ML20196C128

Reactor Vessel Heatup & Cooldown Limit Curves for Normal Operation

ML20196C128
Person / Time
Site:	North Anna
Issue date:	05/31/1988
From:	Palusamy S, Schmertz J, Yanichko S WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20196B903	List: ML20196B900 ML20196B918 ML20196C128
References
WCAP-11791, NUDOCS 8812070182
Download: ML20196C128 (36)

Text

g WeSTINO440088 CLASS 3 CUSTOssem DESIGNATED DISTRieUTION WCAP-11791 ANALYSIS OF CAPSULE U FRON THE VIRGINIA POWER CONPANY NORTH ANNA UNIT 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAN NORTH ANNA UNIT 1 REACTOR VESSEL HEATUP AND C00LDOWN LIMIT CURVES FOR NORMAL OPERATION May 1988 Prepared by: MCfM y C. Schmertz /

Verified by: N[ 4 40

5. e.gantenxo Approvedby:/ # # _f f 5. 5. falusamy, Manager Structural Materials Engineering Work Performed Under Shop Order VHHJ-106 '

Prepared by Westinghouse for the Virginia Power Company Although information contained in this report is nonpreorietary, no distribution shall be made outside Westinghouse er its licensees without the customer's approval.

WESTINGHOUSE ELECTRIC CORPORATION Generatieri Technology Systems Divisien P.O. Box 2728 Pittsburgh, Pennsylvania 15230-2728

"***T#t ro70182 881130 8 FDR ADOCK 0500

TABLE OF CONTENTS Section Title Page

1.0 INTRODUCTION

1 2.0 FRACTURE TOUGHNESS PROPERTIES. 2 3.0 CRITERIA FOR ALLOWA8LE PRESSURE-TEMPERATURE RELATIONSHIPS 3 J

4.0 HEATUP AND C00LDOWN LIMIT CURVES 6 i

5.0 ADJUSTED REFERENCE TEMPERATURE 8

, REFERENCES 21 APPENDIX A - HEATUP C00LDOWN DATA POINTS A-1 I

1 j

s acess-esMes10 jj

L I

LIST OF TABLES Table Title Page ,

I Chemistry Factor For Welds, 'F ,

14 i

!! Chemistry Factor for Base Metal. 'T 16 l

!!! Reactor Vessel Toughnen: Data (Unirradiated) 18  ;

i IV Fast Neutron (E > 1.0 MeV) Exposure at the Pressure 19 Vessel Inner Radius - 0' Azimuthal Angle '

V Calculations of Fluence Factors Based en Surveillance Data 20 i

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LIST OF FIGURES I

Figure Title Page f 1 Fluence factor for Use in the Expression for ART 11

, NDT 2 North Anna Unit 1 Reacter Coolant Systen Heatup Limitations 12 Applicable for the First 10 EFPY I c

i 3 North Anna Unit 1 Reacter Coolant Systea Cooldown Limitations 13 l l

Applicable for the first 10 EFPY l

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t NORTH ANNA UNIT 1 HEATUP AND C00LDOWN LIMIT CURVES FOR NDRMAL CPERATION l

1.0 INTRODUCTION

Heatup and ecoldown limit curves are calculated using the most limiting value of RTNDT (reference nil-ductility temperature) for the reactor vessel. The most limiting RTNDT f the material in the core region of the reacter vessel is determined by using the preservice reacter vessel material fracture tough-ness preperties and estimating the radiation-induced ART RT is NDT. NDT casignated as the higher of either the dren weight nil-ductility transition

emperature (NDTT) er the temperature at which the material exhibits at least 50 f t-lb of impact energy and 35 mil lateral expansion (normal to the major working direction) minus 60*F.

RT NDT increases as the material is exposed to fast-neutron radiation.

Therefore, to find the most limiting RTNDT at any time period in the reacter's life, ART NDT due to the radiation exposure associated with that  !

ti:na period must be added to the original unirradiated RT The extent of '

NDT.

the shift in RT NDT is ennanced by certain chemical elements (such as cepper, nickel and phospherus) present in reacter vessel steels. Westinghouse, other NSSS vencers, the U.S. Nuclear Pegulatory Ccmmission and others have developed trend curves for predicting adjustment of RTNDT as a function of fluence and cepper, nickel and/or phosphorus centent. The Nuclear Regulatory Ccmmission I (NRC) trend curve is puolished in Regulatory Guide 1.99 (Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials)III.

Regulatory Guide 1.99 was originally published in July 1975 with a Revision 1 being issued in April 1E77. Currently, a Revision 2(2) to Regulatory Guide ,

1.99 is under censideration within the NRC. The chemistry facter, "CF" (*F),  !

a function of cepper and nickel centent identified in Regulatory Guide 1.99, Revision 2 is given in table I for welds and table !! for base retals (plates and fergings). Interpolation is permitted. The value. 'f', given in figure 1 is the calculated value of the neutren fluence at the location of interest (inner surface,1/4T, er 3/4T) in the vessel at the location of the postulated defect, n/cm2 (E > 1 MeV) divided by 1019 The fluence factor is determined frem figure 1.  !

nu.mme i

m '

. , i i

Giv:n the copper and nickel contents of the most limiting material, the  ;

radiation-induced ARTNOT can be estimated from tables I and !! and figure 1. j When two or more credible survei,llance data sets are available, they may be

{

used to predict the radiation induced ARINDI without using Tables I and !!,

in accordance with another procedure of Regulatory Guide 1.99 Rev. 2. In this other procedure, the chemistry factor is calculated by multiplying each measured ART NOTf or each capsule by its corresponding fluence factor, sussiing the products, and dividing by the sum of the squares of the fluence f actors. If this latter procedure gives the higher shift in ARTNOT, its use is meadator.. If it gives the lower shift, either procedure may be used.

For North Anna Unit 1 the surveillance data method of Regulatory Guide 1.99 Rev. 2 is t sed to predict ART NOT*

2.0 FRACTURE TOUGHNESS PROPERTIES The preirradiation fracture-toughness properties of the North Anna Unit i reactor vessel materials are presented in table !!!. The fracture-toughness properties of the ferritic material in the reactor coolant pressure boundary are determined in accordance with the NRC Regulatory Standard Raview PlanW The postirradiation fracture-toughness properties of the reactor vessel beltline material were obtained directly from the North Anna Unit i Vessel Material Surveillance Program. Credible surveillance data are currently available for two capsules for North Anna Unit 1. These are Capsule U (which is the more recent of the two) and Capsule V.

The shift in RTNOT and the adjusted reference temperature (ART) were calculated using the two methods described in the introduction. The surveillance data method was used to predict the ART for generating the heatup and cooldown curves.

NOT

-" 2

3.0 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS The ASME approach for calculating the allowable limit curves for various heatuo and cooldown rates specifies that the total stress intensity facto.r.

Kg , for the ccmbined tha: mal and pressure stresses at any time during heatup or ecoldown cannot be greater than the reference stress intsnsity facter, Kgg, fer the metal temperaturo at that time. 'X IR is obtained frem the reference fracture toughness curve, d2 fined in Appendix G to the ASME CodeI43 The KIR curve is given by the following equation:

K;g = 26.78 + 1.223 exp (0.0145 (T-RTNDT + 160)] (1) anere K

gg a

rsfarence stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temcarature RT NDT Therefore, the governing equation ier the heatuo-ecoldewn analysis is defined in Appendix G of the ASME CodeI43 as fd lows-CKgg + KIT 1KIR (21 dare K;g a

stress intensity factor caused by membrane (pressure) stress KIT = stress intensity factoi caused by the thermal gradients K;g = func*.ien of tamperature relative to the RTNDT of the material C = 0.0 for Level A and Level B service limits C = 1.~ for hydrostatic and leak test conditions during which the reactor core is cet critical an.wwe 3

l At any time during the heatup or cooldown transient, KIR is determined by tho metal temperature at the tip of the postulated flaw, the appropriate value for RTNOT, and the reference fracture toughness curve. The thermal stresses resulting from the temperature gradients through the vessel wall are calculatedandthenthecorresponding(thermal)stressintensityfactors, K1f, for the reference flaw are computed. From equation 2, the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated.

For the calculation of the allowable pressure versus coolant temperature dr,1nc cooldown, the reference flaw of Appendix G to the ASME Code I43 is assumet. to estat at the inside of the vessel wall. During cooldown, the centrolling location of the flaw is always at the inside of the wall because the thermal gradients produce tensile stresses at the inside, which increase l with increasing cooldown rates. Allowable pressure-temperature relations are l generated for both steady-state and finite cooldown rate situations. Frem these relations, composite limit curves are constructed for each cooldown rate of interest.

Tho use of the compcsite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on the measurement of reacter coolant tosperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw.

I' During cooldown, the 1/4 T vessel location is at a highe temperature than the fluid adjacent to the vessel 10. This condition, of course, is not true,for the steady-state situation. It follows that, at any given reactor coolant temperature, the AT developed during cooldomn results in a higher value of Kgg at the 1/4 T locatica for finite cooldown rates than for steady-state operation. Furthermore, if conditions e.tist so that ths increase in Kgg exceeds Kli, the calculated allowable pressure during cooldowr. will be greater than the steady-state value.

Tho above procedures are needed because there is no direct control on tesperature at the 1/4 T location and, therefore, allowable p'ressures may meereu.ece 4

unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this preblem and ensures conservative cperation of the system for the entire ecoldown period.

Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditiens assuming the presence of a 1/4 T defect at the insido of the wall that alleviate the tensile stresses produced by internal

ressure. The m, .1 temoerature at the crack tip lags the coolant temoerature; therefore, the K 7g fer the 1/4 T crack during heatup is icwor -

nan the K;g for the 1/4 i crack during steady-stste cencitiens at the same

colant temperature. During heatuo, esoecially i. the end of the transient, conditiens may esist so that the effects of compressive thermal stresses and lower X;g's do not offset each other, and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves fer finite heatup ratas when the 1/4 i flaw is considered.

Therefore, both cases have to be analyzed in order to ensure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained.

The seceni portien of the heatuo analysis c:ncerns the calculatien of the i pressure-temperature limitations for the case in which a 1/4 7 deep outside surface flaw is assumed. Unlike the situation at the vessel inside surface, the thereal gradients established at the outside surface during heatup produce stresses which are tensile in nature and therefore tend to reinforce any pressure stresses present. These thermal stresses are dependent on both the rate of heatup and the tiet (or coolant ttverature) along the heatuo ramp.

Since the thermal stresses at the outside are tensile and increase with increasing heatuo rates, each heatuo rate must be analyzed en an individual basis.

I Following the generation of pressure-temoerature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced a w.-c " '

5

by constructing a composite curve based on a point-by point comparison of the stcady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the least of the three values taken from the curves uncer consideration. The use of the composite curve is necessary to set conservative heatuo limitations because it is possible for conditions to exist wherein, over the course of the heatup ramp, the controlling condition suitches frem the inside to the outside, and the pressure limit must at all times be based on analysis of the most critical criterion.

Finally, the 1983 Amenement to 10CFR50(5] has a rule which addresses the metal tauperature of the closure head flange and vessel flange regions. This rule states that the metal temoerature of the closure flange regions must exceed the material RT NDT by at least 120*F for normal coeratien wnen the pressure exceeds 20 percent of the preservice hydrostatic test pressure (621 psig for North Anna Unit 1). From Section !!! of the ASME Code, the preservice hydrostatic test pressure is 1.25 times the design pressure, or 3105 psig. Table 111 indicates that the limiting RTNOT of -22'F or less occurs in the flange of North Anna Unit 1, so the minimum allowable temperature of this region is conservatively taken as 98'F at pressures greater than 621 psig. These limits are less restrictive than the limits shewn on figures 2 and 3.

4.0 HEATUP AND COOLDOWN LIMIT CURVES Licit curves for normal heatup and cooldown of the primary Reactor Coolant System have been calculated using the methods discussed in section 3, and the procedure is presented in reference 6.

Trcnsition temperature shif ts occurring in the pressure vessel materials due to radiation espesure have been obtained directly frem the reacter pressure vessel surveillance pregram.

Allowable corbinations of temperature and pressure fer specific temcerature change rates are below and to the right of the limit lines shewn in figures 2 and 3. This is in addition to other criteria which must be met before the reacter is made critical, nu.wwe 6

The reactor must not be made critical until pressure-temperature combinations are to the right of the criticality limit line shown in figure 2. The straight line portion of the criticality limit is at the minimum permissible temocrature for the 2485 psig inservice hydrostatic test as required by Appenoix G to 10CFR part 50. (5) The governing equation for the hydrostatic test is defined in Appendix G to Soution !!!.of the ASME Code I43 as follows:

1.5 X;g < KIR Where. X;g is the stress intensity factor covered by membrane (pressure) stress Kgg = 25.78 + 1.223 exo (0.0145 (T - RTNOT + 160))

T is the minimum permissible metal temperature and RT is the metal NDT reference nil-ductility temperature The curved portion of the criticality limit is shif ted 40'T to the right of and parallel to the hea'up t curve as required by Appendix G to 10 CFR50(5) ,

It should be noted that there are other criteria which must be met before the reacter 'is made critical. For examole, the reactor must not be made critical until a steam bubble is formed in the pressuri:er. The leak test linit curve shown en the heatum curve in figure 2 represents minimum temerature requirements at leak test pressures ranging from 2000 psig to 2485 psig. The leak test limit curve was determined by the same method used to ecmute the l inservice hydrostatic test temperature. Thia cathed used a 1.5 safety factor en the pressure stress intensity factor as explained previously.

The leak limit curve shown in figure 2 represents minimum temperature requirements at the leak test pressure specified by applicable cocesI3'43 i

The leak test limit curve was determined by methods of references 3 ana 5.

figures 2 and 3 define limits for ensuring prevention of nonductile failure.

1 aw.*m e 7

5.0 ADJUSTED REFERENCE TEMPERATURE From Regulatory Guide 1.99 Rev. 2 the adjusted reference temocrature (ART) for eacn material in the beltline is given by the following expression:

ART = Initial RTNDT + ARTNDT + Mar %n (3)

Initial RT NDT is the reference temperature for the unirradiated material as defined in paragraph NB-2331 of Section !!! of the ASME Boiler and Pressure V n sel Code. If measured values of initial RTNM fer the material in questien are not available, generi: mean values for that Lists of material may be used if there are sufficient test results to establish a Jn and standard aaviatien fer tne class.

t.RT is the mean value of the adjustment in reference temcerature caused l NDT by irradiation and should be calculated as follows:

(4)

ART NDT surface = (CF]f(0.28-0.10 log f)

To calculate ARTNDT at any depth (e.g., at 1/4T or 3/4T), the following attenuation formula was used N

ARTNDT = (ARTNDT surfacele

• O chere x (in inchas) is the capth into the vessel wall measured frem the vessel ,,

inner (aetted) surf ace.

CF ('F) is the chemistr y f acter. If it is not based en surveillance data, it is a function of cepper and nickel content only as given in Table ! :at welds and in Table 11 for base metals (plates and forgings). Linear interpolation F is permitted. In Tables ! and !! "weight-percent cepper" and ' weight-cercent l nickel" are the best-estimate values for the material, which will normally be ,

the eaan of the reasured values for a plate er forging or for weld samples made with the weld wire heat nu-ter that matches the critical vessel weld. As  !

discussed earlier, the CF(*F) may also be calculated using two er more sets of ,

P I

me.*me se 8

\

i credible surveillance data by multiplying each measured ART for each NOT capsule by its corresponding fluence factors, summing the products, and dividing by the sum of the squares of the fluence factors.

Using the information provided in 7able IV (7), the calculated neutron fluence for 10 effective full power years (EFPY) is 1,39 x 1019 n/cm 2 at the vessel inside radius. The corresponding fluence factor is 1091.

Applying Tables ! cnd !! to sl1 the beltline materials indicates that the t 1 ewer snell forging 03 is tu mest limiting base metal material with a chemistry factor of 115'F. Applying these same tables to the girth weld gives a chemistry facter of 50.2*F.

When credible surveillance data from capsules V (7) and V (8) are used, it is founo that the base utal has a chemistry factor of 73.5'T while the weld i metal has s chemistry factor of 93.1*F.

In cecoliance with Regulatory Guide 1.99 Rev. 2, the surveillance capsule data is used, rather than the chemistry table data.

i From the reasured data for Caosule U, the fluence is 8.28 x 1018 n/cm2 ,

and fer this fluence Reg. Guide 1.99 Rev. 2 gives a fluence factor of i 0.9471. Similarly, frem measured data fer Capsule V, the fiuence is 2.49 x  !

18 2 10 n/cm . For this fluence, Reg. Gute 1.99 Rev. 2 gives a fluence '

factor of 0.623.  !

The surveillance data provides ARTNOT shifts for the asial and tangential directions for the base retal for both Capsulec U and Y. Data is also provided fer the weld utal for both of the capsules. Table V shows the mothed for ebtaining the chemistry factors for the base metal and the weld retal. These are 73.5'F for the base eetal and 93.1'T for the weld estal. l Using tre chemistry f acter for the weld eetal, the ART is calculated for 10 l effective full power years (EFPY).

P f

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l Using table IV the fluence is 1.39 x 10 19 n/cm2 at 10 EFPY at the inside radius, and it gives a fluence fector of 1.091. Multiplying this fluence by the chemistry factor of 93.1'T gives an RTNDT shift (ARTNDT) at the surface of 101.6'F. Since the initial RT NDT for the weld is a measured value, and since the standard deviation may be cut in half (when surveillance data are available) the margin = 2 v go 2 , ,,2 has the numerical value of 28'F.

Combining the initial RTNDT, the margin, and the s'hift give. the ART for the girth weld at the inside surface of the vessel:

Su, ' ace ART = 19 + 28 + 101.6 = 148.6'F.

At the 1/4T thickness location, the shift is found by equation 5 to be 89.34*F. B" the same ecuatien, the shift at the 3/4T location is found to be 69.08'F.

Combining the initial RTNOT, the margin, and the shift gives the ART for the girth weld at the 1/4T and 3/4T locations:

1/4T t.ocation ART = 19 + 28 + 89.34 = 136.34*F 3/4T Location ART = 19 + 28 + 69.08 = 116.08'F The above analysis was used to develop the North Anna Unit I heatup and cooldown curves shown in figures 2 and 3, respectively.

Appendix A provides the numerical values on which the heatuo and cooldown curves are bastd.

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MATERIAL PROPERTY Basis CENTRG1ING MA1 TRIAL: CIRLDFERENIAL WEth COPPER CONTINT: 0.086 m ECEIL COMEN: 0.11 m DEITIAL ITET ' LIT RT ET ATTER 10 D7T: 1/47,136.39 3/47,116.18F CtNtVI3 APP!JCABLZ FOR MA1UP RATES UP 70 60*F/HR FOR 1HE SERVICE PERICD UP TO 10 IFFT AE CORAIMS NO MARGIES FOR FCSSIBLZ IE311t!MDIT ERROPS 2500 i i i i I 12AE T137 LIMIT, '

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250 UP 10101777 --

8 0 so too 150 200 260 300 aso 400 450 soo sesneatro truPreatunt (oce.r) l

, Figure 2. North Anna Unit 1 Reactor Coolant Syntas Heatup Limitations

! Applicable for the First 10 EFPY

12

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l MATDlAL PROPERTT BASIS CofrTRCU_TNG MATERI.11: CIRCUMFERENTIAL WELD .

COPPER CONTENT: 0.086 m NICKEL CONTENT: 0.11 m INITIAL RTg: 19 7 RT g AFTER 10 EF7T: 1/4T,136.32 3/47, 116.1 7 I 0 CURVES APPLICABLE FOR COOLDOWN RATES UP M 100 F/HR FOR THE SERVICE PERIOD UP 1010 IFPT AND COMAINS NO MARGINS FOR POSSIBM INSTRUMMT BRORS .

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Applicable for the First 10 EFPY  ;

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Table I CHEMISTRY FACTOR FOR WELDS. 'F Cooper, Nickel. Wt-%

Wt-% 0 0.20 0.40 0.60 0.80 1.00 1.20 0 20 20 20 20 20 20 20 0.01 20 20 20 20 20 20 20

.0.02 21 26 27 27 27 27 27 0.03 22 35 41 41 41 41 41 0.04 24 43 54 54 54 54 54 0.05 26 49 67 68 68 68 68 0.06 29 52 77 -

82 82 82 82 0.07 32 55 85 95 95 95 95 0.08 36 58 90 106 108 108 108 0.09 40 61 94 115 122 122 122 0.10 44 65 97 122 133 135 135 0.11 49 68 101 130 144 148 148 0.12 52 72 103 135 153 161 161 0.13 58 76 106 139 162 172 176 0.14 61 79 109 142 168 182 1 88 0.15 66 84 112 146 175 191 200 0.16 70 88 115 149 178 199 211 0.17 75 92 119 151 184 207 221 0.18 79 95 122 154 187 214 230 0.19 83 100 126 157 191 220 238 0.20 88 104 129 160 194 223 245 0.21 92 108 133 164 197 229 252 0.22 97 112 137 '167 200 232 257 0.23 101 117 140 169 203 236 263 0.24 105 121 144 173 206 239 268 as i.

14

l l

Table I (Cont'd.)

CHEMISTRY FACTOR FOR WELDS, 'F Copper, Nickel. Wt-%

Wt-% 0 0.20 0.40 0.60 0.80 1.00 1.20 0.25 110 126 148 176 209 243 272 0.26 113 130 151 180 212 246 276 0.27 119 134 155 184 216 249 280 0.28 122 138 160 187 218 251 284 0.29 128 142 164 191 222 254 287 0.30 131 146 167 194 225 257 290 0.31 136 151 172 198 228 260 293 0.32 140 155 175 202 231 263 296 0.33 144 160 180 205 234 266 299 0.34 149 164 184 209 238 269 302 C .: 153 168 187 212 241 272 305 0.36 158 172 191 216 245 275 308 0.37 162 177 196 220 248 278 311 0.38 166 182 200 223 250 281 314 0.39 171 185 203 227 254 285 317 0.40 175 189 207 231 257 288 320 me u i.

15

Table II CHEMISTRY FACTOR FOR 8ASE METAL, 'F Cooper, Nickel. Wt-%

Wt-% 0 0.20 0.40 0.50 0.80 1.00 1.20 0 20 20 20 20 20 -

20 20 0.01 20 20 20 20 20 20 20 0.02 20 20, - . 20 20 20 20 20 0.03 20 20 20 20 20 20 20 0.04 22 26 26 26 26 26 26 0.05 25 31 31 31 31 31 31 0.06 28 37 37 37 37 37 37 0.07 31 43 44 -

44 44 44 44 0.08 34 48 51 51 51 51 51 O.09 37 53 58 58 58 58 58 0.10 41 58 65 65 67 67 67 0.11 45 62 72 74 77 77 77 0.12 49 67 79 83 86 86 86 0.13 53 71 85 91 96 96 96 0.14 57 75 91 100 105 106 106 0.15 61 80 99 110 115 117 117 0.16 65 84 104 118 123 125 125 0.17 69 88 110 127 132 135 135 0.18 73 92 115 134 141 144 144 0.19 78 97 120 142 150 154 154 0.20 82 102 125 149 159 164 165 0.21 86 107 129 155 167 172 174 0.22 91 112 134 161 176 181 184 0.23 95 117 138 167 184 190 194 0.24 100 121 143 172 191 199 204 me.-mamie 16 '

Table II (Cont'd.)

CHEMISTRY FACTOR FOR BASE HETAL, 'F Copper, Nickel, Wt-%

Wt-% 0 0.20 0.40 0.60 0.80 1.00 1.20 0.25 104 126 148 176 199 208 2J4 0.26 109 130 151 180 205 216 221 0.27 114 134 155 184 211 225 230 0.28 119 138 160 187 216 233 239 0.29 124 142 164 191 221 241 248 0.30 129 146 167 194 225 249 257 0.31 134 151 172 -

198 228 255 266 0.32 139 155 175 202 231 260 274 0.33 144 160 180 205 234 264 282 0.34 149 164 184 209 238 268 290 0.35 153 168 187 212 241 272 298 0.36 158 173 191 216 245 275 303 0.37 162 177 196 220 248 278 308 0.38 166 182 200 223 250 281 313 0.39 171 185 203 227 254 285 317 0.40 175 189 207 231 257 288 320 me mm ie -

17

TA8L' III REACTOR VESSEL TOUGHNESS DATA (UNIRRADIATED)

~

MWD (b) NMWO UPPER UPPER SHELF SHELF HEAT OR HATERIAL Cu Ni P TNOT RTNOT ENERGY ENERGY MATERIAL CODE NO. SPEC. NO. {%1 (%) (*F) (*F) (FT LB) (FT LB)

{%1 Closure head done 53565-1 A533B C1. 1 -

.60 .014 -31 -26' 87 -

Head flange 4984 A508 C1. 2 -

.82 .011 -40 -40 141 -

Vessel flange 4982 A508 C1. 2 -

.77 .015 -22 -22 161 -

Inlet nozzle 4964 A508 C1. 2 -

.80 .006 -31 -26 100 -

Inlet nozzle 4966 A508 C1. 2 -

.79 .007 -22 -22 88 g,)

Inlet nozzle 4968 A508 Cl. 2 -

.78 .013 -22 3 80 -

Outlet nozzle 4963 A508 C1. 2 -

.80 .010 -13 3 100 -

Outlet nozzle 4965 A508 Cl. 2 -

.81 .006 -22 -22 90 -

Outlet nozzle 4967 A508 C1. 2 -

.81 .006 -4 -4 90 -

Upper shell 05 4952 A508 C1. 2 .16 .74 .013 2 6, 60 -

Intermediate shell 04 4958 A508 Cl. 2 .12 .82 .010 -31 17 , 94 92 Lower shell 03 4979 A508 C1. 2 .15 .80 .009 -13 38 74 85 8ottom Head Segment 53648-3 A5338 C1. 1 -

.63 .016 -31 -13' 65 -

Bottom Head Segment 53648-4 A533B C1. 1 .62 .014 -13 -13 (a) 77 -

Bottom Head Dome 53774 A533B C1. 1 -

.60 .012 -22 - 8, 67 -

Intermediate to Lower Seit 40 Heat 25531 and .086 .11 .020 -13 19 -

102 Shell Girth Weld Seam Seit 89 Flux Lot 1211 (a) Estimated based on U.S. NRC Standard Review Plan, NUREG 0800, Revision 1, July 1981.

(b) Average energy at highest test temperature, % shear not reported.

m - .u. i.

TABLE IV NORTH ANNA UNIT 1 FAST NEUTRON (E > 1.0 MeV) EXPOSURE AT THE PRESSURE VESSEL INNER RADIUS - 0* AZIMUTiiAL ANGLE (a)

Plant Joacific Elapsed -

Beltline Pagion Irradiation Irradiatien Avg. Flux Cumulative Fluence Interval 2 2 (EFPY) (n/cm -sec) (n/cm ) ,

Cycle 1 1.1 5.33 x 10 10 1.89 x 10 18 l Cycle 2 1.9 6.59 x 10 10 3.52 x 10 18 Cycle 3 2.9 4.52 x 10 10 4.88 x 10 18 Cycle 4 3.8 5.07 x 10 10 6.42 x 10 18 Cycle 5 4.8 3.59 x 10 10 7.50 x 10 19 ,

Cycle 6 5.9 3.7d x 10 10 8.83 x 10 18 l Cycle 7 to 30.7 3.92 x 10 10 3.95 x 10 19 expiration of operating License To 20 calendar 46.7 3.92 x 10 10 . 93 x 10 19 years beyond expiration of 3

eperating license (a) Applicable to the peak azimuthal locations (O', 90',180', 270') on the core beltline.

2 -

?

. w ie '

19 ,

6 o

TABLE V CALCULATIONS OF FLUENCE FACTORS BASED ON SURVEILLANCE DATA Fluence Fluence ART 2 NOT ARTNOT x (FF)

Caosule Material (n/cm ) Factor (FF) ('F) ('F) (FF)2 V Forging 03 2.49 x 10 18 0.623 39 24.2970 0.388129 (tangential)

U forging 03 8.28 x 10 8 0.9471 95 89.9745 0.897000 (tangential)

V Forging 03 2.49 x 10 18 0.623 21 13.0830 0.388129 (axial)

U Forging 03 8.28 x 10 10 0.9471 65 61.5615 0.897000 (axial) ,

188.9160 2.57026 ,

9 Forging 03 CF = f]j ff = 73.5'F V Weld Metal 2.49 x 10 18 0.623 78 48.594 0.388129 U Wald Metal 8.28 x 10 18 0.9471 75 71.0325 0.897000 119.6265 1.285129 Weld Metal CF = jf26l=93.085'F 4

mwman i.

20

REFERENCES

1. Reguletory Guide 1.99, Revision 1, "Effects of Residual Elements on Predicted Radiation Drinage to Reactor Vessel Materials," U.S. Nuclear Regulatory Commissier., April 1977.

2. "Proposed Revision 2 to Regulatory Guide .1.99, Radiau:r Damage to Reactor Vessel Matarials," U.S. Nuclear Regulatory Commission February, 1986.

3. "Fracture Toughness Requirements," Branch Technical Position MTEB 5-2, Chapter 5.3.2 in Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Pewer Plants, LWR Edition, NUREG-0800,1981.

4. ASME Boiler and Pressure Vessel Code.Section III, Division 1 -

Appendixes, "Roles for Construction of Nuclear Vessels, Appendix G.

Protection Against Nonductile Failure," pp. 559-564, 1983 Edition.

American Society of Mechanical Engineers, New York,1983.

5. Code of Federal Regulations,10CFR50, Appendix G, "Fracture Toughness Raquirements," V'.S. Nuclear Regulatory Commission, Washington, D.C.,

, Amended May 17,1983(48 Federal Register 24010).

! 6. 'Drecedure for Generating Heatuo and Cooldown Limit Curves," W. T.

Kaiser, Westinghouse Report MT-SME-1419, February, 1980. (Westinghouse proprietary).

7. "Analysis of Capsule U From the Virginia Electric and Power Company North Anna Unit 1 Reactor Vessel Radiation Surveillance Program,"

S.E. Yanichko, et. al., WCAP-11777, February 1988.

8. "Analysis of Capsule V Virginia Electric and Power Company North Anna Unit 1 Reactor Vessel Materials Surveillance Pregiam", A. L. Lowe, Jr.,

et. al . , BAW-1638, May 1987.

am n i.

21

O

?h APPENDIX A .

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