ML17228B235

Fm Analysis of St Lucie Pressurizer Instrument Nozzle.

ML17228B235
Person / Time
Site:	Saint Lucie
Issue date:	07/11/1995
From:	NANA A D, YOON K K BABCOCK & WILCOX CO.
To:
Shared Package
ML17228B236	List: L-95-220, Forwards Suppl to ISI Plan Second ten-yr Interval Revised Stress & Fracture Mechanics Evaluations Pressurizer Instrument Nozzles ML17228B235 ML17228B237
References
32-1235128-02, 32-1235128-2, NUDOCS 9508100179
Download: ML17228B235 (35)

Text

BNT-20697-2 (11/B9)(BNHP.20697.1)

Il jBBMI NUCI.EAR%M SERll!CE COMPANY CALCULATION"

SUMMARY

SHEET (CSS)DOCUHENT IDENTIFIER 32-1235128-02 FM Ana 1 sis of St Lucie Pressurizer Instrument Nozz 1 e PREPARED BY: Ashok D.Nana COST CENTER 41020 REFT PAGE(S)SIGNATURE TITLE Princi al En ineer REVIENEO BY: Kenneth K.Yoon'IGNATURE/A/01$7//F95 TTTRE Technical nenltant 01$7 II'H STATEHENT:

REVIENER INDEPENDENCE PURPOSE AND SUHHARY OF RESULTS: Purpose To provide a bounding flaw evaluation for the six 1" instrument nozzles located in the spherical heads of the pressurizer.

The evaluation will consider a conservative flaw size and will determine the acceptability of the postulated bounding flaw for the forty year design life of the plant (30 future years).This flaw evaluation will be performed in accordance with IWB-3612 of Section XI, ASME Boiler and Pressure Vessel Code.Summary of Results The postulated flaw size of 0.875 inches in the instrument nozzles (6)of the spherical heads of the St.Lucie Unit 2 pressurizer was found to be acceptable for the design life of the plant, per IWB-3612 of the ASME Code Section XI.***BWNT NON-PROPRIETARY

• • • THE FOLLOJING COHPUTER CODES HAVE BEEN USED IN TNIS DOCUMENT: CODE/VERSION/REV CODE/VERSION/REV THIS DOCUHENT CONTAINS ASSUHPTIONS THAT HUST BE VERIFIED PRIOR TO USE ON SAFETY-RELATED IIORK"-'P508100179-950802 PDR ADOCK 05000389

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• • • 32-1235128-02 RECORD OF REVISIONS Revision 00 01 02 Descri tion of Revision Original Release Issue of"Non-Proprietary" Version Re-analysis considering only the instrument nozzles (6)located in the spherical heads and.using fracture toughness value of 200 ksiV i n Date Released 12/94 7/95 Prepared by: A.D.Nana Reviewed by: K.K.Yoon D:~JI 995 D':~JI 1995 Page 2 of 29 BA&Nuclear Technologies

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• • • 32-1235128-02 TABLE OF CONTENTS Page EXECUTIVE

1.0 INTRODUCTION

1.1 Assumptions

2.0 DESIGN INPUTS~~~~I 7 3.0 GEOMETRY, FLAW SIZE AND ORIENTATION..........

3.1 Geometry of Bounding Pressurizer Nozzle Penetration...

3.2 Flaw Size and Orientation

.......9 4.0 MATERIAL TOUGHNESS...13 5.0 LOADING CONDITIONS/STRESSES

~~5.1 Normal and Upset Loading Conditions

~~~~~~0~~~~14~~~~~~~~14 5.2 Emergency and Faulted Loading Conditions...........

16 6.0 FLAW EVALUATION.....................

'.~..'...............

17 6.1 Flaw Evaluation for Normal and Upset Loading Condition Loads 6.2 Flaw Evaluation for Emergency and Faulted Condition Loads.....

18 20

7.0 CONCLUSION

S 27

8.0 REFERENCES

28 Prepared by: A.D.Nana Reviewed by: K.K.Yoon Date: June 1995 Date: June 1995 Page 3 of 29 B&W Nuclear Technologies

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• • • 32-1235128-02 EXECUTIVE

SUMMARY

During the 1994 refueling outage external leakage was identified at the pressurizer instrument nozzle"C" of Florida Power&Light Company's St.Lucie Unit 2.Subsequent NDE identified indications on the J-welds for three of four steam space instrument nozzles.Modifications were made and justifications perforined to determine the potential for crack growth during plant operation.

The evaluation performed at the time was conservatively limited to one fuel cycle.The purpose of this evaluation was to justify acceptability of indications in the J-weld for the six 1" instrument nozzles in the pressurizer for 30 future years of plant life.The six nozzles are located in various regions of the pressurizer and are horizontally and vertically oriented.Four of the instrument nozzles are horizotally oriented and contained in the pressurizer head steam-space region.The remaining two nozzles are vertically oriented and located in the lower head of the pressurizer.

A detailed finite element stress analysis was performed that accounted for all six nozzle penetration regions.The stress analysis considered and evaluated all significant design transients in the evaluation.

The most significant transient produced maximum tensile stresses in the inside of the pressurizer shell at the nozzle penetration region (J-weld location)~For the normal and upset condition category, the maximum tensile stress (hoop)was developed during an upset condition reactor trip transient (loss of load transient).

This transient was conservatively evaluated for 375 cycles to bound all future cycles of plant heatup/cooldown.

For the emergency and faulted condition, the loss of secondary pressure transient was evaluated since the significant cooldown during this transient produced maximum tensile stresses at the J-weld location.The fracture mechanics analysis postulated a nozzle corner flaw with a conservative flaw size and determined its acceptability for thirty future years of plant life.A nozzle corner flaw with an initial flaw size of 0.875 inches was postulated in the analysis.The flaw size is considered to bound the structural and buttering weld depth around the nozzle area.A fatigue flaw growth Prepared by: A.D.Nana Reviewed by: K.K.Yoon Date: June 1995 Date: June 1995 Page 4 of 29 B&W Nuclear Technologies

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• • • 32-1235128-02 analysis was performed for the normal and upset condition loads.Considering all the applicable design transients, the initial postulated flaw size of 0.875 inches in the instrument nozzle of the St.Lucie pressurizer was determined to reach a final flaw size (af)of 0.966 inches at the end of the design life of the plant.The maximum applied stress intensity factor at the final flaw size is 46.42 ksiV i n and results in a safety factor of 4.31.This safety factor is greater than the required safety factor of 410 (3.16)per IWB-3612(a) of ASME Code Section XI.For the emergency and faulted condition, the maximum applied stress intensity factor at the final flaw size is 84.6 ksiV i n and results in a safety factor of 2.36.This safety factor is greater than the required safety factor of V2 (1.414)per IWB-3612(b) of ASME Code Section XI.Therefore, it is concluded that the postulated flaw size in the instrument nozzle of the St.Lucie pressurizer is acceptable for the design life of the plant (thirty future years)per IWB-3612 of the ASME Code Section XI.Prepared by: A.D.Nana Reviewed by: K.K.Yoon Date: June 1995 Date: June 1995 Page 5 of 29 B&W Nuclear Technologies

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• • • 32-1235128-02

1.0 INTRODUCTION

The purpose of this analysis is to provide a bounding flaw evaluation.

for six of the seven instrument/temperature 1" nozzles in the pressurizer.

All the six nozzles are located in the spherical heads of the pressurizer.

The evaluation will consider a conservative flaw size and will determine the acceptability of the postulated bounding flaw for thirty future years of plant life.This flaw evaluation will be performed in accordance with IWB-3612 of Section XI, ASME Boiler and Pressure Vessel Code.1.1 Assumptions a.A nozzle corner flaw with an initial fiaw size of 0.875 inch is postulated'n this analysis.This flaw size is considered to bound the structural and buttering weld depth around the nozzle area.b.It is assumed that the postulated flaw covers the entire SMAW I-182 weld region so that primary water stress corrosion cracking (PWSCC)is no longer active for the pressurizer.

c.Three hundred and seventy five future cycles of heatup/cooldown are conservatively assumed for the remaining design life of the plant.d.Eight future cycles of pressure tests at 10%of the operating pressure (2475 psia)are assumed over the next 30 years.Prepared by: A.D.Nana Reviewed by: K.K.Yoon Date: June 1995 Date: June 1995 Page 6 of 29 BAW Nuclear Technologies I***BWNT NON-PROPRIETARY

• • • 32-1235128-02 2.0 DESIGN INPUTS a)Geometry of Pressurizer Nozzle Penetrations The penetration configuration of the pressurizer upper head steam space instrument nozzles (four)with the modified nozzle design is contained in Drawing 2998-19321 of Reference 1.The penetration configuration of the pressurizer bottom head (two)instrument nozzles is contained in Drawing 2998-18709 of Reference 2.minimum pressurizer head thickness=3.875 in b)Design Transients/Number of Cycles The following information was taken from Reference 3, with the transient specific information from Reference 4 (for the forty year design life of the plant).i)500 cycles of normal heatup/cooldown for the design life of the component.

The normal operating pressure per Table 5.4-6 of Reference 3 is 2250 psia.ii)A total of 480 cycles of upset condition transients.

The maximum pressure range during upset condition transient is 660 psi and occurs between 2400 psia (abnormal loss of turbine generator load)and 1740 psia (reactor trip transient) with associated temperature difference of 50'F during loss of load transient (Reference 4).iii)200 cycles of leak test at 2250 psia (Reference 4)iv)The remainder of the normal operating transients i.e.15,000 cycles of power change cycles from 15%to 100%power, 2,000 cycles of step power changes of 10%of the full load and 1 x 10'ycles of normal variations of 100 psi and temperature differences of less than 20'F (Reference 4).v)5 cycles of emergency condition transient (complete loss of secondary pressure transient), given in Reference 4.Since the analysis was performed for 30 future years, only 75%of the above number of cycles for a given transient were considered in the evaluation.

Prepared by: A.D.Nana Reviewed by: K.K.Yoon Date: June 1995 Date: June 1995 Page 7 of 29

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• • • 32-1235128-02 c)Materials The pressurizer head and shell material is made of SA-533 Grade B Class 1 per Reference 1 and Addendum 2 of Reference 4.Per Table 5.2-9 of Reference 5, the RTpy of the pressurizer shell material is 10'F.d)Applicable ASME Section XI Code Per Reference 6, the applicable ASME Section XI code is 1989 Edition.Prepared by: A.D.Nana Reviewed by: K.K.Yoon Date: June 1995 Date: June 1995 Page 8 of 29 7I B&W Nuclear Technologies

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• • • 32-1235128-02 3.0 GEOMETRY, FLAW SIZE AND ORIENTATION 3.1 Geometry of Bounding Pressurizer Nozzle Penetration There are six 1" instrument nozzles in the pressurizer of St.Lucie Unit 2 as depicted by the drawing of Reference 2.Four of the instrument nozzle are contained in the pressurizer upper head steam space region.These nozzles are horizontally oriented in the lower spherical part of the upper head as illustrated in Figure 1.The remaining two instrument nozzles are located in the lower region of the pressurizer as illustrated in Figure 2.These nozzles are vertically oriented and located in the lower head of the pressurizer.

The minimum wall thickness of the upper and the lower spherical heads is 3.875 inches.The stress analysis of Reference 7 took each of the six nozzle penetration regions in the spherical heads into consideration and constructed a nozzle penetration finite element model to bound all six instrument nozzle locations.

For additional details refer to Section 3.3 of Reference 7.3.2 Flaw Size and Orientation It is postulated that there exists a nozzle corner flaw (as depicted in Figure 3)with an initial depth equal to the structural and buttering weld depth around the nozzle area.Therefore, a flaw size of 0.875 inches is assumed.The orientation of this flaw was assumed to be in the x,y plane (see Figure 3)which is normal to the hoop direction.

This is the worse case flaw orientation since the maximum stress is primarily due to pressure induced hoop stress as can be seen from the results of the stresses along the flaw plane in Section 6.0 of Reference 7.The analysis will evaluate maximum stress intensity factor and perform fatigue flaw growth analysis based on consideration of all crack front angles i.e.from 6 equal to 0 degrees (vessel side)to the 45 degree flaw plane to 90 degrees (nozzle bore side).Prepared by: A.D.Nana Reviewed by: K.K.Yoon Date: June 1995 Date: June 1995 Page 9 of 29

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• • • 32-1235128-02 Figure 1: Upper Pressurizer Region Ul Qv Z I Eh r~~~m m>r 4g P>>/g x m SE'he.R.(65 C~Prepared by: A.D.Nana Reviewed by: K.K.Yoon Date: June 1995 Date: June 1995 Page 10 of 29 I I I I~~I~~~~I PEI~'~.~I'~~~~~.~.~~~~~~

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• • • 32-1235128-02 Figure 3: Nozzle Corner Flaw Q pter)n/RL Nozzt.6 PgasSua<<~~HGRb e-'/I POS g~g A l 6'D Nozzle CoRNEP, FLAW Zhl$7g UPTON!4T hlozz~E.C, I o.ld~wg x,y coordinates in the plane of the crack 8 45 degrees Prepared by: A.D.Nana Reviewed by: K.K.Yoon Date: June 1995 Date: June 1995 Page 12 of 29 BdkW Nuclear Technologies

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• • • 32-1235128-02 4.0 MATERIAL TOUGHNESS The pressurizer shell and head is SA-533, grade B, class 1 per Reference 1 and Addendum 2 of Reference 4.The RT>>of this material is 10'F.According to IWB-3612, the arrest toughness curve, KI, in Appendix A,Section XI of ASME Boiler&Pressure Vessel Code (Reference 6)was used for this evaluation.

Since the RTN>>of the pressurizer is 10'F, the material is considered to be at the upper shelf region for temperatures above 192'F.Because the maximum stress is primarily due to pressure, the corresponding temperatures during the transient when the maximum stresses occur in the pressurizer shell/head are above 500'F.An upper shelf value of 200 ksiV i n was conservatively used in the analysis.It is noted that any shift due to irradiation is negligible, i.e.no changes in RT>>value of the pressurizer with increases in Effective Full Power Years (EFPYs').Prepared by: A.D.Nana Reviewed by: K.K.Yoon Date: June 1995 Date: June 1995 Page 13 of 29 B&W Nuclear Technologies

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• • • 32-1235128-02 5.0 LOADING CONDITIONS/STRESSES 5.1 Normal and Upset Loading Conditions The stresses due to normal and upset conditions are contained in Section 6.0 of Reference 7.The composite transient evaluated in the analysis consisted of 100'F/hr heatup, 100%power steady state condition, a bounding upset condition transient (represented as a 53'F step-down to pressure of 1740 psia and a 53'F step up with a pressure of 2400 psia)and a 200'F/hr cooldown rate as described in Section 5.0 of Reference 7.The normal and upset condition transient cases are summarized in Table 1.The results of the analysis in Reference 7 showed that the maximum stresses occur during anupset condition step down transient (transient case 2c as given in Table 1).The next largest stress state occurs during steady state conditions when the pressure is 2400 psia (transient case 2a).These maximum stress states occurs at temperatures well above 500'F when the material is at upper-shelf.

Transient case 2c was conservatively evaluated for 375 cycles (from an initial stress-free state to the maximum upset condition), in Section 6.1, to bound the 360 cycles associated with all the upset condition transients as well as the 375 cycles of plant startup and shutdown and 150 cycles of leak tests.In addition, 8 cycles of pressure tests (case 3, Table 1)were evaluated.

During normal cooldown the maximum stress occurs at 595'F (transient case 1c as given in Table 1)when the material is at upper-shelf.

To ensure that the fracture toughness margin (factor of safety os 10 per IWB-3612)is maintained, throughout the entire cooldown transient, the time at the end of the 200'F/hr cooldown is also evaluated (transient case 1d as given in Table 1).At this time, the bulk fluid temperature is at 70'F and maximum thermal stresses are developed in the pressurizer shell/head.

Also, the fracture toughness is low.However, the component is depressurized so that the resulting stresses are not significant as can be seen in Section 6.0 of Reference 7 for this transient case.Adequate fracture toughness margin during the entire heatup/cooldown was demonstrated in Section 6.1.Prepared by: A.D.Nana Reviewed by: K.K.Yoon Date: June 1995 Date: June 1995 Page 14 of 29 B&W Nuclear Technologies

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• • • 32-1235128-02 Table 1: Normal and Upset Condition Transient Cases Transient Category, Case Normal, la Normal, lb Normal, lc Normal, ld Description of Transient Time End of 100'F/hr heatup (max.stress during heatup)100%power steady state Cooldown at 560'F (max.stress during cooldown)Cooldown at 70'F (max.thermal stress during cooldown)Pressure (psia)2250 2250 1472 Temperature

('F)653 653 595 70 Number of Cycles'75 Normal, le Pressure and temperature fluctuations during operation d,P 5 100 b,T<20 765,000 Upset, 2a Upset, 2b Upset, 2c Test, 3 At max.pressure (loss of turbine generator load)53'F step up 53'F step down 110%of operating pressure 2400 2400 1740 2475 653 600-653 653-600 653 360 4 Associated with 30 future years of plant life.Based on considering 75%of the design cycles given in References 3 and 4.This case is not specifically evaluated in Reference 7.Conservatively assumed to be one-half the stresses due to the transient cases 2b and 2c.11,250 cycles of plant loading/unloading, 1,500 cycles of 10%step load increase/decrease and 750,000 cycles of normal pressure variation are conservatively grouped by this transient case.There are only 30 cycles of loss of turbine generator load, however, 300 cycles of reactor trip transient and 30 cycles of loss of primary flow transient are conservatively grouped by this transient case.Prepared by: A.D.Nana Reviewed by: K.K.Yoon Date: June 1995 Date: June 1995 Page 15 of 29 BOW Nuclear Technologies

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• • • 32-1235128-02 In addition to the 375 cycles of plant startup/shutdown (includes 360 cycles of upset condition transients) and 8 cycles of pressure tests there are 11,250 cycles of plant loading/unloading, 1,500 cycles of 10%step load increase/decrease and 750,000 cycles of normal pressure variation (+/-100 psi,+/-7'F)as given in References 3 and 4.Review of these transients show that these transients can be grouped as a single transient with 765,000 associated cycles of maximum pressure variation of 100 psi and temperature variation of less than 20'F (transient case le as defined in Table 1).The associated stress range due to this transient is given in Tables 6-14 and 6-15 of Reference 7.5.2 Emergency and Faulted Loading Conditions The only emergency and faulted condition design transient (pressurizer pressure and temperature versus time)provided in References 3 and 4 is the loss of secondary pressure transient (an emergency condition transient).

The faulted condition transients described in Referenc'es 3 and 4 are;i)those due to safe shutdown earthquake with normal operation at full power and with and without pipe rupture condition and ii)those due to LOCA.However, per Table 3.9-3B of Reference 3, there are no associated cycles for the faulted condition transients.

Therefore, the only transient evaluated (in.Reference 7)for this loading condition is the loss of secondary pressure transient.

During this transient the pressurizer experiences a significant cooldown rate.As a result of this cooldown rate, high tensile stresses at the inside surface of the nozzle corner region can be produced.This is reflected in the stress results given in Section 6.0 of Reference 7 which produced the maximum hoop surface stress amongst all the transients analyzed.This transient case will be evaluated in Section 6.2.There are 4 cycles associated with this transient case.Prepared by: A.D.Nana Reviewed by: K.K.Yoon Date: June 1995 Date: June 1995 Page 16 of 29 B&W Nuclear Technologies

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• • • 32-1235128-02 6.0 FLAW EVALUATION A three dimensional nozzle corner crack is postulated for this analysis.The stress intensity factor, Kfor this flaw geometry is given in Reference 8 and reported below: K,=~iia[0.706A

+0.537(2a/m)A,+OA48(a/2)A+0.393(4a/3n)A]where AAAand A, are the polynomial coefficients of the stress profile expressed as: o(r)=+A,r+A~r+A,r'he three dimensional nozzle corner flaw solution given above is utilized to evaluate the postulated flaw in the one inch pressurizer instrument/temperature nozzles of St.Lucie unit 2.The solution given above is applicable for the 45 degree flaw plane as illustrated in Figure 3.Hence, the stresses are obtained along this flaw plane as illustrated in Figure 6.2 of Reference 7.To address the stress intensity factors at other crack front angles, the information contained in Reference 9 is utilized.Reference 9 has evaluated the stress intensity factors due to pressure induced hoop stresses in a nozzle corner with a quarter circular crack geometry.Three nozzle corner flaw sizes with flaw size to thickness ratios of 0.15, 0.26 and 0.34 were investigated in this study.This study provided the non-dimensional stress intensity factors as a function of the crack front angle, e for each of the three flaw sizes as illustrated in Figure 11 of Reference 9.From this figure it is clear that the stress intensity factor near the surfaces of both the vessel and the nozzle bore side is slightly greater than the stress intensity factor along the 45 degree plane.For the two larger flaw sizes (flaw size to thickness ratio comparable to this evaluation), the stress intensity factor near the surfaces is about 5 to 10 percent higher than along the 45 degree plane.Therefore, to determine maximum flaw growth with consideration of all crack front angles, the stress intensity factors obtained using the above equations will be increased by 10 percent.This is a conservative practice.An initial flaw depth of 0.875 inches is assumed to bound the structural and buttering weld depth around the nozzle area (J-welds).

The postulated flaw in the instrument nozzle is evaluated for normal/upset condition and emergency/faulted condition as given below.Prepared by: A.D.Nana Reviewed by: K.K.Yoon Date: June 1995 Date: June 1995 Page 17 of 29 B&W Nuclear Technologies

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• • • 32-1235128-02 r 6.1 Flaw Evaluation for Normal and Upset Loading Condition Loads As discussed in Section 5.1, the following bounding transient case was analyzed for the normal and upset condition loading.Transient case 2c (reactor trip transient) was evaluated for 375 cycles.The maximum tensile stress state along the flaw plane occurs during this condition when the pressurizer is assumed to cycle from an intial stress-free state (0 psi at 70'F)to a reactor trip transient.

This stress state will be conservatively assumed to occur for all 375 cycles of normal heatup/cooldown.

As a first step, a third order polynomial equation to the stresses from the finite element analysis results was made.The coefficients for the polynomial equation were obtained using a least square flit.The resulting stresses using the polynomial equation agree very well with the finite element model (FEM)stresses as illustrated in Figure 4.The FEM stresses are for the maximum upset condition pressure stress at 2400 psia as given in Table 6.4 of Reference 7.The stress intensity factor, Kfor the initial flaw size of 0.875 inches is: K,(a;)=44.89 ksiV i n A fatigue flaw growth analysis was performed for 375 cycles using the above maximum upset condition stresses as given in Table 2a.The fatigue crack growth rate is: da/dN=C,(b,K,)" where da/dN is the crack growth rate in micro-inch per cycle, b,K, is the maximum K, minus minimum K, (in this case the minimum K, is zero), C, and n are material constants and are obtained from the fatigue crack growth rate curve which is given in Figure A-4300-1 of Reference 6.From this figure, it can be seen that for a surface flaw (water reactor environment) with an R ratio 5 0.25 and b,K, 2 19 ksiV i n, the applicable material constants are C,=1.01 x 10'in/cycle and n=1.95.The flaw size at the end of 375 cycles, is 0.94 inches and the maximum applied K,=46.00 ksiV i n.Also, 8 cycles of pressure tests at 2475 psia were Prepared by: A.D.Nana Reviewed by: K.K.Yoon Date: June 1995 Date: June 1995 Page 18 of 29 B&W Nuclear Technologies

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• • • 32-1235128-02 Figure 4: FEM throughwall stresses versus polynomial fit stresses 45000 For Upset Condition Step Doom Transient FEM stresses Polyriomial-Fit-stresses--

35000 30000 CO D.0 0 25000 15000 0 0.5 1.5 Distance (s), along 45 degree flaw plane Prepared by: A.D.Nana Reviewed by: K.K.Yoon Date: Jurie 1995 Date: June 1995 Page 19 of 29 1

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• • • .32-1235128-02 considered in the analysis.The flaw size at the end of 8 cycles of pressure tests, is 0.942 inches with a maximum applied K,=43.26 ksiV i n.In addition, there are 765,000 cycles of pressure and temperature variations (transient case le of Table 1).The R ratio (K,;JK,~for this case is 0.92 and since b,K, is less than 2.9 ksiV i n, the applicable material constants are C,=1.2 x 10" in/cycle and n=5.95.The fatigue flaw growth due to 765,000 cycles of the above transient is computed using the flaw size after 375 cycles of heatup/cooldown and 8 cycles of pressure tests (0.942 inches)as the initial flaw size.The results are given in Table 2b which shows that after consideration of 765,000 cycles the flaw size is 0.964 inches.In addition, after consideration of 4 cycles of loss of secondary pressure transient as given in Table 4 the final flaw size (af)is 0.966 inches.The maximum applied stress intensity factor at the final flaw size is: K,(af)=46.42 ksiV i n.Since the upper shelf toughness is 200 ksiV i n, this results in a safety factor of 4.31 (as given in Table 3)which is greater than the required safety factor of v 10 (3.16)per IWB-3612(a) of Reference 6.Also, as discussed in Section 5.1, to ensure that the fracture toughness margin is maintained, through the entire cooldown transient, the time at the end of the 200'F/hr cooldown is evaluated (transient case 1d of Table 1).The maximum applied K, at the end of cooldown (70'F)was determined to be 8.69 ksiV i n.The associated fracture toughness, K,~, was obtained from the equation given on Page C-18 of Reference 8.Using this equation, the fracture toughness at (T RTNpr)=70'F-10'F or 60'F is 56.5 ksiV i n.Therefore, there is a safety factor of 6.50 for this condition which is significantly greater than the required safety factor of V'10 per IWB-3612(a) of Reference 6.6.2 Flaw Evaluation for Emergency and Faulted Condition Loads As discussed in Section 5.2, the only emergency and faulted condition transient requiring evaluation is the loss of secondary pressure transient which has 4 cycles associated with it.The results of the analysis are provided in Table 4.This transient occurs following a steady state condition.

The d,K, associated with this transient is 44.84 ksiV i n.The flaw growth associated Prepared by: A.D.Nana Reviewed by: K.K.Yoon Date: June 1995 Date: June 1995 Page 20 of 29 B&W Nuclear Technologies

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• • • 32-1235128-02 with this transient is 2 mils.The maximum applied stress intensity factor at the final flaw size (a,)of 0.966 inches for the emergency and faulted condition is: K~(a,)=84.6 ksiV i n.As previously noted in Section 5.2, the material remains at upper shelf during this transient (K<<=200 ksiV i n).Therefore, this results in a safety factor of 2.36 for the emergency and faulted condition which is greater than the required safety factor of V'2 per IWB-3612 (b)of Reference 6.Prepared by: A.D.Nana Reviewed by: K.K.Yoon Date: June 1995 Date: June 1995 Page 21 of 29 B&W Nuclear Technologies

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• • • 32-1235128-02 Table 2a: Fatigue Flaw Growth Analysis for 375 cycles of normal heatup/cooldown Fatigue Flaw Growth Analysis of Postulated Nozzle Corner Crack ai 0.875 in increment=

0.005 in GEOMETRIC FACTORS GO=1.251 G1=0.606 G2~0.397 G3~0.296 For dK>STRESS AO A1 A2 A3 FACTORS 40.21-12.22-1.53 1.26 dN=C1 m=R=Kmin=Kmax=KI=GO*AO*a (1/2)+G1*A1*a" (3/2)+G2*A2+a" (5/2)+G3*A3*a (7/2)Factor for worse case flaw angle=1.1 19 ksi din 375 cycles 1.01E-07 in/cycle 1.95 0 0.00 ksi*in" 0.5 44.98 ksi*in"0.5 KIA=Safety factor=200 ksi*in" 0.5 3'6 ai (in)aj (in)Fatigue Group 1 KI (~a)C1 (dKI)"m dN ksi din in/cycle cycles Total dN cycles Check KIA/KI Accept 7 0.875 0.880 0.885 0.890 0.895 0.900 0.905 0.910 0.915 0.920 0.925 0.930 0.935 0.875 0.880 0.885 0.890 0.895 0.900 0.905 0.910 0.915 0.920 0.925 0.930 0.935 0.940 0.95 44.89 44.98 45.07 45.15 45.24 45.33 45.41 45.50 45.58 45.67 45.75 45.84 45.92 46.00 46.16 1.69E-04 1.70E-04 1.70E-04 1.71E-04 1.71E"04 1.72E-04 1.73E-04 1.73E-04 1.74E-04 1.75E"04 1.75E-04 1.76E-04 1.76E-04 1.78E-04 29.6 29.5 29.4 29.3 29.2 29.0 28.9 28.8 28.7 28.6 28.5 28.4 28.3 29.6 59.1 88.5 117.7 146.9 175.9 204.9 233.7 262.4 291.1 319.6 348.0 376.4 continue continue continue continue continue continue continue continue continue continue continue continue stop 4.45 4'4 4.43 4.42 4.41 4.40 4.40 4.39 4.38 4.37 4.36 4.36 4.35 4.33 OK OK OK OK OK OK OK OK OK OK OK OK OK OK Prepared by: A.D.Nana Dn'ltl&'l%>Drl hv K K'OOn Date: June 1995 Date;June 1995 Page 22 of 29 BOW Nuclear Technologies

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• • • 32-1235128-02 Table 2b: Fatigue Flaw Growth Analysis (cont'd)for remaining normal operating transients For 8 cycles of pressure tests at 110%-of operating pressure or 2475 psig Stress Factors for 2400 psi are ratioed by (2475/2400) 1.031 STRESS FACTORS For dK>19 ksi~in AO=34.00 dN=8 Al-1.98 Cl~1.01E-07 A2-3.51 m 1.95 A3 1.37 R 0 Kmin=0.00 Kmax 43.24 KZ=Go*AO*a"(1/2)+G1*A1*a"(3/2)+G2*A2+a"(5/2)+G3*A3*a"(7/2)

Factor for worse case flaw angle=1.1 increment 0.001 in ai (in)aj (in)Fatigue Group 2 KE (a j)C1(dKE)"m dN ksi gin in/cycle cycles Total dN Check cycles KlA/KX Accept 7 0.940 0.941 0.941 0.942 43.24 43.26 1.56E-04 1.57E-04 6.39 6.39 6.4 continue 12.8 stop 4.63 4.62 OK OK Prepared by: A.D.Nana Reviewed by: K.K.Yoon Date: June 1995 Date: June 1995 Page 23 of 29 8'Nuclear Technologies

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• • • 32-1235128-02 Table 2b: Fatigue Flaw Growth Analysis (cont'd)for remaining normal operating transients For 765,000 cycles of remaining normal condition transients GO Gl G2 G3 1.251 0.606 0.397 0.296 GEOMETRIC FACTORS STRESS AO Al A2 A3 FACTORS Group 1 32.54<<12.42 6.41-1.17-Group 2 42.80-26.42 13.41-2.70 dN=Cl 765000 1.20E-11 5.95 and R>0.65 0.92 For dK<19 ksi gin KI~GO*AO*a" (1/2)+Gl*A1*a (3/2)+G2*A2*a (5/2)+G3*A3*a (7/2)Factor for flaw angle~increment 0.002 in 1.1 Kmin=43.88 ksi*in"0.5 Step up transient-Kmax=47.95 ksi*in"0.5 Step down transient Factor=0.5 ai (in)aj (in)Group 1 Min KI(aj)ksi<finn Fatigue Group 3 Group 2 Max KI (aj)ksi Jin Delta dKI (a j)Cl (dKI)"M ksi gin in/cycle dN cycles Total dN cycles Check 0.942 0.944 0.946 0.948 0.950 0.952 0.954 0.956 0.958 0.960 0.962 0.944 0.946 0.948 0.950 0.952 0.954 0.956 0.958 0.960 0.962 0.964 38.04 38.07 38.11 38.14 38.17 38.20 38.23 38.27 38.30 38.33 38.36 45.44 45.47 45.50 45.53 45.56 45.59 45.62 45.65 45.68 45.71 45.74 3.70 3.70 3.70 3.70 3.70 3.69 3.69 3.69 3.69 3.69 3.69 2.88E-OB 2.87E-OB 2.87E-OB 2.87E-OB 2.86E-OB 2.86E-OB 2.85E-OB 2.85E-OB 2.85E-OB 2.84E-OB 2.84E-OB 69502.0 69596.2 69691.4 69787.7 69885.0 69983.3 70082.7 70183.0 70284.4 70386.9 70490.3 69502.0 continue 139098.1 continue 208789.6 continue 278577.3 continue 348462.2 continue 418445.6 continue 488528.2 continue 558711.2 continue 628995.7 continue 699382.5 continue 769872.8 stop Prepared by: A.D.Nana Ps vi<w<rl hv'K Ynnn Date: June 1995 Date: June 1995 Pace 24 of 29 B&W Nuclear Technologies

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• • • 32-1235128-02 Table 3: Summary of flaw sizes an check with acceptance criteria for normal and upset condition KIA 200 ksi din Lumped Transients:

Considers 375 cycles of worst upset loads (covers future normal heatup and cooldown)8 cycles of 2475 psig pressure tests as well as 765,000 cycles of remaining normal condition transient cycles ai (in)aj (in)KI(aj)KIA/KI(aj) ksi Pin Transient Group ASME ACCEPT CODE 0.875 0.935 0.941 0.962 0.880 0.940 0.942 0.964 0.966 44.98 46.00 43.26 38.20 46.42 4.45 4.35 4.62 5.24 4.31 Beginning-1 end of 1 end of 2 end of 3 end of all 3.16 3.16 3.16 3.16 3.16 OK OK OK OK OK Prepared by: A.D.Nana Date: June 1995 Reviewed by: K.K.Yoon...Date: June 19)5 Page 25 of 29

B&W Nuclear Technologies

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• • • 32-1235128-02 Table 4: Summary of flaw growth analysis and check with acceptance criteria for emergency and faulted condition For 4 cycles of loss of secondary pressure transient GO~Gl G2 G3.~-1.251 0.606 0.397 0.296 GEOMETRIC FACTORS Ao A3 STRESS FACTORS Group 1 30.91 A1-1.80 A2"3.19 1.25 Group 2 101.23-113.14 56.99-12.44 dN~Cl 4 1.84E-07 1.95 0.47 1.82 For dK v 15 ksi din KI~GO*AO*a" (1/2)+Gl*A1*a" (3/2)+G2*A2~a" (5/2)+G3*A3*a" (7/2)Kmin=Kmax 39.76 ksi*in"0.5 steady state at 2250 psia 84.60 ksi+in"0.5 loss of secondary pressure Factor increment ai (in)for worse case flaw 0.002 in Group 1 Min aj KI(aj)(in)ksi fin angle Group 2 Max KI (aj)ksi Pin Delta dKI (a j)'l (dKI)"M ksi~in in/cycle cycles Total dN cycles Check 0.9640 0.9660 39.76 84.60 44.84 3.06E-04 6.5'.5 stop'Summary of final flaw sizes and check with acceptance criteria KIC 200 ksi Sin For Emergency and Faulted Condition:

ai (in)af (in)KI (af)KIC/KI (af)ksi din ASME ACCEPT CODE 0.9640 0.9660 84.60 2.36 1.41 OK Prepared by: A.D.Nana Reviewed by: K.K.Yoon Date: June 1995 Date: June 1995 Page 26 of 29 Ol>"d B&W Nuclear Technologies

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• • • 32-1235128-02

7.0 CONCLUSION

S Considering all the applicable design transients, the initial postulated flaw size of 0.875 inches in the instrument nozzle of the St.Lucie pressurizer was determined to reach a final flaw size (a,)of 0.966 inches after 30 future years plant life.For the normal and upset condition the maximum applied stress intensity factor at the final flaw size is 46.42 ksiV i n and results in a safety factor of 4.31.This safety factor is greater than the required safety factor of 410 (3.16)per IWB-3612(a)of Reference 6.The analysis considered all crack front angles to determine the maximum applied stress intesity factor and ensure bounding fatigue flaw growth.For the emergency and faulted condition, the maximum applied stress intensity factor at the final flaw size is 84.6 ksiV i n and results in a safety factor of 2.36.This safety factor is greater than the required safety factor os 2 per IWB-3612(b) of Reference 6.Therefore, it is concluded that the postulated flaw size in the instrument nozzle of the St.Lucie pressurizer is acceptable for the thirty future years of plant life per IWB-3612 of the ASME Code Section XI.Prepared by: A.D.Nana Reviewed by: K.K.Yoon Date: June 1995 Date: June 1995 Page 27 of 29 B&W Nuclear Technologies I***BWNT NON-PROPRIETARY

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8.0 REFERENCES

1~'lorida Power&Light Drawing No.2998-19321, Rev.0,"Top Head Instrument Nozzles Repair".2.Florida Power&Light Drawing No.2998-18709, Rev.1,"Pressurizer General Arrangement".

3.BWNT Document 38-1210589-00,"Pressurizer Instrument Nozzles, FM Design Input," for St.Lucie Unit 2, dated 11/11/94 (FP&L Number JPN-PSLP-94-603, File: PSL-100-14).4.BWNT Document 38-1210588-00,"Pressurizer Instrument Nozzles, FM Design Input," for St.Lucie Unit 2, dated 11/11/94 (FP&L Number JPN-PSLP-94-631, File: PSL-100-14).5.'t.Lucie Unit 2 Updated Final Safety Analysis Report, through Amendment No.9, dated October 1994.6.ASME Boiler and Pressure Vessel Code,Section XI, 1989 Edition.7.BWNT Document 32-1235127-02,"Stresses for St.Lucie Unit 2, Pressurizer LEFM," by T.M.Wiger, dated June 1995.8.EPRI Report Number NP-719-SR,"Flaw Evaluation Procedures," with errata for subject report dated April 14, 1980, prepared by ASME Task Group on Flaw Evaluation, Electric Power Research Institute, Palo Alto California, August 1978.9."Solution of Three Dimensional Crack Problems using the Boundary Integral Equation Method," by J.Heliot, R.Labbens and A.Pellissier-Tanon, presented at the Second Prepared by: A.D.Nana Reviewed by: K.K.Yoon Date: June 1995 Date: June 1995 Page 28 of 29 B&W Nuclear Technologies

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• • • 32-1235128-02 International Conference on Numerical Methods in Fracture Mechanics, Swansea, Great Britain, July 1980.References marked with an"asterisk" are retrievable from the Utilities Record System.Authorized Project Manager's Signature Prepared by: A.D.Nana Reviewed by: K.K.Yoon Date: June 1995'ate: June 1995 Page 29 of 29 NUCLEAR ENGINEERING DEPARTMENT CoMPoNENT, SUPPoRT AND INsPEcTIoNs P.O.Box 14000 JUNo BEAcHg FLoRIDA 33408 St.Lucie Nuclear Power Plant Unit 2 ATTACHMENT B STRESSES FOR S7;LUCIE UNIT 2 PRESSURIZER LEFM Prepared by B&W NUCLEAR TECHNOLOGIES For St.Lucie Nuclear Power Plant 10 Miles South of Ff.Pierce on A1A Ft.Pierce, Florida 33034 NRC Docket Number: Document Number: Revision Number: 2 Date: Commercial Service Date: August 8, 1983 50-389 32-1235127-02 July 14, 1995