ML20065K148

Nonproprietary Rev 1 of Heatup & Cooldown Limit Curves for Normal Operation for North Anna Unit 1

ML20065K148
Person / Time
Site:	North Anna
Issue date:	11/30/1993
From:	Chicots J, Malone M WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20065K100	List: ML20065K098 ML20065K142 ML20065K148
References
WCAP-13831, WCAP-13831-R01, WCAP-13831-R1, NUDOCS 9404190196
Download: ML20065K148 (29)

Text

..

l ATTACHMENT 4 WESTINGHOUSE REPORT. WCAP-13831, REVISION 1 VIRGINIA ELECTRIC AND ' POWER ' COMPANY 9404190196 940415

.fDR ADOCK'0500o33g PDR

Westinghouse Class 3 (Non Proprietary).

WCAP 13831 Revision 1 HEATUP AND COOLDOWN LIMIT CURVES FOR NORMAL OPERATION FOR NORTH ANNA UNIT 1 J. M. Chicots M. J. Malone November 1993 Work Performed Under Shop Order VKLP-139 Prepared by Westinghouse Electric Corporation for the Virginia Electric and Power Company P

Approved by: 4 AALC 3

T! A. Meyer, Manker Structural Reliability and Plant ufe Optimization WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Division P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355 01993 Westinghouse Electric Corp.

All Rights Reserved

r PREFACE This report was revised to change the margin values for the upper shell forging and weld seams WOSA and WOSB. In addition, the maximum temperature difference between RCS fluid and reactor vessel 1/4-T and 3/4-T locations for heatup and cooldown rates was added.

1

~.

I TABLE OF CONTENTS Section Title Pgle LIST OF ILLUSTRATIONS

. ii I

LIST OF TABLES ii 1

INTRODUCTION 1-2 FRACTURE TOUGHNESS PROPERTIES 1:

3 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS

'2' 4

HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES 4

5 CALCULATION OF ADJUSTED REFERENCE iTUPERATURE 6

6 REFERENCES 16 APPENDlX A: DATA POINTS FOR HEATUP AND COOLDOWN CURVES WITHOUT MARGINS FOR INSTRUMENTATION ERRORS APPENDIX B: PRESSURIZED THERMAL SHOCK EVALUATION RESULTS 9

4

(

i e

]

g-UST OF ll.LUSTRATIONS Fiaure Title Paae 1

North Anna Unit 1 Reactor Coolant System Heatup Umitations 12 (Heatup rate of 20*F/hr) Applicable for the First 30.7 EFPY (Without Margins for Instrumentation Errors) 2 North Anna Unit 1 Reactor Coolant System Heatup Umitations 13 (Heatup rate of 40*FAr) Applicable for the First 30.7 EFPY (Without Margins for instrumentation Errors) 3 North Anna Unit 1 Reactor Coolant System Heatup Umitations 14 (Heatup rate of 60*F/hr) Applicable for the First 30.7 EFPY (Without Margins for instrumentation Errors) 4 North Anna Unit 1 Reactor Coolant System Cooldown Umitations 15 (Cooldown Rates up to 100*F/hr) Applicable for the First 30.7 EFPY (Without Margins for instrumentation Errors)

UST OF TABLES Table Title Pace 1

North Anna Unit 1 Reactor Vessel Toughness Table 8

(Unitradiated) 2 Summary of Adjusted Reference Temperatures (ART's) at 1/4-T 9

and 3/4 T Locations for 30.7 EFPY 3

Calculation of Adjusted Reference Temperatures at 30.7 EFPY 10 for the Umitirg North Anna Unit 1 Reactor Vessel Material-Circumferential Weld Seam i

4 Maximum Temperatures Difference Between RCS Fluid and Reactor 11 Vessel 1/4-T and 3/4 T Locations for Heatup and Cooldown Rates ii

1 2

1

1. INTRODUCTION 1

Heatup and cooldown limit curves are calculated using the most limiting value of RTuo7 # derence nilductility temperature) corresponding to the limitmg beltline region material for the reactor vessel. The most limiting RT r of the material in the core region of the reactor vessel is determined by using the e

unitradiated reactor vessel material fracture toughness properties and estimating the radiation-induced ART 7 The unirradiated RT., is designated as the higher of either the drop weight nil-ductility transition temperature (NDTT) or the temperature at which the material exhibits at least 50 ft-lb of impact energy and 35-mil lateral expansion (normal to the majcr working direction) minus 60*F.

RT 7 increasas as the material is exposed to fast-neutron radiation. Therefore, to find the most limiting RT 7 at any time period in the reactor'slife, ART 7due to the radiation exposure associated with that time period must be added to the original unirradiated RT 1 The extent of the shift in RTum is enhanced by certain chemical elements (such as copper and nickel) present in reactor vessel steels. The Nuclear Regulatory Commission (NRC) has published a method for predicting radiation embiittlement in Regulatory Guide 1.99 Rev. 2 (Radiation Embrittlement of Reactor Vessel Materials)D3 Regulatory Guide 1.99 Revision 2 is used for the calculation of ART values at 1/4-T and 3/4 T locations. T is the thickness of the vessel at the beltline region measured from the clad / base metal interface.

The pressure-temperature limit curves in Figures 1 through 4 of this report correspond to allowable pressure-temperature values at the limiting beltline region of the reactor vessel and do not include margins for instrumentation errors or for pressure differences between the wide-range pressure transmitter and the limiting reactor vessel bottline region.

2.

FRACTURE TOUGHNESS PROPERTIES The fracture-toughness properties of the ferritic material in the reactor coolant pressure bounda.y are determinod in accordance with the NRC Regulatory Standard Review Plar/23 The pre-irradiation fracture-toughness properties of the North Anna Unit 1 reactor vessel are presented in Table 1. The post-irradiation fracture toughness properties of the intermediate to lower shell circumferential weld and the lower shell forging were obtained directly from the North Anna Unit 1 Reactor Vessel Radiation Surveillance Program. Credible surveillance data are currently available for two capsules (Capsules U and V) for North Anna Unit 1.

1

3. CRITERIA FOR ALLOWABLE PRESSURE TEMFERATURE RELATIONSHIPS The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, K,, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, K, for the in metal temperature at that time. K,n is obtained from the reference fracture toughness curve, defined in Appendix G of the ASME CodeA. The K curve is given by the following equation:

in K,n -

26.78 + 1.223

• e

'"5 f'-"*" * ""

(1) 8

where, K-reference stress intensity factor as a ftnction of the metal temperature T and the metal in reference nil-ductility temperature RT,,

Therefore, the goveming equation for the heatup-cooldown analysis is defined in Appendix G of the ASME CodeM as follows:

C

• K + K s K,3 (2) u y

where, K. =

stress intensity factor caused by membrane (pressure) stress K=

stress intensity factor caused by tho thermal gradients n

K,n -

function of temperature relative to the RT of the material um C=

2.0 for Level A and Level B service limits C-1.5 for hydrostatic and leak test conditions during which the reactor core is not critical.

At any time during the heatup or cooldown transient, K is determined by the metal temperature at the tip in of the postulated flaw, the appropriate value for RTum, and the reference fracture toughness curve. The thermal stresses resulting from the temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, Kg, 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 during cooldown, the reference flaw of Appendix G to the ASME Code is assumed to exist at the inside of the vessel wall. During 2

4 cooldown, the controlling location of the flaw is always at the inside of the wall because the thennal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates. Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations.

From these relations, composite limit curves are constructed for each cooldown rate of interest.

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

During cooldown, the 1/4 T vessel location is at a higher 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 cooldown results in a higher va!ue of K n at the 1/4-Tlocation i

for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist so that the increase in K,n exceeds K, the calculated allowable pressure during cooldown will be greater than the g

steady-state value.

The above procedures are needed because there is no direct control on temperature at the 1/4-T location and, therefore, allowable pressures may 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 problem and ensurer cmservative operation of the system for the entire cooldown period.

Three separate calculations are required to determine the limit curves for finite heatup rates. /s is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4 T defect at the inside of the wall. The heatup results in compressive stresses at the inside surface that alleviate the tensile stresses produced by intomal pressure. The metal temperature at the crack tip lags the coolant '

temperature; therefore, the K for the 1/4 T crack during heatup is lower than the K,n for the 1/4 T crack in during steady-state conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist so that the effects of compressive thermal stresses and lower Kgs do not offset each other, and the pressure-temperature curve based on steaoy state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/4 T flaw is considered.

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

The second portion of the heatup analysis concems the calculation of the pressure-temperature limitations 3

for the case in which a 1/4-T deep outside surface flaw is assumed. Unlike the situation at the vessel inside surface, the thermal 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 time (or coolant temperature) along the heatup ramp. Since the thermal stresses a; the outside are tensile and increase with increasing heatup rates, each heatup rate must be analyzed on an individual basis.

Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are pruduced by constructing a composite curve based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration. The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist wherein, over the course of the heatup ramp, the controlling condition switches from 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 Amendment to 10CFR50W has a rule which addresses the metal temperature of the -

closure head flange and vessel flange regions. This rule states that the metal temperature of the closure flange regions must exceed the material unirradiated RT, by at least 120*F for normal operation when the pressure exceeds 20 percent of the preservice hydrostatic test pressure (621 psig for North Anna Unit 1). Table 1 indicates that the limiting unirradiated RT of 22*F occurs in the vessel flange of North Anna Unit 1, so the minimum allowable temperature of this region is 98'F at pressures greater than 621 psig This limit is shown in Figures 1 through 4 whenever applicable.

4. HEATUP AND COOLDOWN PRESSURE TEMPERATURE LIMIT CURVES Pressure-temperature limit curves for normal heatup and cooldown of the primary esctor coolant syste have been calculated for the pressure and temperature in the reactor veM Nho region using the methods discussed in Section 3. Since indication of reactor vessel beltline pressure is nut available on the plant, the pressure difference between the wide-range pressure transmitter and the limrting beltline re must be accounted for when using the pressure-temperature limit curves presented in this report.

Figures 1 through 3 present the heatup curves usr'ng heatup rates of 20*F/hr, 40*F/hr and 60*F/hr applicable for the end-of-license life (30.7 EFPY), respect'vsly. Figure 4 presents the cooldown curves 4

using cooldown rates up to 100*F/hr applicable for the end-of-license life (30.7 EFPY). No margins for possible instrumentation errors are included in the development of heatup and cooldown curves.

Allowable combinations of temperature and pressure for specific temperature change rates are below and to the right of the limit lines shown in Fgures 1 through 4. This is in addition to other aiteria which must be met before the reactor is made entical The reactor must not be made critical until pressure-temperature combinations are to the right of the criticality limit line shown in Fgures 1 through 3. The straight line portion of the criticality limit is at the minimum permissible temperature for the 2485 psig inservice hydrostatic test as required by Appendix G to 10CFR Part 50. De goveming equation for the hydrostatic test is defined in Appendix G to Section lli of the ASME Code as follows:

1.5 K, s; K,

where.

K, is the stress intensity factor covered by membrane (pressure) stress, K,n - 26.78 + 1.233 e ""*"" " *2; f

T is the minimum permissible metal temperature, and RTNDT is the metal reference nil-ductility temperature.

The curved portion of the criticality limit is shifted 40*F to the right of and parallel to the heatup curve as required by Appendix G to 10 CFR Part 50. It should be noted that there are other criteria which must be met before the reactor is made critical. For example, the reactor must not be made critical until a steam bubble is formed in the pressurizer. The leak test limit curve shown on the heatup curve in Figures 1 through 3 represents minimum temperature requirements at leak test pressures ranging from 2000 psig to 2485 psig. Tne leak test limit curve was determined by the sama method used to compute the inservice hydrostatic test temperature. This method used a 1.5 safety factor on the pressure stress intensity factor as explained previously.

The leak limit curve shown in Fgures 1 through 3 represents minimum temperature requirements at the leak test pressure specified by applienhle Codes *'I. The leak test limit curve was determined by methods of References 2 and 4.

Fgures 1 through 4 define limrts for ensuring prevention of nonductile failure for the North Anna Unit 1 5

reactor vessel.

The data points used to develop the heatup and cooldown pressure-temperaturo limit curves shown in Figures 1 through 4 are presented in Appendix A.

The maximum temperature difference between the RCS fluid and the vessel at the 1/4-T and 3/4-T locations for both heatup and cooldown cases were calculated and the results are provided in Table 4.

5. CALCULATION OF ADJUSTED REFERENCE TEMPERATURE From Regulatory Guide 1.99 Rev. 201 he edjusted reference temperature (ART) for each material in the t

beltline is given by the following expression:

ART - Initial RTuor + ARTuor + Margin (3) initial RT i

uoy s the reference temperature for the unirradiated material as defined in paragraph NB-2331 of l

Section ill of the ASME Boiler and Pressure Vessel Code. If measured values of initial RTuo for tha material in question are not available, generic mean values for that class of material may be used if there are sufficient test results to establish a mean and standard deviation for the class.

ART is the mean value of the adjustment in reference temperature caused by irradiation and should be uor calculated as follows:

ARTuor = [CF)

• f 5 8 W 0 (4)

To calculate ARTuor at any depth (e.g., at 1/4-T or 3/4-T), the following formula must first be used to attenuate the fluence at the specific depth.

f,,3 = f.

• e '

(5) b where x (in inches) is the depth into the vessel wall measured from the vessel clad / base metal interface.

The resultant fluence is then put into equation (4) to calatlate ARTuoy at the specific depth. The calculated surface fluence at end-of-license life (30.7 EFPY) for North Anna Unit 1 is 3.95 x 10" n/cm2 l'I.

CF (~F) is the chemistry factor, obtained from Tables in Reference 1, using the values of the copper and

-l 6

nickel content as reported in Table 1. If plant-specrfic surveillance' data has been deemed credible per I

~ Reg. Guide 1.99, Revmon 2, it may be considered in the calculation of the chemistry factor.

-All materials in the belthne region of North Anna Unit 1 reactor vessel were consdered in determining the -

~

limiting material.- The results of the ART's at 1/4-T and 3/4-T are summarized in Table 2. From Table 2,

- it can be seen that the limiting material is the circumferential weld seam for heatup and cooldown curves apphcable up to end-of-license life (30.7 EFPY). Sample calculations to determine the ART values for the

- circumferential weld seam at endd-license life (30.7 EFPY) are shown in Table 3.

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TABLE 1 NORTH ANNA UNIT 1 REACTOR VESSEL TOUGHNESS TABLE (Unirradiated)

Material Description CU (%)

NI(%)

lRT m(*F) u Closure Head Flange 0.82

-40 (b)

Vessel Flange 0.77 22 (b)

Upper Shell Forging 05 0.16 0.74 6 (c)

Intermediate Shell Forging 04 0.12 0.82 17 (a)

Lower Shell Forging 03 0.15 0.80 38 (a)

Intermediate Shell to Lower Sheli 0.086 0.11 19 (a)

Circumferential Weld Seam, WO4 Weld Seam, WOSA 0.30 0.10 0 (c)

Weld Seam, WO58 0.11 0.10 0(c)

(a)

The initial RT values for the plates and circumferential weld are measured values.

um (b)

Initial RT,or values for the closure head and vessel flange were estimated per U.S. NRC Standard Review Plan A. These values are used for considering flange requirements for the heatup/cooldown curves l.

H (c)

Initial RT,e7 values are estimated per U.S. NRC Standard Review Plan. For margin A

calculations for these materials, o, = 30'F for forging and o, - 20'F for wolds per BAW 1911, Rev.1

• Reactor Pressure Vessel and Surveillance Progrem Materials Ucensing information for North Anna Units 1 and 2", A. L lowe, dated August 1986.

Copper and nickel concentrations are taken from North Anna Unit 1 Table 7 " Beltline Materials" of BAW-2168, Revision 1, " Response to Generic Letter 92-01 for Virginia Electric and Power Company North Anna Units 1 and 2", dated September 1992.

8

i

. TABLE 2 -

SUMMARY

OF ADJUSTED REFERENCE TEMPERATURES'(ART's)

AT 1/4 T and 3/4-T LOCATIONS FOR 30.7 EFPY Material 1/4-T ART (*F) 3/4 T ART (*F) l Upper Shell Forging 05 140.3 117.3 Inter Shell Forging 04 158.1 136.8 Lower Shell Forging 03 215.2 (146.5) 186.7 (128.3)

Circumferential Weld Seam, WO4 137.5 (162.9)*

119.1 (139.9)*

Weld Seam, WOSA 1'43.4 111.2 Weld Seam, WOSB 82.4 65.4 ART numbers within ( ) are based on chemistry factors' calculated using surveillance capsule data.

"These ART numbers were used to generate heatup and cooldown curves.

9 1

l J

TABLE 3 CALCULATION OF ADJUSTED REFERENCE TEMPERATURES AT 30.7 EFPY FOR THE LIMITING NORTH ANNA UNIT 1 REACTOR. VESSEL MATFRIAL-CIRCUMFERENTIAL WELD SEAM Regulatory Guide 1.99 Revision 2 30.7 EFPY Parameter 1/4-T 3/4-T Chemistry Factor, CF (*F) 93.09 93.09 Fluence, f (10" n/cm')(*

2.492 -

0.992 Fluence Factor, ff 1.245 0.998 ART, = CF x ff (*F) 115.9 92.9 initial RT 7, I (*F) 19 19 Margin, M (*F)

• 28 28 Revision 2 to Rcgulatory Guide 1.99 Adjusted Reference Temperature, 162.9 139.9 ART = lnitial RT 7 + ART., + Margin (a)

Fluence, f, is based upon f,, (10" n/cm, E>1 Mov) = 3.95 at end-of-license life (30.7 EFPY).

2 The North Anna Unit 1 reactor vessel wall thickness is 7.677 inches at the beltline region.

(b)

Margin is calculated as, M = 2 [ o' + o,8]". The standard deviation for the initia! RT.,

margin term, o,, is assumed to be 0*F since the initial RTor is a measured value. The standard deviation for ARTer term,0, is 28'F for the weld, except that o, need not exceed 3

0.5 times the mean value of ART.1 o, is 14*F for the weld (half the value) when surveillance data is used.

10

TABLE 4 MAXIMUM TEMPERATURES DIFFERENCE BETWEEN RCS FLUID AND REACTOR VESSEL 1/4-T AND 3/4 T LOCATIONS FOR HEATUP AND COOLDOWN RATES Water 1/4 T 3/4 T AT-AT Rate Time Temp.

Temp. =

Temp.

@ 1/4-T

.@ 3/4-T Case.

'F/hr

.(sec)

(*F)

(* F)

(*F)

(*F)

(* F)

Heatup.

60 28,800 550 535 519 15 31 Cooldown 100-5940 385 408 434 23 100 6120 380 403 429 49 e

,l

)

MATERIAL PROPERTY BASIS LIMITING MATERIAL: CIRCUMFERENTIAL WELD SEAM LIMITING ART AT 30.7 EFPY: 1/4-T, 162.9'F 3/F

' 39.9'F i

2.,500 i,,,,,,,,,,,,

,,,,,,i,,,,i,

/

/

[

LEAK TEST LIMIT x 2,250

}

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n l

l l

O 2.,000

/

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l i

UNACCEPTABLE

/

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h 1.,750 OPERATION

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l

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L 1,500

/

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D l

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ACCEPTABLE

/

OPERATION O

1'250 HEATUP RATE.\\

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LA TO 2D ' F/ HR

/

1.,000

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/

Cd 750

/

O

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~

500 -

c

CRITICALITY LIMIT BASED ON 250 -- INSERVICE NfDROSTATlC TEST

. /

I'. TEMPERATURE C292 h) FOR THE 0 ~~

SERVICE PERIOD UP TO 30.7 EFPr O

50 100 150 200 250 300 350 400 450 500 Indicated Temperature (Deg.

F)

Figure 1.

North Anna Unit 1 Reactor Coolant System Heatup Umitations (Heatup rate of 20*F/hr)

Applicable for the First 30.7 EFPY (Wrthout Margins for instrumentation Errors) 12

l MATERIAL PROPERTY BASIS LIMITING MATERIAL CIRCUMFERENTIAL WELD SEAM LIMITING ART AT 30.7 EFPY: 1/4 T, 162.9'F 3/4-T, 139.9'F 2,500 iiiiiiiiiiii, I I I iI I I I I I I I i j

LEAK TEST LIMIT

[

/

x f

2.,250

/

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i I

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O UNACCEPTABLE

/

/

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/

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b'

/

/

L 1,500

/

/

/

/

ACCEPTABLE 0

/

OPERATION e

1,250 f

HEATUP RATE

/

\\

UP TO 40

• F/ t R j

N

/

1.,000

/

Oo

/

e (d

750

/

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f 500

~'~

CRITICALITY LIMIT BASED ON 250 -- 1NSERV1CE HYDROSTATlC TEST

_ 7 2 TEMPERATURE (292 b FOR THE SERVICE PERIOD UP TO 30.7 EFPY 0

O 50 100 150 200 250 300 350 400 450 500 Indicated Temperature (Deg.

F)

Figure 2.

North Anna Unit 1 Reactor Coolant System Heatup Limitations (Heatup rate of 40*F/hr)

Applicable for the First 30.7 EFPY (Without Margins for instrumentation Errors) 13

r i

MATERIAL PROPERTY BASIS LIMiriNG MATERIAL: CIRCUMFERENTIAL WELD SEAM LIMITING ART AT 30.7 EFPY: 1/4-T, 162.9'F 3/4-T, 13S.9'F 2,500 iiiiiiiiiiiii iiiiiiiiiiiii i

/

LEAK TEST LIMIT -

[

[

[

x 2.,250 l

l I

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g i

w l

l L

1,500 U

f

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ACCEPTABLE e

1.,250

[

[

OPERATION m.ATUP RATE k

1 UP TO SO

• F/ HR \\

c x

/

1.,000 y

f e

/

H (d

750 O

/

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1 f

500 CRITICALITY LIMIT BASED ON 250 - inservice wronosTATIC TEST

, /

TEMPERATURE C292 b FOR THE

SERVICE PERIOD UP TO 30,7 EFPY 0

O 50 100 150 200 250 300 350 400 450 500 Indicated Temperature CDec.

FJ 1

Figure 3.

North Anna Unit 1 Reactor Coolant System Heatup Limitations (Heatup rate of 60'F/hr) i Applicable for the First 30.7 EFPY (Wdhout Margins for instrumentation Errors) 14 l

5 MATERIAL PROPERTY BASIS 4

- LIMITING MATERIAL: CIRCUMFERENTIAL WELD SEAM

'LIMirlNG ART AT 30.7 EFPY: 1/4 T 162.9'F -

3/4 T,.139.9'F 2.,500 2.,250

)

I I

O

0) 2,000

,/

W O.

y 1.,750

/

UNACCEPTABLE.

OPERATlON

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.-OPERATION i

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()

RATES h/ m.-

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==

--.w

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ao

g: y 40

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((((

BD 250 _.___

100 0

0 50 100 150 200 250 300 350 400 -450 50'O.

Indicated Temperature CDeg. F}

u 1

l 1

Figure 4.

North Anna Unit 1 Reactor Coolant System Cooldown Umitations (Cooldown rates up to 100*Fhir) Applicable for the First 30.7 EFPY (Without Margins for Instrumentation Errors)

i l

15 1

.m.

1

6. ' REFERENCES

[1].

Regulatory Guide 1.99, Revision 2, " Radiation Embnttlement of Reactor Vessel Materials". U.S. Nuclear Regulatory Commission, May,1988.

[2]

" Fracture Toughness Requirements", Branch Tedinical Position MTEB 5 2, Chapter ;

5.3.2 in Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition, NUREG-0800,1981.

[3]

ASME Boiler and Pressure Vessel Code. Section lil, Division 1 - Appendixes, " Rules for Construction of Nuclear Power Plant Components, Appendix G, Protection Against Nonductile Failure", pp. 558-563,1986 Edition,' American Society of Mechanical Engineers, New York,1986.

[4]

Code of Federal Regulations,10CFR50, Appendix G, " Fracture Toughness Requirements". U.S. Nuclear Regulatory Commissa, Washington, D.C.,

Federal Register, Vol. 48 No.104, May 27,1983.

[5]

WCAP-8771, " Virginia Electric and Power Company North Anna Unit No.1 Reactor Vessel Radiation Surveillance Program", J. A. Davidson and J. H. Phillips September 1976. (Westinghouse Proprietary Class 3)

[6]

WCAP-11777, " 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., February 1988. (Westinghouse Proprietary Class 3)

[7]

Material Certification Report, Rotterdam Order No. 30661, Lab No. J 575.

t T

16 T'

-4m

m. a7 g.

e.

i APPENDIX A DATA POINTS FOR HEATUP AND COOLDOWN CURVES (Without Margins for instrumentation Errors) l A-0 t

q t

North Anna Unit I fleatup/Cooldown Cbrve Desa WTD10UT mar 8 ns i

Cooldown Curves Steady State 20 DEO CD 40 DEO CD 60 DEO CD.

100 DEG CD -

T P

T P

T P

T P

T P

85 547.28 85 513.64 85 479.44 85 444.59 85 373.22 90 552.65 90 519.15 90 485.12 90 450.47 90. 379.55 95

'558.42 95 525.11 95 491.20 95. 456.85'

- 95 386.47 100 564.63 100 531,51 100 4 97.82 100 463.73 100 393.94 105 57130 105 53832 105 505.00 105 471.20 105 402.08 110 578A7 I10 545.75 110 ' 512.72 110 479.24 110 410.82 115 586.06 115 553.77 115 521.07 115 487.96 115 42038 120 59435 120 56238 120 530.05 120 497.26 120 430.71 125 603.26 125 571.68 125 539.66.

125 507.42 125. 441.85 130 : ' 612.85 130 :. 581.66 130 550.10 130 51837 130 453.95 135 ~ ' 623.15 135 592.32 135. 561.39 135 : 530.21 135 467.07 140 ~ 634D8 140 603.90 140 573.52 140 542.85 140 ' 481.22 145 645.99 145 616.38 145 586.50 145 556.63 ~

145-4 %.47 150 658.79 150 629.77 150 600.59

' 150 571.46 150 513.00 155 672.54 -

155 644.11 155 615.00 155 58736 155 530.89 160 687.19 -

160 659.65 160 632.00 160. - 604.60 160 550.08 165 703.09 165 ' 67635 165 - 649.64 165 - ' 623.22 165 ' 570.94 170 720.16 170 694.21 170 668.61 170 643.11

!?O.59330 175 '73839 175 713.59 175 688.90 175 - 664.74 175 617.60 1MD 758.16

. I80 734.23 180 710.90 180 687.84 180 - 643.63 :

185 779.19 l 185 756.66 185 734A2 185 712.94 I85 671.90

-190 - 802.03 190 780.55 190 759.89 190 739.76

.190 - 702.20.

195 82634 195-806.48 195 787,16 195

-768.82 195. 734.90 200 852.71 200 834.16 200 816.63 200 799.99 200. ' 770.25 f 205 880.84 205 864.11 205 848.28 205 833.52 205- ' 808.29 210 911.22 -

210 896.19 210 882.22 210' 869.58 210" B49.22 215 943.79 215 930.62 215 918.77 215 908.64 '

' 215 ' 893.33-220 978.72 220 M7.62 220 - 958.26 220 950.46 - ~ 220 940.81 225 1016.23.

225 1007.61 225 1000.60 225 995.48 225 ' 991.95 '

230 1056.54 230 -1050.44 230 1046.08

'230 1043.85 230 1046.95 235 1100.08 235 1096.47 235 1094.99 235 1095.93 240 - l 146.64 240 1145.88 245 1196.67 250 1250.15 255 1307.90 260.1369.68 265 1435.87 270 1507.19 275 1583.43 280 1664.94 285 1752.74 290 A846.59 295 1946.83 300 2054.44 305 2169.27 310 2292.07 315 2423.12

.i A1 t

-%.r tw w

i North Anna Unit 1 Hestup/Cooldomi Curve Data WITHOI.TT mar 8 ins Heatup Curves 20 DEG HU Criticality Umit 40 DEO HU Cnucality Limit T

P T

P T

P T

P 85 541.05 292 0

85 523.08 292 0

90 544.65 292 544.65 90

$23D8 292 523.08 95 551.19 292 551.19 95 523.91 272 523.91 100 558.02 292 558.02 100

$26.71 292 526.71 105 566.12 2 92 566.12 105 531.26 292 531.26 110 574.67 292 574.67 110 536.M 292 536.96

!!5 - 583.99 292 583.99 115 543.97 292 543.97 120 594.05 292 594.05 120 551.92 292 551.92 125 603.26 292 603.26 125 560.88 292 560.88 130 612.85 292 612.85 130 570.68 292 570.68 135 623.15 292 623.15 135 581.44 292 581.44 140 634.08-292 634D8 140 592.98 292 592.98 145 645.99 2 92 645.99 145 605.64 292 605.64 150 658.79 292 658.79 150 619.27 292 619.27 155 672.54 292 672.54 155 633.87 292 633.87 160 687.19 292 687.19 160 649.71 292 649.71 165 703.09 292 703.09 165 666.81 292 666.81 170 - 720.16 292 720.16 170 685D0 292 685.00 175 73839 292 73839 175 704.79 292 704.79 180 758.16 292 758.16 12 725.84 292 725.84 1 85 779.19 292 779.19 185 748.72 292 748.72 190 802.03 292 802.03 190 773.22 292 773.22 195 826.34 292 82634 195 799.50 292 799.50 200 852.71 292 852.71 200 827.64 292 827.64 205 880.84 292 880.84 205 858.12 292 858.12 210 911.22

.292 911.22 210 890.69 292 890.69 215 943.79 292 943.79 215 925.64 292 925.64 220 978.72 292 978.72 220 963.17 292 963.17 225 1016.23 292 1016.23 225 1003.73 292 1(D3.73 230 1056.54 292 1056.54 23 0 1047.10 292 1047.10 135 1100.08 292 1100D8 235 1093.62 292 1093.62 240 1146.64 292 II46.64 240 1143.61 2 92 1143.61 245 1196.62 292 11 %.62 245 1196.62 292 11 %.62 250 1247.04 292 1247.04 250 1246.41 292 1246.41 255 1301.47 295 1301.47 255 1297.66 295 1297.66 260 1359.75 300 1359.75 260 1352.70 300 1352.70 265 1422.22

'305 1422.22 265 1411.$5 305 1411.55 270 1489.!!

310 1489.11 270 1474.75 310 1474.75 275 1561.00 315 1561.00 275 1542.37 315 1542.37 280 1637.87 320 1637.87 280 1614.86 320 1614.86 285 172033 3 25 172033 285 16923 9 325 1672.59 290 1808.46 330 1808.46 290 1775.78 330 1775.78 295 1903.06 335 1903D6 295 1864.70 335 1864.70 300 20N.25 340 2004.25 300 1960.06 340 1960.06 30$ 2112.48 345 2112.48 305 2061.90 345 2061.90 310 2227.78 350 2227.78 310 2170.71 350 2170.11 315 235131 355 2351.31 315 2286.98 355 2286.98 320 2482.97 360 2482.97 320 2410.81 360 2410.81 i

A-2

L Nonh Anna Unit i Heacup/Cooldown Curve Data WITHOUT mar 8 ins -

. Ileatup Curves bak Test Data -

(for all heatup curves) 60 DEO HU

- Criticality Limit T

P T

-P' T

P.

85 507.18 -

292 0

270 2000 90 507.18 292 507.18 292

- 2485 7

95 507.18 292 507.18 100 507.18 292 5tT7.18 105 507.82 292 507.82

.I10 510.04 292 510.04

!!5 513.81-292 513.81 120

$18.78 -

292

'518.78' 125 525.00

-292 525.00 130 532.25 292 532.25 135

.540.50 292 540.50

- t 140 549.81.

292 549.81 145 560.20 292 560.20 150 571.56 292

'571.56 155 583.91

'292 583.91 160 597.43 ~

292 597.43 165 : 612.13 292 612.13 170 - 628.00 292-628.00 175 645.05 292 645.05 180 663.57 292 663.57.

.185 68338 292 68338 190 704.87 292 704.87 r

195-727.84 292 727.84 200 752.74, 292 752.74 205 77933 292 77933 210 808.11 292 : 808.11 215 838.87 -

_ 292 838.87 -

220. 871.89

_-292 871.89 225. 907.59 292 ' 907.59 230 945.78 292 945.78 235.. 986.80 292 986.80 -

-1 240 1030.80 292 1030.80 245 1078.06 292 1078.06 250. I128.77 292. I128.77 255 1183.17 295 1183.17 260 1241.50 300 1241.50 265 1304.05 305 1304.05 270 1371.17 310 1371.17 275 1443.05 3 15 1443.05 280 1520.15 320 1520.15-285 1602.65 325 1602.65 290 1691.00 330 1691.00 295 ' 1785.48 335 1785.48 300 - 1886.69.

340 1886.69 305 1994.52 345 1994.52 310 2110.10 350 ' 2110.10 315 2229.35 355 222935 320 2346.49 360 2346.49 -

325 " 247130 365 247130 i

+

r a

A-3 p

i' t

s 4

r

,.s v1

I APPENDIX B PRESSURIZED THERMAL SHOCK EVALUATION RESULTS B-0

A limiting condition on reactor vessel integnty known as Pressurized Thermal Shock (PTS) may occur during a severe system transient such as a Loss-Of Coolant-Accident (LOCA) or a steam line break.

Such transients may dialienge the integnty of a reactor vessel under the following conditions:

severe overcooling of the inside surface of the vessel wall followed by high repressurization; significant degradation of vessel material toughness caused by radiation embrittlement; and the presence of a critical-size defect in the vessel wall.

In 1985 the Nudear Regulatory Commission (NRC) issued a formal ruling on PTS. It established screening criteria on pressurized water reador (PWR) vessel embrittlement as measured by the nil-ductility reference temperature, termed RT,n". RT,n screening values were set for beltline axial welds, forging or plates and for beltiine circumferential weld seams for the end-of-license plant operation. The screening criteria were determined using conservative fracture mechanics analysis techniques. All PWR vessels in the United States have been required to evaluate vessel embrittlement in accordance with the enteria through end-of-license. The NRC has amended its regulations for light water nudear power plants to change the procedure for calculating radiation embrittlement. The revised PTS Rule was published in the Federal Register, May 15,1991 with an effective date of June 14,199154 This amendment makes the procedure for calculating RT,n values consistent with the methods given in Regulatory Guide 1.99, Revision 283 The PTS Rule requires that the PTS submittal be updated whenever there are changes in core loadings, surveillance measurements or other information that indicates a significant change in projected RT,n values.

The Rule outlines regulations to address the potential for PTS events on pressurized water reactor vessels in nudear power plants that are operated with a license from the United States Nudear Regulatory Commission (USNRC). PTS events have been shown from operating experience to be transients that resuit in a rapid and sevwe cooldown in the primary system coinddent with a high or increasing primary system pressure. The PTS concem arises if one of these transients acts on the beltline region of a reactor vessel where a reduced fracture resistance exists because of neutron irradiation. Such an event may resutt in the propagation of flaws postulated to exist near the inner wall surface, thereby potentially affecting the integrity of the vessel.

The Rule establishes the following requirements for all domestic, operating PWRs:

B-1

All plants must submit projected values of RT,3 for reactor vessel beltline materials by giving values for time of submittal, the expiration date of the operating license, and the projected expiration date if a change in the operating license or reacall has been requested. This assessment must be submitted within six months after the effective date of this Rule if the value of RT,n for any material is projected to exceed the screening criteria. Otherwise, it must be submitted with the next update of the pressure temperature limits, or the next reactor vessel surveillance capsule report, or within 5 years frcm the effective date of this Rule diange, whichever comes first. These values must be calculated based on the methodology specified in this rule. The submittal must include the following:

1) the bases for the projection (including any assumptions regarding core loading patterns), and 2) copper and nickel content and fluence values used in the calculations for each bettiine material. (if these values differ from those previously submitted to the NRC, justification must be provided.)

Using the prescribed PTS Rule methodology, RT,n values were generated for all beltline region materials of the North Anna Unit 1 reactor vessel as a function of present time (11 EFPY) und end-of-life (30.7 EFPY) fluence values.

Table B 1 provides a summary of the RT,3 values for all beitline region materials for 11 EFPY and end-of-license (30.7 EFPY), using the PTS Rule. As shown in Table B 1 all the RT,n values remain below the NRC screening values for PTS using the fluence values for the present time (11 EFPY) and the projected fluence values for the end of-license (30.7 EFPY).

B-2 s

r1 TABLE B-1 NORTH ANNA UNIT 1 RT,n VALUES ('F) FOR 11 EFPY AND 30.7 EFPY MATERIAL 11 EFPY 30.7 EFPY Upper Shell Forging 05 1052 132.9 intermediate Shell Forging 04 146.0 167.4 Lower Shell Forging 03 199.1 (1532) 227.7(171.5)

Circumferential Weld Seam, WO4 140.9 (177.9)

.155.7 (201.0)

Weld Seam, WOSA 124.3 155.9 Weld Seam, WOSB 90.6 104.0 Note values in ( ) were calculated based upon surveillance capsule data.

REFERENCES

[B1]

10CFR Part 50, " Analysis of Potential Pressurized Thermal Shock Events," July 23,1985.

[B2]

10CFR Part 50.61, " Fracture Toughness Requirements for Protection Against Pressurized

]

Thermal Shock Events," May 15,1991. (PTS Rule)

(B3]

Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials," U.S.

Nuclear Regulatory Commission, May 1988.

B3 f