Date: 12/15/94 Reviewed by: D E. Killian Date: 1/25/95 Page 8

l B&W NUCLEAR TECHNOLOGIES 32-1236235-00 1 l

3.0 Resolution of Actual Flaw into a Simple Flaw to be Analyzed 3.1 Characterize Flaw in Accordance with IWB-3610' Using Figure 3-1, it is possible to characterize the subsurface flaw contained entirely in the ferritic steel as either a surface or a subsurface flaw depending on tl:.e relationship between S and d. As seen, the flaw can be characterized as subsurface.

Flaw Ferritic Steel 4

2d A

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Nearest Surface If S 2: 0.4d, then a subsurface flaw is assumed.

If S < 0.4d, then a surface flaw is assumed.

S = 1.21 in. 2d = 0.62 in.

1.21 > 0.4(.31) 1.21 > 0.12 Therefore, a subsurface flaw will be used.

Figure 3-1 Characterization and Proximity Rules for Analytical Evaluation of Components' 1

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Prepared by: L.T. Hill Date: 12/15/94 j Reviewed by: D.E. Killian Date: 1/25/95 Page 9

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B&W NUCLEAR TECHNOLOGIES 32-1236235-00.

I 3.2 Flaw Shape and Location ne flaw indication is completely' circumscribed by an elliptical planar area as outlined in IWA-3300'. .

For purposes of this analysis, the flaw is considered in its actual location per IWA-3300'.

3.3 Flaw Orientation  ;

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As shown in Figure 2-2, the indication is a non-planar elliptical subsurface flaw and should be i projected in planes normal to the stresses that are related to the longitudinal and radial directions per

- IWA-3340'. However, the radial stress, as shown in Ref. 6, is found to be insignificant. In order to ,

perform a conservative analysis, the flaw will be assumed to be normal to the longitudinal stress. l 1

The flaw model used for this analysis is depicted below in Figure 3-2.  ;

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2a = 0.62 in, e= 0.855 in.

1 = 1.70 in, t = 4.75 in.

4 Figure 3-2 Simplified Flaw Geometry Used for Analysis Prepared by: L.T. Hill Date: 12/15/94

' Reviewed by: D.E. Killian Date: 1/25/95 Page 10 ,

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B&W NUCLEAR TECHNOLOGIES 32-1236235-00 4.0 Fracture Mechanics Assessment 4.1 IWB-3612 Flaw Acceptance Criteria Based on Evaluation A flaw is acceptable if the applied stress intensity factor satisfies the following IWB-3612' criteria.

. (a) For normal and upset condition:

i Ku K,<

06 where K, - the maximum applied stress intensity factor for normal and upset conditions ,

based on final crack length. '

Ku = crack arrest fracture toughness for the corresponding crack tip temperature.

-(b) For emergency and faulted condition:

K K,<l A

I K, = the maximum applied stress intensity factor for emergency and faulted .

conditions based on final crack length.

Ku = fracture initiation fracture toughness for the corresponding crack tip temperature.

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9 Prepared by: L.T. Hill Date: 12/15/94 Reviewed by: D.E. Killian Date: 1/25/95 Page 11 l l

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i B&W NUCLEAR TECHNOLOGIES 32-1236235-00 4.2 Stress Intensity Factor The stress intensity factor for the flaw model is calculated using the following equation:

K, = (a,,M,,+ a,M,) ".

, O where a., c. = membrane and bending stresses, ksi, a = minor half-diameter, in.

Q = flaw shape parameter M. = correction factor for membrane stress M. = correction factor for bending stress In Section III- Appendix G and Section XI- Apeendix A of the ASME Code, the small-scale ,

plasticity adjustment is hidden in the flaw shape parameter, Q. The flaw shape parameter is assumed to follow Figure A-3300-l' as follows:

G = 1 + 4.593 (a/I)us - 0.212 (de)2 where a is conservatively assumed to be the sum of the absolute value of the membrane stress and the absolute value of the bending stress.

M., M. are the loading type correction factors from the ASME Section XI, Appendix A procedure given graphically in Figures A-3300-2 and A-3300-4 for subsurface flaws in Appendix A. Note that no corrections for aspect ratio, all, are provided for the buried flaws. The "Ptl" designation in Figures A-3300-2 and A-33004 refers to the point on the crack front nearest the closest surface and "Pt2" is the other point on the minor axis. A polynomial form by Cipolla* (listed as part of the FORTRAN program listing in Appendix B of this document) is available which corresponds to the ASME Section XI, Appendix A curves. It is this form that will be utilized to determine the loading type coefficients when the stress intensity factor calculations are made.

i Prepared by: L.T. Hill Date: 12/15/94 Reviewed by: D.E. Killian Date: 1/25/95 Page 12

B&W NUCLEAR TECHNOLOGIES 32-1236235-00 4.3 Stresses for Evaluation Stresses in the surge nonle-to-pressurizer weld (the flaw location) are caused by the pressure, '

thermal, and external loadings associated with thermal stratification of the surge line.

1 As given by Ref. 6, the opening mode longitudinal stresses are given at the inside and outside surface l of the pressurizer at the location of the surge nozzle-to-lower head weld. However, as defined by the l stress intensity factor solution being used, the stresses should be broken into membrane and bending components. Fig. 4-1 demonstrates how the membrane, a., and bendmg, a., stress components are defined. Note that the bending stress is defined as being positive on the outboard side of the neutral axis. Section 4.6 gives magnitudes of the membrane and bending stresses utilized for analysis of the .;

Normal / Upset and Emergency / Faulted conditions.  :

Figure 4-1 Definition of Membrane and Bending Stress Components I

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Prepared by: L T. Hill Date: 12/15/94 Reviewed by: . D.E. Killian Date: 1/25/95 Page 13  ;

32-123f,235-00 B&W NUCLEAR TECHNOLOGIES 4.4 Limit Load Solution In addition to failure by brittle fracture, a net section collapse failure needs to be assessed.

The safest approach in a fracture mechanics assessment is to adopt an analysis that spans the entire range from linear clastic to fully plastic behavior. Such an analysis accounts for the two extremes of brittle fracture and plastic collapse.

The limit load solution for a semi-elliptical surface crack in a flat plate subject to combined tension and bending is used in this analysis, see Figure 4-2. This is a conservative as-sumption for the embedded flaw being analyzed, since the bending stress will have a great-er effect on a surface flaw than on an embedded flaw. The crack depth, a, is postulated to be the 0.62 in. for the initial crack depth while the initial crack length, 2c, is modeled as 1.70 in. In addition, the width of the plate is assumed to be infinite.

Since no margin of safety is applied to the limit load condition, the component will not fail by plastic collapse if o,, / am < 1 Figure 4-2 Plastic Collapse Solution Utilized in this Analysis' og -3ay(1-a)2 /(A+(A2+9(1 a)2)3)

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Prepared by: LT. Ilill Date: 12/15/94 Reviewed by: D.E. Killian Date: 1/25/95 Page 14

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B&W NUCLEAR TECHNOLOGIES 32-1236235-00 4.5 Fatigue Crack Growth In order to determine the final crack length, a r, as required by IWB-36128, a fatigue crack growth will be required to be performed. The fatigue crack growth rate da/dN of the material i

'is characterized in terms of the range of applied stress intensity factor AKi . The Paris Law  ;

characterization is of the form:  !

f da 1

=C,(AK,)" i dN where

• AK,= - (Aa.M. + AaN v'(ra/Q)

C, = scaling constant i n= the slope of the log (da/dN) versus log ( AK) i Using Figure A-4300-1,8 the coefficients C, and n are 2.67E-11 and 3.726. Note that the scaling constant C, produces fatigue crack growth rates in units of in./ cycle where AK, is in units of ksiv'in. Also if the R ratio (K. /KJ < 0 (i.e. K. < 0) then K. is set equal to

0. This is to prevent the non-conservative effect of crack closure (i.e. reverse yielding at the crack tip which retards fatigue crack growth).

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Prepared by: L.T. Hill Date: 12/15/94 Reviewed by: D.E. Killian Date: 1/25/95 Page 15

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'32-1236235-00' l B&W NUCLEAR TECHNOLOGIES a 4.6 Spectrum Analysis Required for Fatigue Crack Growth Reference 2 states the number of cyclic stresses that the current flaw indication could undergo in the life of this component. Table 4-1 demonstrates the number of transients per design life  ;

and the associated number of stress cycles per transient.

To predict the life of a component subjected to a variable load history (such as for the surge nozzle-to-lower head weld geometry), it is necessary to reduce the complex history into a -

• number of events which can be compared to the available constant amplitude test data.

First, it is necessary to determine the severity of each individual Peak-Valley, PV, within a given transient. This is done in this analysis by computing the stress intensity factor at point I for flaw dimensions corresponding to the initial flaw size. From this, the PVs are sorted '

based on this stress intensity factor.

Once the PVs have been categorized, they must be combined to form completed fati:ue cy-cles. The following methodology is the most conservative method for matching PVs. The most damaging combination of PVs (matching of the PV with the highest stress intensity  ;

factor with the PV having the lowest stress intensity factor) is obtained by first forming the largest possible cycle. The next largest cycle possible is then formed by using the remaining PVs available, and so on, until all PVs have been used. The procedure is repeated for each of the Normal / Upset transients studied. Section 5.0 gives further details on the application of

- this technique through the development of a FORTRAN computer program. The results of this analysis are shown in Table 4-2.

L.T. Hill Date: 12/15/94 Prepared by: f Date: 1/25/95 Page 16 Reviewed by: D.E. Killian  ;

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B&W NUCLEAR TECHNOLOGIES 32-1236235-00 I Table 4-1 Number of Normal / Upset Fatigue Cycles that the Pressurizer Surge Nozzle Un-dergoes Number of Fatigue Total Transient Transients Cycles Fatigue

/ Transient Cycles hulal 10 35 350 ,

hula 2 8 32 256 hula 3 45 33 1485 hula 4 29 34 986 hula 5 148 33 4884 cdibt 40 38 1520 cdib2 200 34 6800 2a 1440 1 1440 2b 1440 2 2880 3 48000 1 48000 4 48000 1 48000 7 310 2 620 8a 80 1 80 8b 162 1 162 8c 88 1 88 i

8d 70 1 70 14 40 1 40 20b 20000 1 20000 20d2 34000 1 34000 22a1 5 1 5 15 1

22b1 15 1 22c1 10 1 10 ,

22dl 10 1 10 22a2 7 5 35 22b2 42 5 210 22c2 7 5 35 22d2 100 5 500 172481 Fatigue Cycles in Design Life i

Prepared by: L.T. Hill Date: 12/15/94 Reviewed by: D.E. Killian Date: 1/25/95 Page 17 F

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B&W NUCLEAR TECHNOLOGIES 32-1236235-00 Table 4-2 Spectrum History For WP-15 Surge Nozzle-to-Lower Head of Pressurizer Weld Cycles in Max Min e max S1F Opoint 1 Design Stress State Stress State Crack Tip Max Min Life Mem Bend- Mem Bend - Temp SIF S1F (ksi) (ksi) (ksi) (ksi) (degF) (ksVin) (ksVin)

Transient: hulal 10 13.95 21.75 .95 -29.55 521.00 24.69 -11.79 10 11.10 21.20 1.75 -24.25 342.00 20.64 -8.39 10 12.55 15.35' 1.80 -14.50 513.00 18.79 -4.06 10 8.70 20.40 2.35 -12.15 249.00 17.31 -2.61 10 11.15 13.55 2.35 -11.65 521.00 16.27 -2.41 '

10 11.45 12.65 4.75 -14.55 414.00 16.12 -1.39 10 11.95 -10.45 2.70 -9.10 508.00 15.53 -1.09 10 10.65 12.25 2.25 -7.15 521.00 15.08 .74 10 8.45 16.55 2.40 -6.30 223.00 14.95 .27 10 11.40 10.40 2.25 -5.95 407.00 14.94 .27 10 11.30 9.70 2.55 -6.35 508.00 14.50 .16 10 8.30 15.40 3.30 -8.00 216.00 14.21 .12 10 9.95 10.75 2.45 -5.85 510.00 13.63 .06 10 9.30 12.10 .05 .05 513.00 13.62 .06 10 10.30 7.50 .4.70 -10.50 504.00 12.47 .17 10 10.25 7.55 2.45 -2.65 503.00 12.45 1.17 10 10.35 6.35 2.45 -2.65 502.00 12.01 1.17 -

10 4.05 19.85 4.90 -7.60 241.00 12.00 1.48 10 - 9.70 6.80 3.85 -2.65 504.00 11.58 2.42 10 6.40 11.90 6.60 '8.40 189.00 10,64 2.73 10 5.95 12.85 8.10 -10.10 276.'JO 10.63 3.48 10 9.00 5.60 3.10 2.50 5G400 10.39 3.72 10 7.55 8.55 6.85 -5.95 114.00 10.28 3.89 10 6.65 10.25 6.60 -5.30 171.00 10.16 3.90 10 7.20 9.00 8.20 -7.20 171.00 10.14 4.66 10 5.70 11.40 8.00 -5.70 189.00 9.75 5.04 10 6.55 8.15 7.35 -3.45 257.00 9.16 5.28 10 -5.80 9.50 8.35 -5.05 264.00 9.02 5.61 10 5.80 7.90 4.30 4.70 114.00. 8.35 5.66 10- 3.80 12.10 5.30 -3.20 177.00 8.23 5.97 10 8.80 .40 8.55 4.55 446.00 7.74 5.98 10 8.45 .35 8.60 -4.00 547.00 7.70 6.23 10 7.95 .85 8.30 -2.40 449.00 7.45 6.54 10 8.15 .35 8.55 -1.25 547.00 7.43 7.19 10 5.65 5.95 5.90 5.20 503.00 _

7.40 - 7.33 Transient: hula 2 8 12.70 18.50 . 20 -22.10 427.00 20.86 -9.15 8 10.90 19.90 .55 -17.75 521.00 19.58 -6.54 8 10.40 15.60 1.05 -12.55 521.00 16.55 -3.94 8 11.90- 10.70 1.55 -12.65 407.00 15.61 -3.53 8 9.25 15.95 1.35 -9.65 521.00 15.49 -2.52 8 10.75 11.25 1.50 -9.50 508.00 14.69 -2.32 8 9.35 12.15 3.90 -14.60 513.00 13.69 -2.19 Prepared by: L.T. Hill Date: 12/15/94 Reviewed by: D.E. Killian Date: 1/25/95 Page 18

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-* . 1

' B&W NUCLEAR TECHNOLOGIES 32-1236235-00 l 1 Table 4 2 (continued)

Cycles in Max Min @ max SIF Opoint 1 Design Stress State . Stress State Crack Tip - Max - Min Life Mem - Bend Mem Bend Temp SIP SIF l (ksi) (ksi) (ksi) ' (ksi) (degF) (ksVin)(ksVin) '!

l 8 9.85 9.05 1.20 -8.40 407.00 12.74 -2.16 l 8 6.40 16.20 1.40 -7.80 223.00 12.64 -1.75 8 10.30 7.50 1.55 -7.85 504.00 12.47 -1.63 8 6.90 12.30 1.70 -6.50 189.00 11.31 .98 8 6.25 13.55 1.30 -4.90 177.00 11.24 . .72 8 6.10 .13.80 3.90 -10.80 177.00 11.21 .67 .

8 9.65 5.75 3.90 -10.00 M2.00 - 11.07 .36 8 6.30 12.90 3.20 ' 7.60 -

307.00 11.00 .06 i

8 7.05 9.35 .00 .00 508.00 10.15 .00 8 5.25 12.65 1.75 -3.35 189.00 9.86 .28 .;

8 5.70 11.50 4.05 -7.55 189.00 9.79 .73 8 5.10 12.60 1.55 -1.65 276.00 9.69 .75 8 5.85 10.25 -4.10 -7.60 283.00 9.39 .75 8 5.10 9.40 3.95 -5.85 264.00 8.32 1.29  :

8 5.05 7.85 3.30 -2.70 114.00 7.63 1.91

8 8.55 -1,25 4.90 -4.20 542.00 7.19 2.77 8 5.05 6.75 ' 4.05 .75 511.00 7.17 3.32 I

8 8.30 .80 - 1.75- 4.85 542.00 7.13 3.42 8 5.75 4.85 7.65 8.85 503.00 7.05 3.53 8 2.55 11.85 8.00 -8.60 177.00 6.95 3.95 8 2.50 11.80 3.45 4.45 177.00 6.88 -4.79 8 5.70 3.90 7.55 -5.15 502.00 6.62 4.83 8 -7.95 -1.45 4.50 2.90 - 542.00 6.57 5.13 8 9.45 -5.45 4.80 4.10 513.00 6.48 5.88  ;

8 9.20 -5.50 3.70 6.90 513.00 6.23 5.99 Transient: hula 3 45 11.25 13.85 .10 -20.40 521.00 16.54 8.29 45 11.55 12.35 .50 -16.80 513.00 16.08 -6.18 r 45 10.95 12.55 1.80 -11.20 513.00 15.55 -2.73 45 10.75 11.25 1.80 -10.60 508.00 14.69 -2.49 45 11.40 8.40 1.35 -8.65 503.00 13.98 -2.12 45 8.50 13.80 1.95 -9.75 521.00 13.62 -2.02  ;

45 7.25 15.75 1.50 -8,50 - $21.00 13.29 -1.93 45 6.20 16.20 3.75 -12.75 223.00 12.43 -1.58 45 '10.25 7.35 1.35 -6.55 503.00 12.36 -1.31 ,

45 9.75 7.75 1.55 -6.75 503.00 12.05 1.21 ,

45 10.10 6.90 1.80 -5.80 503.00 12.01 .62 45 9.65 5.75 3.70 -9.20 502.00 11.07 .23 45 6.65 11.75 . 1.80 -4.70 283.00 10.82 .20 45 6.00 12.90 .05 .05 ' 177.00 10.70 .06 45 5.95 11.95 4.10 -7.80 307.00 10.23 .68 ,

- 45 3.15 17.15 3.40 -5.70 241.00 9.82 .85 45 5.45 11.15 1.90 -1.90 283.00 9.40 .%  ;

45 5.30 10.90 3.95 -6.55 141.00 9.15 1.02 45- 6.20 7.80 2.15 -1.75 511.00 8.68 1.24  ;

Prepared by: L.T. Hill Date: 12/15/94 ,

Reviewed by: D.E. Killian Date: 1/25/95 Page 19

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B&W NUCLEAR TECHNOLOGIES 32-1236235-00 l

1 Table 4-2 (continued)

Cycles in Max Min @ max SIF @ point 1 Design Stress State Stress State Crack Tip Max Min Life Mem Bend Mem Bend Temp SIF SIF (ksi) (ksi) (ksi) (ksi) (degF) (ksVin) (ksVin) 45 5.45 9.25 4.25 -5.05 283.00 8.58 1.87 l 45 4.65 9.65 3.80 2.70 264.00 8.00 2.36 45 4.60 8.30 4.60 -3.70 171.00 7.39 2.69 45 8.55 -1.25 4.75 -3.55 542.00 7.19 2.89 45 8.25 .55 1.70 4.50 542.00 7.18 3.24 45 7.90 .40 4.00 .10 542.00 6.92 3.60 45 4.75 6.75 3.25 3.75 98.00 6.90 4.34 45 8.80 -2.70 3.65 3.25 532.00 6.89 4.50 i 45 8.30 -2.30 4.05 2.55 532.00 6.58 4.59 45 5.40 3.80 7.75 -5.85 502.00 6.30 4.75 45 4.75 5.05 4.20 3.20 504.00 6.21 4.98 45 8.55 -4.35 8.30 -6.40 532.00 6.05 5.06 45 2.75 8.95 4.45 3.25 141.00 5.95 5.23 45 2.35 9.65 7.60 -3.30 141.00 5.86 5.57 Transient: hula 4 29 11.35 14.05 .35 -28.15 513.00 16.75 -11.58 29 11.15 13.55 1.05 -22.75 521.00 16.27 -8.31 29 9.40 15.70 1.35 -10.65 521.00 15.52 -2.91 29 10.65 12.25 1.95 11.45 521.00 15.08 -2.69 29 7.40 18.00 2.45 -12.45 249.00 14.58 -2.64 29 11.25 9.75 1.50 -9.50 508.00 14.47 -2.32 29 9.30 12.10 1.35 -7.75 513.00 13.62 -1.77 29 7.20 15.10 1.50 -7.50 223.00 12.92 -1.54 29 10.30 7.50 4.85 -14.75 504.00 12.47 -1.38 29 7.15 13.35 2.50 -9.30 189.00 12.04 -1.35 29 10.35 6.35 2.20 -7.90 502.00 12.01 -1.07 I

29 8.90 8.80 1.45 -4.35 508.00 11.69 37 29 6.90 12.30 2.35 -6.05 189.00 11.31 .22 29 6.10 13.10 .00 .00 276.00 10.89 .00 29 9.00 5.60 3.00 -5.90 502.00 10.39 .4?

29 5.75 11.55 2.35 -2.45 189.00 9.86 1.15 29 5.85 10.15 2.60 -2.50 283.00 9.35 1.36 29 5.35 10.75 5.00 -7.60 141.00 9.13 1.57 29 6.40 7.60 2.40 -1.40 511.00 8.78 1.60 29 6.95 6.35 3.55 -3.15 504.00 8.77 1.96 29 4.90 10.10 3.95 2.85 171.00 8.43 2.43 29 5.95 7.55 6.15 7.55 257.00 8.34 2.64 29 5.10 9.40 4.40 2.20 264.00 8.32 3.08 29 3.35 12.25 4.50 -2.40 177.00 7.87 3.10 l 29 3.35 12.05 5.80 -4.60 177.00 7.79 3.44 l 29 8.45 .35 4.90 .90 547.00 7.70 4.02 1 29 5.80 6.10 2.25 5.55 503.00 7.60 4.14 l 29 8.15 .35 3.75 3.15 547.00 7.43 4.55 l 4.98 I 29 6.00 5.10 8.10 -6.10 403.00 7.38 29 8.50 -1.30 7.05 -3.45 542.00 7.13 5.01 Prepared by: L.T. Hill Date: 12/15/94 Reviewed by: D.E. Killian Date: 1/25/95 Page 20

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.- l B&W NUCLEAR TECHNOLOGIES 32-1236235-00 Table 4-2 (continued)

Cycles in Max Min @ max SIF Opoint 1  :

Design Stress State Stress State Crack Tip Max Min Life Mem Bend Mem Bend Temp SIF SIF i (ksi) (ksi) (ksi) (ksi) (degF) (ksVin) (ksVin) 29 8.30 -2.40 8.00 -5.70 532.00 6.54 5.04 29 4.40 6.60 4.10 4.20 161.00 6.51 5.28 29 8.60 -4.00 4.70 4.10 532.00 6.23 5.79 29 5.45 3.35 4.40 4.80 152.00 6.17 5.79 Transient: hula 5 148 11.35 14.05 .10 -18.10 513.00 16.75 . -7.28 148 11.15 13.55 .60 14.50 521.00 16.27 5.13 148 9.40 15.70 1.80 -11.20 521.00 15.52 -2.73

- 148 10.65 12.25 2.40 -12.40 521.00 15.08 -2.67 148 11.25 9.75 .80 -7.90 508.00 14.47 -2.32 148 9.30 12.10 1.55 -8.35 513.00 13.62 -1.83 148 10.30 7.50 4.85 -14.85 504.00 12.47 -1.42 '

148 4.15 20.25 2.55 -9.35 241.00 -12.29 -1.32 .

148 10.35 6.35 1.60 -6.70 502.00 12.01 -1.14 148 8.90 8.80 2.20 -8.00 508.00 11.69 -1.11 .

148 5.60 15.10 1.45 -5.85 223.00 11.31 .95 148 6.10 13.10 1.85 -5.85 276.00 10.89 .59 148 9.00 5.60 2.35 -6.05 502.00 10.39 .22 148 5.85 11.65 .10 .00 189.00 10.00 .09 148 5.25 11.55 4.85 -10.65 141.00 9.38 .25 148 5.45 11.05 3.00 -5.90 283.00 9.35 .42 l 148 5.35 10.75 2.85 -3.65 141.00 9.13 1.14 148 6.00 9.20 5.00 -7.80 114.00 9.08 1.49 148 6.40 7.60 3.55 -3.15 511.00 8.78 1.96 .

148 6.90 6.40 4.05 -3.55 504.00 8.75 2.26 j 148 3.90 12.40 3.95 -2.65 177.00 8.46 2.51 .

148 5.95 7.55 6.15 -7.55 257.00 8.34 2.64 148 4.85 8.55 1.45 3.65 114.00 7.73 2.69 148 8.45 .35 - 4.40 -2.20 547.00 7.70 3.08 148 5.80 6.10 4.55 -2.35 503.00 7.60 3.16

• 148 6.05 5.35 4.90 . 90 403.00 7.53 4.02 148 8.15 .35 3.75 3.15 547.00 7.43 4.55 148 4.50 8.10 8.10 -6.10 171.00 7.22 4.98 148 8.55 -1.25 7.05 -3.45 542.00 7.19 5.01 148 4.70 6.60 8.00 -5.70 98.00 6.79 5.04 148 8.30 -2.40 4.70 4.10 532.00 6.54 5.79 148 4.30 6.40 4.40 4.80 98.00 6.34 5.79 ,

148 8.60 -4.00 5.45 3.25 532.00 6.23 6.13 ,

Transient: edib1 40 12.90 19.50 .60 -19.20 521.00 21.75 -7.12 40 11.60 7.80 .55 -18.85 504.00 13.90 -7.02 40 7.25 16.55 .95 -15.85 276.00 13.69 -5.38 40 7.00 16.50 .70 15.30 342.00 13.40 -5.38 40 6.00 17.40 2.75 -19.25 216.00 12.81 -5.20  !

40 5.70 17.40 .30 -13.90 216.00 12.50 -5.16

)

Prepared by: L.T. Hill Date: 12/15/94 Reviewed by: D.E. Killian Date: 1/25/95 _ Page 21

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., l B&W NUCLEAR TECHNOLOGIES 32-1236235-00 Table 4-2 (continued)

Cycles in Max Min @ max SIF @ point 1  ;

Design StressState Stress State Crack Tip Max Min Life Mem Bend Mem Bend Temp SIF SIF (ksi) (ksi) (ksi) (ksi) (degF) (ksVin) (ksVin) 40 10.85 6.15 .85 -14.25 502.00 12.40 -4.81 40 6.50 13.10 2.35 -16.25 177.00 11.28 -4.29 40 6.75 12.05 1.15 -13.05 189.00 11.05 -4.05 40 6.20 12.70 2.00 -14.30 189.00 10.81 -3.80 40 9.45 4.35 2.40 13.70 502.00 10.28 -3.19 40 5.55 12.55 1.75 -10.65 177.00 10.11 -2.55 40 5.70 12.20 2.55 -11.65 189.00 10.10 -2.23 40 6.75 9.75 2.30 -10.70 521.00 10.03 -2.08 40 6.50 10.20 2.05 -9.95 171.00 9.99 -2.01 40 6.45 10.05 2.45 -8.95 521.00 9.88 -1.26 40 6.30 9.90 .20 -3.20 $21.00 9.67 -1.04 40 9.75 1.95 2.75 -8.85 449.00 9.55 .95 40 6.15 9.95 3.45 -9.45 171.00 9.55 .55 1 40 5.95 10.05 2.60 -7.40 171.00 9.40 . 52 40 5.15 10.55 2.80 -7.80 217.00 8.85 .50 40 4.35 11.85 3.20 -8.00 521.00 8.65 .21 40 9.15 .95 3.20 -7.30 547.00 8.59 .06 40 4.75 9.15 3.10 -6.90 213.00 7.89 .12 40 4.40 9.90 3.85 -8.45 217.00 7.87 .20 40 4.80 9.00 2.10 -4.00 213.00 7.87 .34 40 5.35 7.65 2.95 -5.85 161.00 7.82 .39 40 5.65 5.45 .35 1.65 155.00 7.20 .94 40 9.50 -4.10 5.05 9.05 445.00 7.02 1.06 40 4.60 7.20 5.30 -8.80 161.00 6.94 1.38 40 5.30 5.40 4.20 -6.10 253.00 6.86 1.42 40 4.85 6.25 2.55 2.75 161.00 6.79 3.32 40 2.90 9.50 2.25 3.85 171.00 6.31 3.48 40 5.50 3.50 5.40 - ",.20 152.00 6.28 3.60 40 5.10 4.20 5.40 -3.10 502.00 6.19 3.64 40 3.55 7.15 5.25 -2.35 114.00 5.% 3.79 40 4.45 5.05 5.10 .60 504.00 5.94 4.31 40 2.65 8.05 7.30 -5.40 171.00 5.50 4.51 Transient: edib2 200 8.85 13.85 .20 -15.60 521.00 14.01 -5.95 200 10.85 6.15 3.20 20.50 502.00 12.40 -5.34 200 9.45 4.35 1.10 -14.50 502.00 10.28 -4.69 200 5.40 12.50 .95 -13.45 189.00 9.94 -4.39 200 5.10 12.00 1.60 ' -14.30 177.00 9.43 -4.16 f 200 9.15 .95 1.00 -12.00 547.00 3.59 -3.76

200 4.70 9.90 1.50 -13.10 171.00 8.15 -3.76 200 4.10 11.00 1.40 -12.10 141.00 8.05 -3.45 l 200 6.25 6.05 1.45 -11.05 504.00 8.00 -2.98 l 200 4.35 9.85 1.00 -9.80 141.00 7.80 -2.89 1 200 4.40 9.20 .10 -7.40 141.00 7.58 -2.75 200 4.60 8.70 1.50 10.10 114.00 7.56 -2.56 Prepared by

L.T. Hill Date: 12/15/94 Reviewed by: D.E. Killian Date: 1/25/95 Page 22

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e 32-1236235-00 B&W NUCLEAR TECHNOLOGIES Table 4-2 (continued)

Max Min @ max SIF Opoint 1 Cycles in Design Stress State Stress State Crack Tip Max Min Life - Mem Bend Mem Bend Temp SIF SIF (ksi) (ksi) (ksi) (ksi) (desF) (ksVin)(ksVin)

-5.85 171.00 7.36 -2.11 200 4.30 8.90 .15

-8.15 217.00 7.21 -1.75 200 4.00 9.20 1.55

-8.95 267.00 7.06 -1.53 200 4.35 8.05 2.15

-7.40 114.00 6.90 -1.24 .

200 4.20 8.00 1.80

-10.55 155.00 6.35 -1.07 200 4.95 4.95 3.35

-3.20 155.00 6.05 -1.04 200 4.90 4.30 .20

-5.80 98.00 5.98 .89 200 3.90 6.40 1.50 200 2.60 8.80 1.95 -6.75 171.00 5.75 .85

-7,30 152.00 5.52 .66 200 4.80 3.20 2.40

-5.85 257.00 5.52 .59 200 3.55 6.05 1.85

-2.15 161.00 5.34 .51 200 3.60 5.50 .35 200 3.35 4.45 1.95 -5.65 253.00 4.70 .43 200 3.30 4.50 1.95 -5.35 253.00 4.68 .31 200 3.50 4.00 2.00 -4.30 253.00 4.66 .14

-3.80 155.00 4.47 .24 200 3.35 3.85 1.90 200 7.25 -5.45 2.35 -4.45 513.00 4.44 .39 200 3.20 4.00 1.10 2.00 155.00 4.39 1.74 200 5.40 -1.70 1.20 2.50 446.00 4.17 2.02 200 2.85 4.05 2.00 2.20 88.00 4.10 2.62 200 2.45 4.25 3.80 -1.30 253.00 3.81 2.89 200 1.30 6.60 1.50 4.40 267.00 3.70 3.02

-3.15 542.00 3.08 3.04 200 4.75 -3.05 4.75 Transient: 2a 8.35 9.00 3.80 504.00 12.91 9.62 1440 10.35 Transient: 2b 1440 10.35 9.75 7.70 1.40 511.00 13.56 6.36 ,

9.25 8.55 3.65 511.00 13.23 9.14 1440 10.25 Tr=nate: 3 48000 10.50 10.80 9.10 3.80 511.00 14.21 9.72 Transient: 4 i

.75 502.00 11.37 7.01 48000 9.80 6.10 8.15 Transient: 7  ;

310 10.10 8.00 8.00 -1.10 504.00 12.50 6.*i5 310 9.35 6.85 8.55 3.65 504.00 11.26 9.14 Transient: 8a 80 9.60 7.60 9.10 -3.50 504.00 11.84 6.87 Transient: 8b 162 10.45 8.25 9.05 2.75 504.00 12.% 9.23 i Transient: Sc 88 10,85 5.85 9.40 1.00 502.00 12.27 8.84 Transient: 8d 70 10.45 5.15 9.75 1.35 502.00 11.57 9.30 Transient: 14 40 9.05 7.25 8.45 3.65 504.00 11.15 9.05 I

l i

12/15/94 i

Prepared by: L.T. Hill Date-. 1 Date: 1/25/95 Page 23 Reviewed by: D.E. Killian l 1

B&W NUCLEAR TECHNOLOGIES 32-1236235-00 Tabic 4-2 (continued)

Cycles in Max Min @ max SIF @ point 1 Design Stress State StressState Crack Tip Max Min Life Mem Bend Mem Bend Temp SIF SIF (ksi) (ksi) (ksi) (ksi) (degF) (ksVin) (ksVin)

Transient: 20b 20000 8.60 5.80 8.45 3.65 502.00 10.09 9.05 Transient: 20d2 34000 8.95 7.25 9.65 3.95 504.00 11.05 10.30 Transient: 22al 5 8.00 14.60 2.25 .95 177.00 13.50 2.36 Transient: 22bl 15 6.00 12.50 1.20 .50 177.00 10.52 1.26 Transient: 22cl 10 7.50 10.50 4.15 1.65 141.00 11.09 4.33 Transient: 22dl 10 6.35 9.25 3.45 1.35 114.00 9.44 3.59 Transient: 22a2 7 7.85 14.55 3.35 -14.65 177.00 13.32 -2.72 7 7.50 11.00 3.60 -12.70 141.00 11.31 -1.70 7 7.45 10.55 3.60 -12.10 141.00 11.06 -1.46 7 7.15 9.05 4.10 -5.60 114.00 10.11 1.52 7 2.20 1.00 2.20 1.00 70.00 2.34 2.34 Transient: 22b2 42 5.90 12,50 2.05 -12.65 177.00 10.42 -3.08 42 5.55 8.85 2.60 -7.40 114.00 8.51 .52 42 5.50 8.50 2.60 -7.10 114.00 8.32 .40 42 5.25 7.45 2.70 -5.40 114.00 7.65 .34 42 1.15 .45 1.15 .45 70.00 1.19 1.19 Transient: 22c2 7 8.00 11.10 4.30 -9.40 521.00 11.85 .24 7 7.85 10.55 4.65 -4.25 521.00 11.45 2.53 7 7.40 10.60 4.95 -2.95 141.00 11.04 3.29 7 7.70 8.80 5.10 -2.50 511.00 10.53 3.60 7 4.05 1.65 4.05 1.65 300.00 4.24 4.24 Transient: 22d2 100 6.80 9.70 3.45 -11.35 521.00 10.06 -1.30 100 6.90 8.80 4.10 -3.00 511.00 9.77 2.51 100 6.25 9.25 4.10 -2.80 114.00 9.34 2.59 100 6.55 7.65 4.20 -2.40 511.00 8.94 2.83 100 3.35 1.35 3.35 1.35 300.00 3.50 3.50 ,

1 I

i l

Prepared by: L.T. Hill Date: 12/15/94 Reviewed by: D.E. Killian Date: 1/25/95 Page 24 l l

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B&W NUCLEAR TECHNOLOGIES 32-1236235-00 i t

i 4.7 Material Properties 4.7.1 ' Fracture Tougimess. l According to the applicable weld data sheet and procedure qualification' (Atachment 1),' weld WP-15 ,

can be conservatively evaluated using a RT. of 10'F.  !

The fracture toughness'of weld WP-15 was assumed to follow the lower bound ASME Section XI fracture toughness curves (Figure A-4200-1)' as follows':

Ku = 26.78 + 1.233 exp [0.0145 (T-RT + 160)]-

.? and Ku = 33.20 + 2.806 exp [0.02 (T-RT + 100)] (Temp in *F and Ku,Kuin ksVin) where T is the crack tip temperature.  ;

Because the upper shelf is not quantified by these equations, a cut-off was imposed at 200 ksVin.  !

4.7.2 Material Strength i

The material yield strenFth is utilized in this analysis for purposes of determining the flaw shape parameter, Q. (this equation contains a small-scale plasticity correction). Also, the material flow stress (taken as the average of the yield and ultimate strengths) is used in the limit load calculations. .

In order to ensure the highest plasticity correction (and hence the largest applied stress intensity fac- l tor), the lowest material yield strength is desired. Hence, the material yield and ultimate strengths taken at 600*F are 26.6 ksi and 70.0 ksi, respectively." Thus, the corresponding flow stress is 48.3 ,

ksi.

r r

!t

.a i

i Prepared by: L.T. Hill Date: 12/15/94 i Reviewed by: D.E. Killian Date: 1/25/95 Page 25 i

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B&W NUCLEAR TECHNOLOGIES 32-1236235-00

- 5.0 Procedure Used in Analysis Based on the information given in the previous sections, two FORTRAN programs were created. The first program takes the required stress and crack tip temperature information from Ref. 6 and creates s

the spectrum history required.

The first program SORT.P. takes the opening mode stress for each PV and converts into membrane and bending components. Next, the stress intensity factor is calculated at the point nearest the surface .

(Point 1) for the initial flaw size for each PV in a transient. Finally, the fatigue cycles are completed by matching the most damaging combination of PVs (matching of the PV with the highest stress intensity factor with the PV having the lowest stress intensity factor). The next largest cycle possible is then formed by using the remaining PVs available, and so on, until all PVs have been used. The procedure is repeated for each of the transients studied. In addition, for each fatigue cycle created the crack tip temperature for the maximum stress state is given (for use in determining the fracture tough-ness of the weld material) and the number of times that the fatigue cycle is expected to occur in the design life of this component are included with the fatigue pair. The program listing can be found in Appendix A, with the program output summarized in Table 4-2 and written to fiche (FRAC 2.INP).

The second program, BURIED.F, calculates the fatigue crack growth based on the required number of cycles per fatigue group, stress level and flaw size as previously discussed. This phase of the pro-gram is schematically shown in Figure 4-3. Since the actual order of the spectrum is unknown, when checking the fracture toughness margins per IWB-3612 and limit load margins after the fatigue crack growth analysis is performed, it is important to check the maximum stress state (and the associated crack tip temperature) within a fatigue group to ensure the margins are satisfied. Recall that the fracture toughness is a function of crack tip temperature, hence the higher temperature, the higher the fracture toughness. This second phase of BURIED.F is schematically shown in Figure 4-4, with the program listed in Appendix B. The output for this program is summarized in Table 4-3 and is written to fiche (BURIED 2.OUT).

It should be noted that the Emergency / Faulted conditions are evaluated by assuming the maximum stress state within a fatigue group is completely membrane (i.e. E/F membrane stress = N/U mem-

'rane + N/U bending). This is an extremely conservative approach of estimating the severity of emergency / faulted conditions.

Verification of SORT.F and BURIED.F can be found in Appendix C.

Prepared by: L.T. Hill Date: 12/15/94 Reviewed by: D.E. Killian Date: 1/25/95 Page 26

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.*' .-. C ,

BhW NUCLEAR TECHNOLOGIES : 32-1236235 00 ,

i tegna nuelasw shee a),e LI r

Inina Peugue one 4 cyass esses tidennesen 4

L cornpute 8K(e#Mensa(sp> Knen(ep) -

8K(e.,c, - : pse).Knen( v e)

Canoues dep-co(dK(apra de. pet-Co(dK(s.pse)) n i

4 a = a+ (dui1+ de.ssays

. . . +(d Int . duer tasioyem m No  ;

Fassus Gee k

Yes No Lasirengue one .,

I Figure 4-3 Proccdure Used in BURIED.F in Calculating Fatigue' Crack Growth of N/U Transients ,

i Prepared by: L.T. Hill Date: 12/15/94 Date: 1/25/95 Page 27 Reviewed by: D.E. Killian

,p 4:

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BkW NUCLEAR TECHNOLOGIES 32-1236235-00 Input Mae IWU Seese Stees and Assoshnsd Omsk TD Tomeunwo  !

wom essemene i

sunnes e sea nesamyrenn e:Peeas1 anda f

i 1.#1 als E/F SWees % Peeler el Peniis I and I ese== w-eemme mem mees Evolueen NN Umit Land i

Evalueen E/F Unit Land e=== e - esame men e===l l i

Pee resan teuenness and uma tand masine  ;

t

- tau ressa ome l t

Figure 4-4 Procedure Used in BURIED.F for Determination of Fracture Toughness and Limit Load Margins for N/U and F/F Conditions (Based on flaw dimensions after Fatigue Crack Growth) b k

Prepared by: L.T. Hill Date: 12/15/94 Reviewed by: D.E. Killian Date: 1/25/95 Page 28

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,).

32-1236235-00 B&i # NUCLEAR TLCliNOLGGIES 6.0 Results and Conclusions

' T-ble 4-3 shows the results' of this aralysis. A small amount of fatigue crack growth is expected to '

occur over the design life ci shis component (-0.0026 in.) Furthermore, fracture toughness margir do not exceed theV10 for Normal / Upset conditions andV2 for the Emergency / Faulted conditions.r.so, i the limit load raargin was found to be less than 1 for expected loading conditions. Hence, the CR-3 WP-15 surge nozzle-to-lower head weld flaw indication has been found to be acceptable by IWB-3612 .

criteria for continued safe operation until end-of-life. l Table 4-3 Results of Fracture Mechanics Assessment of the WP-15 Weld Flaw Evaluation Fatigue Crack Growth Analysis Cumulative a e Transient Fatigue (in.) (in.)

Cycles

0. .31000 .85500 350. .31013 .85503 hulal 606. .31020 .85505 hulc2 2091. .31045 .85510 hula 3 t 3077. .31062 .85514 hula 4

, 7961. .31142 .85531 hula 5 9481. .31166 .85537 cc.lb1 16't81. .31200 .85544 edib2 17721. .31200 .85544 2a 20601. .31205 .85546 2b 68601. .31231 .85555 3 116601. .31254 .85563 4 .

117221. .31254 .85563 7 117301. .31254 .85564 8a i 117551. .31254 .85564 8b 117621. .31254 .85564 Sc .

117661. .31254 .85564 8d 137661. .31254 .85564 14 171661. .31254 .85564 20b 171666. .31254 .85564 20d2 171681. .31254 .85564 22:1 171691. .31254 .85564 22bl  ;

171701. .31254 .85564 22cl 171715. .31255 .85564 22dl 171736. .31255 .85564 22a2 171946. .31256 .85564 22b2 17199. .31256 .85564 22c2 172401. .31258 .85565 22d2 Prepared by: L.T. Hill Date: 12/15/94 Date: 1/25/95 Page 29 Reviewed by: D.E. Killian l

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B&W NUCLEAR TECHNOLOGIES 32-1236235-00 Table 4-3 (continued)

Fracture Tonehness and Limit lead Martins For Max Stress intensity Factor in Stress Pair Craa IT Margin FT Margin Tip Point 1 Point 2 LL Margin Temp N/U E/F N/U E/F N/U E/F Transient 521.00 8.076 5.272 9.156 5.281 .490 .754 hulal ,

427.00 9.558 6.328 10.778 6.339 .432 cs59 hula 2 521.00 12.056 8.265 13.424 8.279 .357 .530 hula 3 513.00 11.902 8.151 13.259 8.164 .361 .537 hula 4 513.00 11.902 8.151 13.259 8.164 .361 .537 hula 5 521.00 9.IC/ 6.023 10.367 6.033 .447 .684 cd1bl 521.00 14.229 9.282 16.136 9.297 .311 .480 cdlb2 504.00 15.442 11.513 16.717 11.531 .287 .395 2a 511.00 14.700 10.637 16.068 10.654 .300 .425 2b 511.00 14.025 9.973 15.420 9.989 .313 .450 3 502.00 17.541 13.706 18.717 13.729 .255 .336 4 504.00 15.943 11.928 17.241 11.947 .278 .382 7  ;

504.00 16.843 12.602 18.213 12.623 .265 .36' 8a 504.00 15.379 11.513 16.627 11.531 .288 .395 8b 502.00 16.251 13.007 17.218 13.028 .275 .353 8e 502.00 17.227 13.986 18.178 14.009 .261 .330 8d 504.00 17.889 13.349 19.348 13.370 .250 .344 14 502.00 19.753 15.219 21.166 15.244 .228 .304 20b 504.00 18.036 13.436 19.530 13.458 .248 .342 20d2 177.00 12.408 9.329 14.246 9.344 .304 .477 22al 177.00 15.919 11.648 18.479 11.667 .245 .391 22b1 141.00 9.942 12.000 11.177 12.019 .251 .380 22cl 114.00 8.815 13.928 9.941 13.950 .216 .330 22dl 177.00 12.573 9.423 14.455 9.438 .301 .473 22a2 177.00 16.067 11.717 18.678 11.736 .243 .389 22b2 521.00 16.820 11.250 18.897 11.268 .267 .404 22c2 1 521.00 19.813 13.176 22.307 13.197 .229 .349 l

22d2 Prepared by: L.T. Hill Date: 12/15/94 Reviewed by: D.E. Killian Date: 1/25/95 Page 30

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i BalW NUCLEAR TECHNOLOGIES 32-1236235-00 7.0 References 1

1. ASME Boiler ar:d Pressure Vessel Code,Section XI,1983 Edition, July,1983.

2. BWNT Document 32-1179379, " Lower Loop Pressurizer Surge Nozzle Analysis," NSS-3.

3. Bloom, J.M., " Assessment of Defects and Design of Components Allowing for De fects,",RDD:89:1012-02-01:01, May 1989 Alliance, OH.

4. Cipolla, R.C., FAA-EPRI-75-4-3, April 1975.

5. Stuckey, K.B. , Internal BWNT Memo, Dated 10 November 1994 (Attachment 2).

6. BWNT Document 32-1235087-00, "CR-3 PRZ Surge Nozzle Stresses", November 1994.

7. Anderson, T.L.,

• Fracture Mechanics: Fundamentals and Applications", CRC Press, Boca Raton, FL,1991.

8. BWNT Document, "CR-3 Pressurizer Surge Nozzle Flaw Evaluation", May 1994.

9. Florida Power Corporation Microfilm Card Numbers IX4054 through IX04061, " Manufacture Data Report for Pressurizer", Microfilm Cards of Original Construction Records, Information Supplied by Bob Reynolds of FPC (Attachment 1).

10. ASME Boiler and Pressure Vessel Code, " Case 1332-4 Requirements for Steel Forgings, Sec-tion 111", Approved by Council, December 7,1966.

11. ASME Boiler and Pressure Vessel Code,Section III,1965 Edition (with addenda through Sum-mer 1967).

Reference 9 is not available for entry into the BWNT Record Center but may be referenced as design input as permitted by BWNT-0402-01, Appendix 2.

Project Manager Approval: A /Z .

R.L. Black Prepared by: L.T. Hill Date: 12/15/94 Reviewed by: D.E. Killian Date: 1/25/95 Page 31

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' B&W NUCLEAR TECHNOLOGIES - 32-1236235-00 I

Appendix A - SORT.F - FORTRAN Program Listing used to Create Fatigue Spectrum History l i

l This appendix contains proprietary information.  !

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B&W NUCLEAR TECHNOLOGIES 32-1236235-00 ,

t Appendix B - BURIED.F - FORTRAN Program Listing Used to Evaluate Fatigue Crack Growth of the'dP-15 Weld Flaw and to Evaluate the Required Fracture Toughness and Limit Load Margins i

This appendix contains proprietary information.

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Appendix C - FORTRAN Programs Verification This appendix contains a set of calculations used to verify the FORTRAN programs listed in Appendi-ces A. SORT.F, and B, BURIED.F. It should be noted that the program listings contained in Apprendices A and B also contain numerous comment cards explaining the various program commands and routines. By reviewing the comment cards and the FORTRAN statements that follow, it is possible to vedfy that the intended procedun has been satisfied.

Feature Used in Both SORT.F and BURIED.F Stnss Intensity Factor Input Parameters: a = 0.31 in.

1 = 1.70 in.

t = 4.75 in, c = 0.855 in.

agu = 25.9 ksi a,,,,, = 0.05 ksi a w= 0.05 ksi Point = 1 (point on crar.k closest to surface) lland Calculation:

M. (e/t a/t,Pt!) = 1 + 0.5948 (a/t)2 + 0.4812 (a/t)' + 0.3%3 (a/t)' + 0.3354*(a/t)' +

0.303* (a/t/(1-c/t))' /[(1-e/t-a/t)5]

= 1.0164 M. (c/t,a/t, Pt!) =.84086850 + (E/T*(1.509002 + E/T*(.12940970 *A/T-0.60377800)

+ A/T*( .7731469 + .04428577*A/T) ) + A/T*(0.8841685 .07410377*A/T)

.8338377)*(1.0/(1.0-E/T-A/T)5)

= 0.4366 Q = 1 +4.593*(a/l)**1.65 - 1./6.*(( a ph+a#/agu)2

= 1.277 K = (a M. + awM.) (ra/Q)5

= 0.06344 ksh/in Summary:

Hand Calc BURIED.F SORT.F Mm 1.0164 1.0163667 1.0163667 Mb 0.4366 0.4365798 0.4365798 l Q 1.277 1.2770770 1.2770770 l K_ptI (ksi in5) 0.06344 0.0634400 0.0634400 l

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l Prepared by: L.T. Hill Date: 12/15/94 Reviewed by: D.E. Killian Date: 1/25/95 Page 34

7 32-1236235-00 l B&iV NUCLEAR TECHNOLOGIES Feature Used in SORT.F l

Breaking Inside and Outside Surface Longitudinal Stresses into Membrane and Bending Compo-l nents input Parameters: Stresses taken from PV #1, transient: hulal a.,,, = 0.0 ksi a.,,, = 0.1 ksi Hand Calculation:

a.,,, = (a ,, + a.)/2. = (0.1 + 0.0)/ 2 = 0.5 ksi au = (a.,. - a,,,,)/2. = (0.1 - 0.0)/ 2 = 0.5 ksi This corresponds exactly to the SORT.F output file, FRAC 2.INP.

Creation of Fatigue Cycles Based on Sorting of PVs {

l Table 4-2 shows the pairing of PVs based on SIFs for each transient. This table confirms that the l maximum PVs are paired with the minimum PVs for each transient.

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Egggts__Qsed in EURIED.F Limit Load Solution input Parameters: a_Il = 0.6243702173 in.

c= 0.85 in, t = 4.75 in.

aw = 52.8 ksi a,,,,, = 13.95 ksi Max stress state in Transient hulal aw = 21.75 ksi Hand Calculation:

a = (a/t) / (1 + t/c)

A = a. / a, om=3aw (1-a)2/ (x+(x + 9(g_,)2)ts

= 31.131 ksi Computer output for this Limit Load Solution = 31.13137195 tsi i

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,4 BsW NUCLEAR TECHNOLOGIES 32-1236235-00 Fracture Toughness Calculation input Parameter: ct_ temp = 521T Hand Calculation:

Ku = 33.2 + 2. 806*exp(.02 *(ctemp-10 + 100)) = 569103.91 ksi/in Ku = 26.78 + 1.233 *exp(.0145 *(ctemp-10 + 160)) = 20748.76 kslVin Computer output for the fracture toughness is the same. Note, however, that because the upper shelf is not quantified by these equations, a cut-off was imposed at 200 ksiVin.

Fatigue Crack Growth for One Cycle Input Parameters:

initial crack depth = 0.3100 in.  :

Point 2 Kpt2_ max = 19.10221035535215 Kpt2_ min = -7.20950034447816 Note: R ratio < 0, therefore, Kpt2_ min = 0.0 dkpt2 = 19.10221035535215 Point 1 Kpt1_ max = 21.93384431786706 Kptl , min = -10.5738326272854 Note: R ratio < 0, therefore, Kpt1_ min = 0.0 dkpt1 = 21.93384431786706 Hand Calculation: Computer Output Determine crack growth in 1 Fatigue Cycle BURIED.F output opt 1 = 2.67E-11(21.9338)S5 = 2.65E-6 in. 2.6515806707%285E-06 dpt2 = 2.67E-1 f(19.1022)'* = 1.58E-6 in. 1.584268698316173E-06 da = (dptl+dpt2)/2 = 2.12E-6 in. .0000021179246845 de = (dpt!- dpt2)/2 = 5.34E-7 in. .0000005336559861 new a = a + da = 0.31000212 in. .3100021179246845

• new e = c + de = 0.855000534 in. .8550005336559861 l

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B&W NUCLEAR TECHNOLOGIES 32-1236235-00 Appendix D - Microfiche ,

File Name Daig Descriotion  ;

fat.inp 1/25/95 Stress results from Reference 6 for the nozzle-to-head region. The i streses include all PVs for each N/U transient. The stresses include tle effect of surge line stratification.

frac 2.inp 1/25/95 Output of SORT.F. Also, serves as input for BURIED.F. ,

buried 2.out 1/25/95 Results of Fatigue Crack Growth and the assessment of the final crack size based on IWB-3612 Fracture Toughness Margins for N/U and E/F conditions. In addition, the limit load condition is checked.

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B&W NUCLEAR TECHNOLOGIES 32-1236235-00 Attachment 1 - Letter from R.B. Reynolds to E.E. Organ Showing Fracture Toughness Data of WP-15 Weld 9M Power CORP 0AATioM B&W Nuclear Technologies November 30, 1994

$NE594-0346 3315 Old Forest Road Lynchburg, VA.'24506-0935 Attention: Mr. E. E. Degan

Subject:

Pressurizer Surge Nozzle to Lower Head Weld Data

Reference:

Pressurizer Manufacturers Data Report FPC Construction Records in Microfiche Card Files 1XO4059, 1XO4060, and 1XO4061 Mr. Organ:

Due to potential problems with legibilit of including the actual weld the data sheets in the surge nozzle flaw anal sis (BWNT 32-1235116-00)dures following data was extracted from the we d data sheets and proce qualifications for WP-15. the weld between the surge nozzle and lower head. Two weld data sheets were located in the Referenced FPC records 4 1.e. WP-15 and WP-15 Alt. 1.

WP-15 identifies Quality Control Specification W-54 " General Specification for Automatic and 5tmi-Automatic Cas Metal Arc Welding of Vessels for Special Products or Commercial Nuclear Applications". The Procedure Qualification number is70A1 PQ-1085 which flux core filleridentifies wire withthe 45 weld CFH material as 7/64" diameter (McKay) heat was 150 degrees F with a maximum CO2 shielding gas. The minimum pre interpass temperature of 500 degrees F. The postweld heat treatment was 1100-1150 degrees F for a minimum of 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. The Charpy V-notch was performed at +10 degrees F with 240 ft-lbs energy applied. The absorbed energy is documented to be 37 ft-lbs. 34 ft-lbs, and 50 fr-1bs.

WP-15 Alt. 1 identifies Quality Control Specification W-50 "Ceneral Specjfication for Manual Metal Arc Weldino of Vessels for Nuclear or Spec 1a1 Applications". The Procedure Quilffication number is PQ-1887A which identifies the weld material as ASTM A316 E7015/7016/7018. The minimum preheat was 150 degrees F with a maximum interpass temperature of 500 degrees F. The postweld heat treatment was 1100-1150 degrees F forF a minimum of 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. The Charp with 240 ft-1bs energy applied.y V-notch was performed at +10 degreesThe abs 84 ft-1bs, 86 ft-lbs, and 212 ft-lbs.

If there are any questions, please call.

R. B. ynolds. Site Nuclear Engineering Services y

A. Petrowsky. Supervisor, Site Nuclear Engineering Services cc: Record Management 15780 W. rower Une street Cytal RW + Fioaos 34425-6708 * (904) T95-6488 A f1onde Progress Company Prepared by: _L.T. Hill Date: 12/15/94 Reviewed by: D.E. Killian Date: 1/25/95 Page 38

,,' i s-B&W NUCLEAR TECHNOLOGIES 32-1236235-00 Attachment 1 . ernal BWNT Memo from K.B. Stuckey to L.T. Hill, dated 11/10/1994 (Gives detailed information concerning the WP-15 Weld) 1 l

l BGW NUCLEAR BWTECHNOLOGIES To BWNS 205538 5 00/89) 1.ance Hill From Customer Ken Stuckey or File CR3 PZR Subj. Date Surge Nozzle to Pressurizer Attachment Weld November 10,1994 Refernce: 1) BWNT Drawing 021185600E00 Schematic Drawing "620-0007-59 Pressurizer".

2) BWNT Drawing 135433E 2 Lower Head Assembly abd Details *620-0007 59 Pressurizer".

Based on review of Canton Manufacturing and Quality Assurance Records, and the references,! have ceincluded the following relevant to the subject:

PWHT (Stress h.hf) was performed at 1100*F to 1150'F for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> total.

The initial weld was performed using the Semi Automatic Gas Metal Arc welding process with fabrie.ated McKay E70A1 flux cored weld wire (WP 15 Rev 4).I was unable to readily detstmine other specific information relative to these consumables.

Weld repairs were performed at various times using the Shielded Metal Arc welding process using E7015 A1 electrodes (B&W manufactured) or E7018 A1 electrodes (WP 15 Alt 1 Rev 31.

Please advise if more information is needed on this matter.

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