ML17228B048

Nonproprietary Version of Stresses for St Lucie Unit 2 Pressurizer Lefm.

ML17228B048
Person / Time
Site:	Saint Lucie
Issue date:	02/15/1995
From:	HARRISON H T, MILLER A M BABCOCK & WILCOX CO.
To:
Shared Package
ML17228B046	List: ML17228B045 ML17228B047 ML17228B048 ML17228B049
References
32-1235127-01, 32-1235127-01-R01, 32-1235127-1, 32-1235127-1-R1, NUDOCS 9503070361
Download: ML17228B048 (43)

Text

FLORIDA POWER AND LIGHT COMPANY NUCLEAR ENGINEERING DEPARTMENT P.O.Box 14000 Juno Beach, Florida 33408 St.Lucie Nuclear Power Plant Unit 2 ATTACHMENT 8 STRESSES FOR ST.LUG I E UNIT 2 PRESSURIZER LEFM Prepared by BBW NUCLEAR SERVICE COMPANY For St.Lucie Nuclear Power Plant 10 Miles South of Ft.Pierce on A1A Ft.Pierce, Florida 33034 Commercial Service Date: NRC Docket Number.Document Number.Revision Number.0 August 8, 1983 50-389 32-1235127-01 Date: February 15, 1995 FHQ: PSL1SNB.RPT Page 3 95030703&i 950302 PDR ADOCK 050003B9 P PDR

BWNT-20697-2(11/89)

JllBBMI NUCLFAH CM TECHNOI.OGIES CALCULATlONAL

SUMMARY

SHEET (CSS)DOCUMENT IDENTIFIER 32-1235127-01 TITLE Stresses for St.Lucie Unit 2, Pressurizer LEFM PREPARED BY'EVIEWED BY: A.M.MIller SIGNATUR TITLE Engr.IV COST 41020 REF.~AME H.T.Harriso SIGNATUR~~9'5 TITLE Princi al En r.38 TM STATEMENT'EVIEWER PURPOSE AND

SUMMARY

OF RESULTS The purpose of this document was to determine enveloping Normal and Upset, Condition stresses for the seven 1" instrumentation/temperature sensing nozzles in the pressurizer at St.Lucie Unit 2.Results from this document were used as inputs to the fracture mechanics evaluation, Reference[7].e stress results are summarized in Tables 6-1 through 6.8 in Section 6.0.Note that thermal stratification effects were not considered in this analysis.**BWNT NON-PROPRIETARY**

THE FOLLOWING COMPUTER CODES HAVE BEEN USED IN THIS DOCUMENT: CODE I VERSION I REV CODE I VERSION I REV THIS DOCUMENT CONTAINS ASSUMPTIONS THAT MUST BE VERIRED PRIOR TO USE ON SAFETY-RELATED WORK YES()No(X)FAG~OF BGW NUCLEAR TECHNOLOGIES

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32-1235127-01 RECORD OF REVISIONS Revision Number Pages Added/Chanched Descri tion All Original issue 01 All Issue of Non-Proprietary Version Prepared By: A.M.Miller Reviewed By: H.T.Harrison Date: 2 95 Date: 2 95 Page: 2 BRW NUCLEAR TECHNOLOGIES

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32-1235127-01

1.0 Introduction

4 2.0 Assumptions 4 3.0 Design Inputs 3.1 Design Characteristics 3.2 Material Properties 3.3 Model Geometry~~~~5~~0~6 8 4.0 Finite Element Model.......................10 5.0 Thermal Analysis...................~.....10 6.0 Stress Analysis..........................18 7.0 ANSYS 5.0A Verification

......................37 8.0 References 38 9.0 Microfiche

~~~~~~~~~~~~~~~39 Prepared By: A.M.Miller Reviewed By: H.T.Harrison Date: 2 95 Date: 2 95 B&W NUCLEAR TECHNOLOGIES

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32"1235127-01

1.0 Introduction

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.Subsecpxent NDE identified indications on the J-welds for all four steam space instrument nozzles.Modifications were made and justifications performed to determine the potential for crack growth during plant operation.

The evaluation performed at the time was conservatively limited to one cycle based on the design information available.

The purpose of this analysis is to provide stress analysis input for a bounding fracture mechanics flaw evaluation so that it is applicable to all seven instrument/temperature 1" nozzles in the pressurizer.

Results f rom this document were used as inputs to the fracture mechanics evaluation, Reference[7].The results were presented in the coordinate system of the postulated flaw as recpxired by Reference[7]since the fracture mechanics evaluation determined the postulated flaws, see Figure 6.2.The component stresses along a postulated flaw plane shown in Figure 6.2 were determined using ANSYS 5.0A finite element software.2.0 Assumptions 1)Material properties for SA-240, Type 304 were assumed for the stainless steel cladding on the pressurizer heads and shell.2)The effects of the nozzle were neglected in determining the shell/head stresses (i.e.the nozzle was omitted from the finite element model).3)The hT between the cladding surface temperature and the bulk fluid temperature was assumed to be a specific hT'F for calculating the natural convection heat transfer coefficient.

4)Piping loads on the instrumentation/temperature sensing nozzles produce negligible stresses on the pressurizer shell/head.

5)Effects of thermal stratification were not considered in this analysis.6)Hydrotest was assumed to be shop hydrotest only.Therefore, no future hydrotests are assumed to occur.Prepared By: A.M.Miller Reviewed By: H.T.Harrison Date:~2 95 Date: 2 95 B&W NUCLEAR TECHNOLOGIES

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32-1235127" 01 3.0 Design Inputs 3.1 Design Characteristics The following design parameters for the pressurizer were taken from Reference[8].Heatup Cooldown Operating Pressure Operating Temperature Minimum Pressure (Reactor Trip transient)

Maximum Pressure (Abnormal Loss of Load transient.)

100'F/hr 200'F/hr 2250 psia 653 F 1740 psia (653-616'F hT)2400 psia (664-614'F hT)Prepared By: A.M.Miller Reviewed By: H.T.Harrison Date: 2 95 Date:~2 99 BOW NUCLEAR TECHNOLOGIES

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32-1235127-01 3.2 Material Properties This section summarizes the material properties used in the thermal/stress analysis.The material types come from References

[8-10]and assumption 1.References for the material properties are given in the tables below.The material property designation and units are: KXX-thermal conductivity, btu/(hr-in-'F)

DENS-density, lb/in'-specific heat, btu/(lb-'F)

C is a calculated value based on C KXX/(DENS x Thermal Diffusivity) where thermal diffusivity is taken from the same source as KXX EX-Young's Modulus, psi x 10'LPX-coefficient of thermal expansion, in/in/'F x 10~Sm-design stress intensity, ksi Sy-yield strength, ksi Su-ultimate strength, ksi v-Poisson's ratio~.3 for all materials PRESSURIZER HEADS AND SHELL SA-533 GR-B CL-1, Low Alloy Steel (Mn-.SMo-.SNi)

TEMP 100 200 300 400 500 600 700 REF DENS.2839.2831.2823.2817.2809.2802.2794 1.8833 1.9500 1.9833 1.9833 1.9583 1.9167 1.8583.1079.1139.1196.1257.1323.1389.1448 EX 29.3 28.8 28.3 27.7 27.3 26.7 25.5 ALPX 7.06 7.25 7.43 7.58 7.70 7'3 7.94 Sm 26.7 26.7 26.7 26.7 26.7 26.7 26.7 Sy 50.0 47.5 46.1 45.1 44.5 43.8 43.1 Su 80.0 80.0 80.0 80.0 80.0 80.0 80.0 Prepared By: A.M.Miller Reviewed By: H.T.Harrison Date: 2 95 Date: 2 95 B&W NUCLEAR TECHNOLOGIES

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32-1235127-01 HEAD AND SHELL CLADDING 304 STAINLESS STEEL, SA-240 ASSUMED (18Cr-8Ni)

TEMP 100 200 DENS.2862.2853.7250.7750.1157.1209 EX 28.1 27.6 ALPX 8.55 8.79 20.0 Sy 30.0 20.0'25.0 Su 75.0 71.0 300 400 500 600 700 REF.2844.2836.2827.2818.2810.8167.8667.9083.9417.9833.1246.1286.1313.1334.1358 27.0 26.5 25.8 25.3 24.8 9.00 9.19 9.37 9.53 9.69 20.0 18.7 17.5 16.4 16.0 22.5 20.7 19.4 18.2 17.7 66.0 64.4 63.5 63.5 63.5 Prepared By: A.M.Miller Reviewed By: H.T.Harrison Date: 2 95 Date: 2 95 BGW NUCLEAR TECHNOLOGIES

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32-1235127-01 3.3 Model Geometry To bound all instrument/temperature sensing nozzle locations in the pressurizer heads and shell, the radius of the modeled nozzle penetration was determined so that the total pressure stress would ecpxal or exceed the total pressure stress present at all nozzle locations.

Penetrations in the spherical heads experience increased stress due to the hillside effect of the skewed penetration.

Penetrations in the cylindrical portion of the shell experience increased stress due to the larger hoop stress and the larger stress concentration effects due to the stress profile around the penetration.

The model inherently included stress to account for the hillside effect and the stress for the stress concentration effect at a hole in the cylindrical shell.Prepared By: A.M.Miller Reviewed By: H.T.Harrison Date: 2 95 Date: 2 95 BGW NUCLEAR TECHNOLOGIES

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32-1235127-01 IJJ CL 0-I-C)(U I CC CC I-D CC Figure 3.1, Finite Element Model Geometry and Materials (inches)Prepared By: A.M.Miller Reviewed By: H.T.Harrison Date: 2 95 Date: 2 95 B&W NUCLEAR TECHNOLOGIES

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32-1235127-01 4.0 Finite Element Model ANSYS 5.0A finite element software, Reference[6], was used to perform the axisymmetric thermal and stress analysis of the nozzle penetration.

A sufficient portion of the head/shell was modeled to attenuate the stress concentration effects at nozzle penetration.

The head/shell and cladding were modeled using axisymmetric elements (PLANE 42 structural and PLANE 55 thermal).5.0 Thermal Analysis The transients in Ref.[8]were reviewed for those likely to produce maximum tensile stress on the inside surface of the pressurizer (conservative for the fracture mechanics analysis in Reference[7]).Based on the review, the following transients were evaluated in this analysis: 100'F/hr Heatup, 200'F/hr Cooldown, a bounding Upset Condition transient which was represented as a 53'F Step-down (pressure 1740 psia)and a 53'F Step-up (pressure 2400 psia), and'oss of Secondary Pressure.The Heatup, Step-down, Step-up and Cooldown transients were combined into one computer run (microfiche THERMAL.OUT) with Heatup from 70'F to 653'F (0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> to 5.83 hours9.606481e-4 days <br />0.0231 hours <br />1.372354e-4 weeks <br />3.15815e-5 months <br />), 53'F Step-down (7 hr instantaneous), 53'F Step-up (8 hr instantaneous) and Cooldown from 653'F to 70'F (9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> to 11.915 hours0.0106 days <br />0.254 hours <br />0.00151 weeks <br />3.481575e-4 months <br />).The Loss of Secondary Pressure transient was run and the results are contained in microfiche LPTHERM.OUT.

Thermal conditions were imposed on the finite element model as applied heat transfer coefficients and bulk fluid temperatures.

Heat transfer was assumed to occur at only the inside surface of the pressurizer as shown in Figure 5.1.All other surfaces were assumed to be insulated.

A constant heat transfer coefficient was used to simplify the analysis in a conservative manner.Since overestimating the tensile stresses at the inside surface of the pressurizer was conservative for the fracture mechanics analysis in Reference[7], the heat transfer coefficient was selected to result in conservatively high tensile stresses on the inside surface of the pressurizer.

During Heatup and the Step-up transients, heating of the inside surface causes compression on the inside surface of the pressurizer.

Therefore, use of a low heat transfer coefficient results in conservative (tensile)stresses.Conversely, during Cooldown and the Step-down transient, a high heat transfer coefficient results in conservative (tensile)stresses.Therefore, nozzles in Prepared By: A.M.Miller Reviewed By: H.T.Harrison Date:~2 99 Date: 2 95 Page: 10 B6cW NUCLEAR TECHNOLOGIES

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32-1235127-01 the steam space were conservatively represented using heat transfer coefficients in the water space (i.e.for heating, condensing steam coefficients are greater than natural convection coefficients in the water space and for cooling, natural convection steam coefficients are less than natural convection coefficients in the water space).The heat transfer correlations in Ref.[2]were reviewed for horizontal and vertical plates.The correlation for a horizontal heating plate, face up (T>Tm)was selected as a representative heat transfer coefficient for the nozzles in the water space.The results of the thermal analyses were reviewed using the ANSYS POST26 post processor to determine times when the radial hT's occurred.The temperature at the inside surface of the pressurizer (essentially the bulk fluid temperature) and the radial hT are shown in Figures 5.2 through 5.5.The temperature distribution at times of extreme hT's were then used as inputs for the stress analyses since the extreme hT's resulted in extreme thermal stresses.Steady state temperature cases were also run at 653 F and 2250 psia without material discontinuity effects (T~T~microfiche STEADYST.OUT) and at 653'F and 2400 psia with material discontinuity effects (T , 70'F, microfiche STRESS.OUT)

.The following table summarizes the critical transient times and identifies the associated pressures.

The location of node pair used for evaluation of the hT's was the same node pair used for the stress path in Section 6.0 and is shown in Figure 6.2 and the POST26 results are contained at the end of the thermal runs Prepared By: A.M.Miller Reviewed By: H.T.Harrison Date: 2 95 Date: 2 95 Page: 11 BOW NUCLEAR TECHNOLOGIES

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32-1235127-01 in the microfiche in Section 9.0.Note that although the node pair used for evaluating the radial hT was at a specified flaw angle through the shell wall, it was representative of the radial gradient since the heat transfer is one dimensional in the radial direction (i.e., the entire inside surface is isothermal and the entire outside surface is isothermal).

TABLE 5.1, CRITICAL TRANSIENT TIMES TRANSIENT TRANSIENT TIME (HR)PRESSURE (PSIA)Heatup 5.83 2, 250 Step-down 7.0667 1, 740 Step-up 8.0667 2,400 Steady State'.000 2,400 Cooldown 9.5247 1, 028'ooldown 11.915 Steady State',250 Loss of Secondary Pressure.051958 200'The pressure was assumed to equal the saturation pressure at 548'F'Includes material discontinuity temperature effects (T,=70'F).'Excludes material discontinuity temperature effects (T ,=T~~653'F).Prepared By: A.M.Miller Reviewed By: H.T.Harrison Date:~295 Page: 12 gmmmlllll~~>

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32-1235127-01 z LU D O K Q nu?nu)0 v-~OOCiE lf Ill/)-U U.U Uj NOX>.NO o K H I-A1VA I I V)(9 1 Uj O 0 I-Figure 5.2, Shell Radial hT Time History, Heatup, 53hT Steps 6 Cooldown Prepared By: A.M.Miller Reviewed By: H.T.Harrison Date: 2 95 Date: 2 95 Page: 14 H 0 8 ANSYS 5.0 A OCT 25 1994 10:02:30 PLOT NO.2 POST26 ZV 1 DIST 0.75 XF 0.5 YF 0.5 ZF-0.5 CENTROID HIDDEN n i 5 0 0 0 I 0 M TEMP RAD 0 0 0 0 0 STLUCIE1.IGS 10 TIME 12 16 18 20 M I hl 4J Vl h)I O 6l 8l r.e 6l Ql 0 0 td W'C'C z V 0 Ul*0 0 n 0 0<C STLUCIE1.IGS Oo2 0.4 0.6 0.8 TIME i+2 l.6 RAD (>>loca-l)ANSYS 5.0 A OCT 25 1994 09:07:48 PLOT NO.POST26 ZV 1 DIST 0.75 XF 0.5 YF 0.5 ZF 05 CENTROID HIDDEN n n 0 El O g I 0 H W 4)M I h>OJ Ul W I O B&W NUCLEAR TECHNOLOGIES

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32-1235127-01 CII Q~g)O~gg<zol 0-Sl-op V)z LU D D X O~u)u)n<r OOOO>~i s a~+~4 I O X CO~4 bl K H 00 0 O O N O A 1VA (D (9 uj O D e~Figure 5.5, Inside Surface Temp and Radial hT, Loss of Secondary Pressure Prepared By: A.M.Miller Reviewed By: H.T.Harrison Date: 2 95 Page: 17 B6cW NUCLEAR TECHNOLOGIES

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32-1235127-01 6.0 Stress Analysis The temperature distributions and pressures at the critical times in Table 5.1 were imposed on the model to obtain the stresses in the pressurizer shell.The pressure boundary conditions are shown in Figure 6.1.Since the fracture mechanics evaluation in Reference[7]requires stresses along a specified flaw plane, the ANSYS POST1 post processor was used to transform the stresses into the flaw plane coordinate system as shown in Figure 6.2.Symmetry boundary conditions were used at the edge of the pressurizer head to restrict the heads motion to only the radial direction.

A nodal force was applied to the head to represent the end cap load developed at the nozzle (nozzle end cap load=mr'(pressure)

~w(.6875)'(pressure)

~1.4849(pressure)).

The nodal force was conservatively applied to the outside surface of the model since it would tend to increase the tensile stresses in the pressurizer shell which is the region of interest.Complete stress results are contained in the microfiche of Section 9.0.The stresses as required for the FM analysis are summarized in the Tables 6.1 through 6.8 and Figures 6.3 through 6.10.Note that the stresses are in psi.Prepared By: A.M.Miller Reviewed By: H.T.Harrison Date: 2 95 Date: 2 95 Page: 18 BOW NUCLEAR TECHNOLOGIES NON-PROPRIETARY**

32-1235127-01 TABLE 6~1 g END OF HEATUP g P~2250 PSI S 0.00000E+00 0.11959 0.23918 0.35876 0.47835 0.59794 0.71753 0.83711 0.95670 1.0763 1.1959 1.3155 1.4351 1.5546 1.6742 1.7938 1.9134 2.0330 2.1526 2.2722 2.3918 2.5113 2.6309 2.7505 2.8701 2.9897 3.1093 3.2289 3.3485 3.4680 3.5876 3.7072 3.8268 3.9464 4.0660 4.1856 4.3052 4.4247 4.5443 4.6639 4.7835 4.9031 5.0227 5.1423 5.2619 5.3814 5.5010 5.6206 5.7402 SX-7283.8-2390.6 2348.7 5917.8 7064.7 8210.2 9242.5 9868.3 10502.11084.11488.11893.12269.12553.12834.13098.13309.13515.13713.13877.14037.14192.14323.14453.14580.14686.14793.14898.14991.15083.15141.15163.15188.15246.15305.15354.15394.15433.15473.15507.15530.15553.15577.15601.15938.16287.16624.16947.17256.SY 5973.2 5126.0 4768.9 5548.9 6676.2 7778.0 8839.1 9535.7 10230.10889.11363.11838.12287.12640.12991.13324.13601.13873.14133.14355.14571.14780.14960.15134.15304.15450.15594.15732.15848.15965.16109.16263.16413.16513.16610.16701~16784.16865.16946.17022.17087.17152.17217.17281.16990.16681.16387.16106.15838.SZ 60276.55404.50677.46241.44495.42714.41030.40141.39242.38404.37926.37440.36991.36728.36456.36206.36064.35908.35768.35692.35604.35527.35487.35441.35400.35380.35356.35334.35323.35313~35295.35291.35286.35264.35241.35234.35214.35192.35170.35150.35120.35090.35060.35031.34986.34940.34894.34848.34802.Prepared By: A.M.Miller Reviewed By: H.T.Harrison Date: 2 95 Date:~2 95 Page: 19 B&W NUCLEAR TECHNOLOGIES NON-PROPRIETARY**

32-1235127-01, TABLE 6.2, 53F STEP DOWN, P~1740 PSI S 0.00000E+00 0.11959 0.23918 0.35876 0.47835 0.59794 0.71753 0.83711 0.95670 1.0763 1.1959 1.3155 1.4351 1.5546 1.6742 1.7938 1.9134 2.0330 2~1526 2.2722 2.3918 2.5113 2.6309 2.7505 2.8701 2.9897 3.1093 3.2289 3.3485 3.4680 3.5876 3.7072 3.8268 3.9464 4'660 4.1856 4.3052 4.4247 4.5443 4.6639 4.7835 4.9031 5.0227 5.1423 5.2619 5.3814 5.5010 5.6206 5.7402 SX-6913.0-1682.6 3307.9 6953.4 7957.1 8893.3 9672.7 10050.10402.10696.10819.10932.11024.11030.11038.11037.10989.10954.10912.10846.10794.10736.10672.10615.10555.10499.10444.10391.10346.10299.10238.10162.10088.10053.10020.9980.3 9950.7 9923.1 9895.9 9869.6 9852.2 9835.4 9819.1 9803.3 10004.10214.10416.10609.10794.SY 8348.2 6909.2 5990.6 6441.3 7527.8 8532.9 9420.0 9888.7 10328.10707.10889.11060.11203.11250.11298.11332.11313.11307.11290.11246.11214.11175.11126.11084.11038.10995.10953.10911.10872.10834.10823.10828.10831'0819.10807.10791.10789.10789.10790.10791.10803.10816.10828.10839.10650.10452.10263.10081.9908.0 SZ 70271.62772.55758.49572.46742.43956.41368.39639.37939.36375.35178.34011.32937.32044.31191.30394.29697.29052.28434.27886.27380.26888.26458.26052.25654.25317.24988.24670.24402.24135.23884.23665.23447.23257.23071.22896.22751.22611.22472.22345.22248.22152.22057.21962.21903.21846.21791.21736.21681.Prepared By: A.M.Miller Reviewed By: H.T.Harrison Date: 2 95 Date: 2 95 Page: 20 B&W NUCLEAR TECHNOLOGZES NON-PROPRZETARY**

32-1235127-01 TABLE 6.3, 53F STEP UP, P~2400 PSZ S 0.00000E+00 0.11959 0.23918 0.35876 0.47835 0.59794 0.71753 0.83711 0.95670 1.0763 1.1959 1.3155 1.4351 1.5546 1.6742 1.7938 1.9134 2.0330 2.1526 2.2722 2.3918 2.5113 2.6309 2.7505 2.8701 2.9897 3.1093 3.2289 3.3485 3.4680 3.5876 3.7072 3.8268 3.9464 4.0660 4.1856 4.3052 4.4247 4.5443 4.6639 4.7835 4.9031 5.0227 5.1423 5.2619 5.3814 5.5010 5.6206 5.7402 SX-7254.5-2602.9 1924.9 5376.5 6556.3 7759.1 8874.0 9593.9 10334.11027.11544.12066.12554.12950.13338.13705.14019.14316.14605.14855.15091.15324.15523.15716.15906.16065.16222.16377.16510.16643.16734.16781.16832.16912.16995.17065.17121.17175'7229.17276.17306.17337.17368.17400.17777.18168.18544.18905.19251.SY 5009.8 4374.6 4204.6 5068.2 6184.4 7297.4 8404'9179.8 9966.4 10731.11319.11914.12486.12961.13433.13885.14279.14658.15025.15349.15656.15955.16217.16466.16711.16919.17123'7319.17481.17644.17833.18028.18220.18347.18471.18589.18691.18790.18889.18981.19056.19132.19207.19282'8956.18611.18281.17966.17668.SZ 55366.51513.47681.43963.42672.41314.40030.39507.38957.38443.38279.38091.37914.37923.37897.37877.37966.38007.38058.38169.38238.38318.38420.38500.38585.38668.38744.38815.38881.38946.38994.39048.39101.39119.39134.39167.39173~39175.39176.39177.39158.39139.39119.39099.39055.39009.38962.38915.38867.Prepared By: A.M.Miller Reviewed By: H.T.Harrison Date: 2 95 Date: 2 95 Page: 21 BOW NUCLEAR TECHNOLOGIES

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32-1235127-01 TABLE 6.4, STEADY STATE, Tref 70F, Tunif 653F, P=2400 PSIA S 0.00000E+00 0.11959 0.23918 0.35876 0.47835 0.59794 0~71753 0.83711 0.95670 1.0763 1.1959 1.3155 1.4351 1.5546 1.6742 1.7938 l.9134 2.0330 2.1526 2.2722 2.3918 2.5113 2.6309 2.7505 2.8701 2.9897 3.1093 3.2289 3.3485 3.4680 3.5876 3.7072 3.8268 3.9464 4.0660 4.1856 4.3052 4.4247 4.5443 4.6639 4.7835 4~9031'.0227 5.1423 5.2619 5.3814 5.5010 5.6206 5.7402 SX-8203.5-2489.2 3016.7 7124.0 8389.6 9630.6 10730.11367.12002.12576.12950.13321.13659.13895.14127.14342.14499.14653.14797.14906.15014.15116.15196.15275.15351.15411.15471.15531.15582.15632.15649.15633.15620.15646.15675.15693.15708.15724.15739.15751.15759.15767~15776.15785.16120.16469.16804.17126.17434.SY 7687.1 6501.0 5884.8 6653.7 7928.7 9155.6 10311.11033.11745.12410.12859.13306.13723.14028.14332.14617.14837.15055'5260.15424.15585.15737.15861.15983.16099.16195.16289.16379.16451.16524, 16627.16743.16857.16924.16989.17048.17106.17164.17221.17276.17327.17377.17428'7478.17179.16864.16563.16275.16001.SZ 72516.66004.59778.54082.51714.49332.47103.45809.44517.43322.42543.41766.41049.40544.40042.39574.39228.38882.38558.38308.38058.37821.37634.37446.37266.37121.36975.36833.36717.36600.36483.36387.36292.36193.36094.36012.35931.35851.35772.35699.35631.35565.35498.35432.35372.35312.35253.35193.35134.Prepared By: A.M.Miller Reviewed By: H.T.Harrison Date: 2 95 Date: 2 95 Page: 22 B&W NUCLEAR TECHNOLOGIES NON-PROPRIETARY**

32-1235127-01 TABLE 6.5, COOLDOWN, Psat=1028 PSIA S 0.00000E+00 0.11959 0.23918 0.35876 0.47835 0.59794 0.71753 0.83711 0.95670 1.0763 1.1959 1.3155 1.4351 1.5546 1.6742 1.7938 1~9134 2.0330 2.1526 2.2722 2.3918 2.5113 2.6309 2.7505 2.8701 2.9897 3.1093 3.2289 3.3485 3.4680 3.5876 3.7072 3.8268 3.9464 4.0660 4.1856 4.3052 4.4247 4.5443 4'639 4.7835 4.9031 5.0227 5.1423 5.2619 5.3814 5.5010 5.6206 5.7402 SX-4251.9-949.12 2183.5 4457.5 5075.9 5642.0 6110.7 6333.8 6537.5 6703.0 6763'6815.3 6851.0 6834.5 6816.9 6791.2 6737.1 6688.0 6633.1 6563.8 6500.6 6433.4 6361.2 6293.0 6222.3 6153.7 6085.9 6019.0 5957.7 5895.3 5825.5 5749.1 5673.4 5623.6 5575.3 5523.1 5481.3 5441.5 5402.0 5364.6 5338.6 5313.0 5287.8 5263.0 5363.8 5470.8 5573.3 5671.3 5764.8 SY 5658.2 4628.9 3925.5 4126.6 4801.8 5420.1 5958.3 6232.9 6488.4 6707.9 6802.5 6889.7 6958.5 6965.9 6972.4 6968.3 6931.0 6898.1 6857.6 6799.7 6746.8 6688.8 6624.2 6562.5 6497.9 6434.8 6372.3 6309.8 6250.4 6190.6 6147.2 6114.0 6079.5 6040.6 6001.6 5960.2 5932.0 5906.1 5880.0 5856.3 5846.3 5835.9 5825.3 5814.4 5707.6 5596.8 5490.3 5388.1 5290.2 SZ 45637.40460.35641.31472.29573.27713.26003.24848.23716.22680.21877.21094.20370.19760.19173.18620.18130.17668.17222.16818.16438.16067.15731.15411.15095.14816.14542, 14276.14041.13807.13586.13386.13186.13011.12839.12674.12536.12403.12271.12149.12059.11969.11879.11791.11743.11699.11655.11612.11569.Prepared Reviewed By: A.M.Miller By: H.T.Harrison Date: 2 95 Date: 2 95 Page: 23 B&W NUCLEAR TECHNOLOGIES

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32-1235127-01 TABLE 6.6, END OF COOLDOWN, P~O PSI S 0.00000E+00 0.11959 0.23918 0.35876 0.47835 0.59794 0.71753 0.83711 0.95670 1.0763 1.1959 1.3155 1.4351 1.5546 1.6742 1.7938 1.9134 2.0330 2.1526 2.2722 2.3918 2.5113 2.6309 2.7505 2.8701 2.9897 3.1093 3.2289 3.3485 3.4680 3.5876 3.7072 3.8268 3.9464 4.0660 4.1856 4.3052 4.4247 4.5443 4.6639 4.7835 4.9031 5.0227 5.1423 5.2619 5.3814 5.5010 5.6206 5.7402 SX-519.07 129.63 701.44 1078.8 1148.3 1181.9 1182.2 1146.2 1094'1031.4 954.61 871.24 785.65 694.22 602.49 510.30 415.58 324.91 233.00 141.57 54.659-33.564-117.81-199.13-281.30-356.35-430.77-504.36-570.83-637.84-699.27-754.44-810.04-858.93-907.49-955.43-994.11-1031.2-1068.3-1102.5-1125.9-1149.3-1172.6-1195.7-1229.8-1263.7"1296.6-1328.4-1359.1 SY 2077.7 1580.4 1144.8 988.46 1099.8 1183.7 1219.9 1190.4 1150.6 1098.6 1021.0 938.62 852'8 754.86 657.42 558.39 454.90 355.00 253.68 152.20 55.383-42.486-136.32-226.71-317.58-400.69-482.67-563.15-635.20-707'0-777.36-843.39-909.32-963.51-1016.9-1070.3-1112.6-1153.1-1193.7-1231 03-1256.9-1282.6-1308.5-1334.6-1317.7-1298.4-1280.5-1263.8-1248.4 SZ 12153.10004.8066.5 6563.8 5865.2 5204.2 4606.3 4128.0 3668.5 3253.8 2882.4 2526.7 2198.2 1888.0 1595.4 1317.9 1048.9 800.61 558'6 323'1 106.80-107'4-310.33-501.66-691.53"864.00-1032.2-1196.1-1343.4-1490.0-1626.1-1752.9-1879.0-1984.9-2088.8-2193.2-2276.2-2355.1-2433.4-2506.0-2555.8-2605.1-2653'-2702.2-2720.1-2735.3-2750.1-2764.5-2778.6 Prepared By: A.M.Miller Reviewed By: H.T.Harrison Date: 2 95 Page: 24 B&W NUCLEAR TECHNOLOGZES

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32-1235127-01 TABLE 6.7, STEADY STATE Tunif Tref 653, P=2250 PSI S 0.00000E+00 0.11959 0.23918 0.35876 0.47835 0.59794 0.71753 0.83711 0.95670 1.0763 1.1959 1.3155 1.4351 1.5546 1.6742 1.7938 l.9134 2.0330 2.1526 2.2722 2.3918 2.5113 2.6309 2.7505 2.8701 2.9897 3.1093 3.2289 3.3485 3.4680 3.5876 3.7072 3.8268 3.9464 4.0660 4.1856 4.3052 4.4247 4.5443 4.6639 4.7835 4.9031 5.0227 5.1423 5.2619 5.3814 5.5010 5.6206 5.7402 SX-7698.3-2331.2 2839.0 6694.9 7882.4 9046.2 10077.10674.11270.11807.12157.12503.12820.13039.13255.13455.13600.13743.13877.13977.14077.14172.14245.14318.14388.14443.14498.14552.14599.14645.14660.14644.14631.14654.14681.14697.14710.14724.14738.14749.14755.14762.14770.14778.15092.15418'5732'6033.16321.SY 7242'6121.4 5535.2 6252.4 7449.3 8600.7 9684.9 10362.11029.11652.12072.12490.12880.13164.13448.13713.13918.14121.14311.14463.14612.14753'4868.14981.15088.15177.15264.15347.15413.15480.15576.15683.15789.15851.15911.15965.16019.16072..16125.16176.16223.16270.16317.16364.16084.15788.15506.15237.14980.SZ 68183.62041.56169.50803.48572.46328.44230.43011.41793.40667.39932.39199.38521.38043.37567.37125.36795.36467.36159.35920.35683.35457.35278.35099.34927.34788.34649.34513.34401.34289.34177.34085.33993.33898.33804.33725.33648.33572.33496.33426.33362.33299.33235.33173.33116.33060.33004.32948.32892.Prepared Reviewed By: A.M.Miller By: H.T.Harrison Date: 2 95 Date: 2 95 Page: 25 B&W NUCLEAR TECHNOLOGIES

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32-1235127-01 TABLE 6.8, LOSS OF SECONDARY PRESSURE, P=200 PSI S 0.00000E+00 0.11959 0.23918 0.35876 0.47835 0.59794 0.71753 0.83711 0.95670 1.0763 1.1959 1.3155 1.4351 1.5546 1.6742 1.7938 1.9134 2.0330 2.1526 2'722 2.3918 2.5113 2.6309 2.7505 2.8701 2.9897 3.1093 3.2289 3.3485 3.4680 3.5876 3.7072 3.8268 3.9464 4.0660 4.1856 4.3052 4.4247 4.5443 4.6639 4.7835 4.9031 5.0227 5.1423 5.2619 5.3814 5.5010 5.6206 5.7402 SX-5205.0 587.39 5836.8 9361.3 9958.3 10290.10363.10092.9703.6 9218.7 8598.4 7929.1 7239.1 6491.2 5747.0 5000.9 4228.6 3502.8 2767.6 2036.5 1355.7 663.80 8.7152-616.39-1248.7-1812.4-2368.7-2915.8-3393.9-3876.5-4314.1-4701.6-5092.1-5422.5-5749.9-6073.0-6325.6-6566.3-6806.8-7026.7-7172.2-7316.9-7460.9-7604.2-7816.6-8028.3-8233.2-8431.3-8622:6 SY 15781.12266.9310.9 8486.6 9430.2 10146.10524.10362.10107.9732'9133.3 8489.4 7802.1 7019.9 6235.6 5437.8 4599.9 3801.6 2992.4 2182.8 1422.8 654.06-76.898-773.97-1474.9-2101.0-2715.4-3315~1-3835.6-4359.0-4853.0-5312.7-5771.4-6135.3-6492.4-6849.9-7122.2'-7380.6-7639.3-7877.5-8031.6-8186.8-8343'-8500.8-8385.2-8254.9-8132.8-8019.2-7913.9 SZ 98478.82490.68060.56501.50752.45286.40344.36430.32662.29217.26146.23197.20462.17888.15461.13164.10941.8912.1 6934.6 5031.0 3296.1 1579.5-22.322-1522.1-3010.1-4329.8-5611.0-6851.7"7933.1-9009.0-9994.7-10897.-11795.-12524.-13236.-13951~-14503.-15023.-15539.-16014.-16330.-16643."16952.-17257."17369.-17463.-17554.-17643.-17729.Prepared By: A.M.Miller Reviewed By: H.T.Harrison Date: 2 95 Date: 2 95 Page: 26 llmml5lmlllll illlmmllm~<

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• • BWNT NON"PROPRIETARY**

32"1235127-01"Intensionally Left Blank-Contains BWNT Proprietary Information" Figure 6.2, Stress Path for Flaw Plane Prepared By: A.M.Miller Reviewed By: H.T.Harrison Date:~295 Date: 2 95 Page: 28 B&W NUCLEAR TECHNOLOGIES

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32-1235127-01 m-e I--og gSgZ>r~Q-I I cnl-cu0cnujm 0-j'I-I-Q~~Z~g ZOe~OI-DR<00 40oQ.Q.cONI-Q.ZZ Z O 0 K O CS~~Q~ann/OOOO s~)sr~+QBU.lLILLU 00 O a 0)H ec O 40 W O X 0 CO P Figure 6.3, End of Heatup Stresses (psi)Prepared By: A.M.Miller Reviewed By: H.T.Harrison Date: 2 95 Date: 2 95 Page: 29 BOW NUCLEAR TECHNOLOGIES

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32-1235127-01 v)l-ol0v)ulcl+l-ZO(yj~OI-Z

<00 40oCLO.SNl-O.ZZ Z ul O O K O CS~~Q+u)u)u)0 v-g OO(5P Il I I IZ)Bu.u.u.uj NOXO-NO O I-N g)g H ec o g c O CQ G.I O CL ul 8 Figure 6', 53'F Step-down Stresses (psi)Prepared By: A.M.Miller Reviewed By: H.T.Harrison Date: 2 5 Date: 2 95 Page: 30 BOW NUCLEAR TECHNOLOGIES

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32-1235127-01 CO geP)Z<cOQ.g g ZOg)9QI-D><QQ COSQ.Q.V)NI-Q.ZZ O O x 0 CS~~.Q v OOOO spssag O a U)a ei O'V cv Pi 5 5 cv 00 0'igure 6.5, 53'F Step-up Stresses (psi)Prepared By: A.M.Miller Reviewed By: H.T.Harrison Date: 2 95 Date: 2 95 Page: 31 B&W NUCLEAR TECHNOLOGIES

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32-1235127-01 ZO"90I->K<00 c(0v-Q.o.cOU)I-0 zz tll CI Q K O O+eee0 r I OOO<If I I IP N M CO n 40 OI H ei O 40 W O X n n Figure 6.6, Steady State 653'F (Tref 70'F)Stresses (psi)Prepared By: A.M.Miller Reviewed By: H.T.Harrison Date: 2 95 Page: 32 BEcW NUCLEAR TECHNOLOGIES

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32-1235127-01

~X~f-%O~<~~i~CfCf ZO"BOf-OR<00

<0 a.a.rnrnf=a.aa O O O CI~~.Q'nnng v-g OOCjP If I I lg M V)N O a N H O h 4 Pt A 5 O CO Q.CO Ol C)E5 0.z 0 9 0 0 O Figure 6.7, Cooldown Stresses (psi)Prepared By: A.M.Miller Reviewed By: H.T.Harrison Date: 2 95 Date: 2 95 Page: 33 BOW NUCLEAR TECHNOLOGIES

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32-1235127-01 g~PAZ~~t-~>+I I CZP-Q.Q.V) gP0.ZZ z O O K O nu?u?u?OK~I OOO)~sang)-U.U.U.Ul NOXO-NO 0-N (OM CO O~0 M H C4 4O o o X cv O V)CD I Z 0 9 0 O 0 O e Figure 6.8, End of Cooldown Stresses (psi)Prepared By: A.M.Miller Reviewed By: H.T.Harrison Date: 2 95 Date: 2 95 Page: 34 BOW NUCLEAR TECHNOLOGIES

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32-1235127-01 Z0~90gg><OQ COSQ.O.V)Nl-fLZZ Z Lll O O K O O~~.g 5 g OOQa I~I I~P N M a O 40 M H C5 O o 8 r g g e 959 Figure 6.9, Steady State 653'F (Tref=Tunif)Stresses (psi)Prepared By: A.M.Miller Reviewed By: H.T.Harrison Date:~2 95 Date: 2 95 Page: 35 BOW NUCLEAR TECHNOLOGIES

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32-1235127-01

<X" Rogg90PD><00

<0 O.O.COCOI-O.aa Z LU Cl O K O~IAIALOQ~g OOO)~a I Sg N 0-r CO a O H cv O CO CL 8 I CL UJ K D CO CO LCj K C 0 R 0 O Llj CO 0 CO CO 0 Figure 6.10, Loss of Secondary Pressure Stresses (psi)Prepared By: A.M.Miller Reviewed By: H.T.Harrison Date: 2 95 Date: 2 95 Page: 36 BEcW NUCLEAR TECHNOLOGIES

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32-1235127-01 7.0 ANSYS 5.0A Verification The ANSYS analysis code, version 5.0A, was verified using closed form solutions for hoop stress in a sphere and stress concentration factors.The following comparison of finite element (FE)stress results and closed form solutions indicated that the software provided accurate results.Therefore, ANSYS BOA was verified for this application.

Finite Element Results from Table 6.7, Pressure 2250 psia Distance Hoo Stress Sz 0.0 4.3052 68183 33648 Hoop Stress Ref.[5, Table 28 case 3a.]e~pr/2t e=2250 (120.5)/[2 (4.094)]~33, 112 psi Compared to 33,648 psi (FE), h%=(33648/33112

-1)(100%)1.6%Stress Concentration at Hole SCF=a,,/a,,>>68183/33648 2.026 (FE)Compared to 2 (from Section A3.3), h%(2.026/2-1)(100%)1.3%Prepared By: A.M.Miller Reviewed By: H.T.Harrison Date: 2 95 Page: 37 B&W NUCLEAR TECHNOLOGIES

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32-1235127-01

8.0 References

1)ASME Boiler and Pressure Vessel Code,Section III, Division 1, Appendices, 1986 Edition with no Addenda.2)BWNS Document No.51-1155656-00,"Standard Correlations for Natural Convection"."BWNT Proprietary Document." 3)BWNS Document No.NPGD-TM-500,"NPGD Material Properties Program User's Manual", Rev.D, March, 1985."BWNT Proprietary Document." 4)Not used.5)Young, W.C.,"Roark's Formulas for Stress&Strain", 6th Edition McGraw-Hill, New York, 1989.6)DeSalvo, G.J.and Gorman, R.W.,"ANSYS User's Manual for Revision 5.0", 1992, Swanson Analysis Systems, Houston, Pennsylvania.

7)BWNT Document 32-1235128-00,"FM Analysis of St.Lucie Pressurizer Instrument Nozzles"."BWNT Proprietary Document." 8)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).

9)'Florida Power&Light Drawing No.2998-19321, Rev.0,"Top Head Instrument Nozzles Repair".10)'Florida Power&Light Drawing No.2998-18709, Rev.1,"Pressurizer General Arrangement".

• References marked with an"asterisk" are retrievable from the Utilities Record System.uthorized Pro'ect Manager's Signature Prepared By: A.M.Miller Reviewed By: H.T.Harrison Date: 2 95 Date: 2 95 Page: 38 BOW NUCLEAR TECHNOLOGIES

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32-1235127-01

9.0 Microfiche

Microfiche are not included in this document because they contain math model which is BWNT Proprietary.

However, for completeness the listing of the computer runs as contained in Rev.0 of the document ("BWNT Proprietary")are given below.FILE NAME THERMAL.OUT STRESS.OUT LPTHERM.OUT LPSTRESS.OUT STEADYST.OUT ENDCD.OUT DESCRIPTION Composite transient thermal runs including 100'F/hr Heatup, 53'F Step-down, 53'F Step-up and 200'F/hr Cooldown Stress runs for the composite transient in"THERMAL.OUT" Loss of pressure thermal run Stress run for the Loss of Pressure transient Steady state stresses (Tref Tunif 653'F)Stress run for the end of the 200'F/hr Cooldown.Prepared By: A.M.Miller Reviewed By: H.T.Harrison Date: 2 95 Date: 2 95 Page: 39

', lt 8 A 0 0