ML20196F701

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BVPS Unit 2 Heatup & Cooldown Limit Curves During Normal Operation at 15 EFPY Using Code Case N-626
ML20196F701
Person / Time
Site: Beaver Valley
Issue date: 01/31/1999
From: Laubham T
WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20137V023 List:
References
WCAP-15139, NUDOCS 9906290284
Download: ML20196F701 (50)
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Text

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-15139 Beaver Valley Unit 2 Heatup and Cooldown Limit Curves During Normal Operation at 15 EFPY.

Using Code Case N-626 T.J. Laubham i

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January 1999 Work Performed Under Shop Order DQYP-139A Prepared by the Westinghouse Electric Company for the Duquesne Light Company Approved:

C. H. Boyd, Manager Engineering and Materials Technology

/

Approved:

D. M. Trombola, Manager l

Mechanical Systems Integration Westinghouse Electric Company Energy Systems P.O. Box 355 Pittsburgh, PA 15230 4355 C1999 Westinghouse Electric Company All Rights Reserved 990656284 990617

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PREFACE

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This report has been technically reviewed and verified by:

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E. Terek _

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TABLE OF CONTENTS LISTOFTABLES...................................................................................................................................iii LISTOFFIGURES..............................................................................................................................iv EXECUTIVE

SUMMARY

...................................................................................................................... v 1

INTRODUCTION....................................................................................................................1 i

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1 2

PURPOSE..................................................................................................................................2 1,

3 FRACTURE TOUGHNESS PROPERTIES................................................................................ 3 I

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4 CRITERIA TOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS............ 7 i

5 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE.....................................10 t

6 HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES................14 7

REFERENCES......................................................................................................

....... 22 0:\\4468. doc 1b.012699 January 1999

lii LIST OF TABLES Table 1 -

~ Calculation of Average Cu and Ni Weight Percent................................................... 3 Table 2 Interpolation of Chemistry Factors from Regulatory Guide 1.99, Revision 2, Posi tion 1.1............................................................................................ 4 Table 3 Calculation of Chemistry Factors Using Surveillance CapsuleData................................................................................................................5 Table 4 Reactor Vessel Beltline Region Material Properties Used in Calcula tions............................................................................................ 6 Table 5 Margins for Adjusted Reference Temperature (ART) Calculations per Regulatory Guide 1.99, Revision 2............................................................................ 11

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l Table 6 Calculation of ART Values for the Limiting Beaver Valley Unit 2 Reactor Vessel Material-Intermediate Shell Plate B9004-1................................................. 12 Table 7 Summary of ART Values at the 1/4T and 3/4T Locations.................................13 Table 8 15 EFPY Heatup and Cooldown Curve Data Points (With Instrumentation Error Margms of 10*F and 60 psig).......................................................................... 18 Table 9 15 EFPY Heatup and Cooldown Curve Data Points (Withcut Instrumentation Error Margins)........................................................................... 21 l

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iv LIST OF FIGURES Figum 1 Beaver Valley Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rates up to 60 F/hr) Applicable for the First 15 EEPY (With Margins of 10*F and 60 psig for Instrumentation Errors)............................16 Figure 2 Beaver Valley Unit 2 Reactor Coolant System Cooldevn Limitations (Cooldown Rates up to 100*F/hr) Applicable for the First 15 EFPY (With Margms of 10*F and 60 psig for InstrumentatiO4 Errors)........................~.17 Includes vessel flange requic. ments of 130*F and 561 psig Per 10CFR50, Appendix G Figure 3.

Beaver Valley Unit 2 Reactor Coolant System Heatup limitations (Heatup Rates up to 60*F/hr) Applicable for the First 15 EFPY (Without Margins for Instrumentation Errors)................................................................. 19 Figum 4 Beaver Valley Unit 2 Reactor Coolant System Cooldovvn Limitations (Cooldown Rates up to 100 F/hr) Applicable for the First 15 EFPY (Without Margms for Instrumentation Errors).................................................... 20 Includes vessel flange requirements of 120'F and 621 psig Pw 10CFR50, Appendix G o:\\4468. doc 1b-012699 armaty1999

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v EXECUTIVE

SUMMARY

This report provides the methodology and results of the generation of heatup and cooldown pressure temperature limit curves for normal operation of the Beaver Valley Unit 2 reactor vessel. These curves were generated based on the latest available reactor vessel information

. (Capsule V analysis, WCAP-14484).

The Beaver Valley Unit 2 heatup and cooldown pressure-temperature limit curves have been updated based on:

the latest weight percent copper and nickel data for the Beaver Valley Unit 2 reactor e

vessel beltline materials (from the Beaver Valley Unit 2 response to Generic Letter 92-01, Revision 1, Supplement 1, Reactor Vessel Structural Integrity),

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the current projected vessel fluence values based on the neutron dosimetry results from the two surveillance capsules removed to date (U and V), using the ENDF/B-VI data set and updated integrated analytical predictions, credible surveillance capsule data from two surveillance capsule specimens tested to e

date, and use of the ASME Code Case N-626 which allows the use of the K, methodology.

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o:\\4MB h lb O12699 January 1999

r 1

1 INTRODUCTION Heatup and cooldown limit curves are calcu:ated using the adjusted RT, (reference nil-ductility temperature) corresponding to the limiting beltline region material of the reactor vessel. The adjusted RT, of the limiting material in the core mgion of the reactor vessel is determined by using the unirradiated reactor vessel material fracture toughness properties,

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estimating the radiation-induced ART, and adding a margin. The unirradiated RT,is designated as the higher of either the drop weight nil-ductility transition temperature (NDIT) or the temperature at which the material exhibits at least 50 ft-lb of impact energy and 35-mil lateral expansion (normal to the major working direction) minus 60*F.

RT, increases as the material is exposed to fast-neutron radiation. Therefore, to find the most limiting RT, at any time period in the reactor's life, ART, due to the radiation exposure associated with that time period must be added to the unirradiated RT,(IRT

). The extent of the shift in RT, 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 embrittlement in Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials."" Regulatory Guide 1.99, Revision 2, is used for the calculation of Adjusted Reference Temperature (ART) values (IRT, + ART, + margins for uncertainties) at the 1/4T and 3/4T locations, where T is the thickness of the vessel at the beltline region measured from the clad / base metal interface. The most limiting ART values are used in the generation of heatup and cooldown pressure-temperature limit curves.

WCAP-15139 January 1999 oA4468. doc 1b.012699

o 2

2 PURPOSE The Duquesne Light Company contracted Westinghouse to regenerate the heatup and cooldown curves documented in WCAP-14485" using K, in place of K, for the calculation of the stress intensity factors. The heatup and cooldown curves from WCAP-14485 were generated with and without margms for instrumentation errors and included a hydrostatic leak test limit curve from 2485 to 2000 psig and pmssure-temperature limits for the vessel flange regions per the mquirements of 10 CFR Part 50, Appendix G".

The purpose of this report is to document the generation of new 15 EFPY Pressure-Temperature (P-T) limit curves utilizing the K, methodology'"'. The P-T curves are developed with the identical adjust reference temperature (ART) values used in WCAP-14485. In addition, this report included all the original text and tables from WCAP-14485 with appropriate changes corresponding to K,. The use of K, will add substantial pressure margin to the heatup and cooldown curves documented in WCAP-14485. This increase in allowable pressure is presented in Section 6 of this report.

WCAP-15139 January 1999 o:\\4%84oc1b-012699

3 3

FRACTURE TOUGHNESS PROPERTIES The fracture-toughness properties of the ferritic material in the reactor coolant pressure boundary are determined in accordance with the NRC Standard Review Plan". The beltline material properties of the Beaver Valley Unit 2 reactor vessel presented in Table 1 are from References 3 through 11.

The average Cu and Ni values were used to calculate chemistry factor (CF) values per Tables 1 and 2 of Regulatory Guide 1.99, Revision 2. (See Table 2 on the following page.) Additionally, l

surveillance capsule data is available for two capsules (Capsules U and V) already removed from the Beaver Valley Unit 2 reactor vessel. This surveillance capsule data was used to calculate chemistry factor (CF) values (Table 3) in addition to those calculated per Tables 1 and 2 of Regulatory Guide 1.99, Revision 2. The closure head flange and vessel flange material properties were obtained from Reference 14.

Table 1 Calculation of Average Cu and Ni Weight Percent" Inter. Shell Plate Inter Shsll Plate Lower Shell Plate Lower Shell Plate B9004-1 B9004-2 59005-1 B9005 2 Weld Metal'*

Ref.

Cu %

Ni %

Cu %

Ni %

Cu %

Ni %

Cu %

Ni %

Cu %

Ni %

3 0.07 0.53 4

0.07 0.59 5

0.08 0.59 6

0.07 0.58 7

0.06 0.57 0.06 0.56 0.08 0.57 0.07 0.56 8

0.05 0.56 9

0.04 10 0.04 0.08 10 0.04 0.09

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10 0.05 0.07 10 0.05 0.07 10 0.08 0.07 11 0.046 0.064 Avg.

0.065 0.55 0.06 0.57 0.08 0.58 0.07 0.57 0.05 0.07 Nates:

(a) The surveillance weldment is a submeasured are weld fabricated using 3/16-inch diameter weld wire type B-4, heat number 83642, with a Linde 0091 type flux, lot number 3536. This weld wire / flux combination is identical to that used for the intermediate and lower shell vertical seams and the girth weld between the intermediate and lower shell plates.

(b) This table has been taken in its entirety from WCAP-14485.

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WCAP-15139 January 1999 o:\\4468. doc:lt> 012699 l

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4 Table 2 Interpolation of Chemistry Factors from Regulatory Guide 1.99, Revision 2, Position 1.1'"

Material Ni, wt %

Chemistry Factor,'F Intermediate Shell Plate B90041 0.40 44 0.55 44 Given Cu wt % = 0.065 0.60 44 Intermediate Shell Plate B9004-2 0.40 37 0.57-37 Given Cu wt % = 0.06 0.60 37 Lower Shell Plate B9005-1 0.40 51 0.58 51 Given Cu wt % = 0.08 0.60 51 Lower Shell Plate B9005-2 0.40 44 0.57 44 Given Cu wt % = 0.07 0.60 44 Weld Metal 0.00 26 0.07 34.1 Given Cu wt % = 0.05 0.20 49 Nets:

(a) This table has been taken in its entirety from WCAP-14485.

WCAP-15139 January 1999 o:\\448. doc 1b-011699

5 Table 3 Calculation of Chemistry Factors Using Surveillance Capsule Data Regulato y Guide 1.99, Revision 2, Position 2.1" Material Capsule Capsule f*

FF" ART FF* ART.

FF' Intermediate ShellPlate B9004-2 U

0.601 0.857 24.26 20.8 0.735 (Longitudinal)

V 2.64 1.26 55.93 70.5 1.59 Intermediate Shell Plate B9004-2 U

0.601 0.857 17.56 15.1 0.735 (Transverse)

V 2.64 1.26 46.27 58.3 1.59 Sum:

164.7 4.66 353"'

Weld Metal U

0.601 0.857 3.64 3.1 0.735 V

2.64 1.26 25.47 32.1 1.59 Sum:

35.2 2.32 15.2'*

j Naiss-(a) f = fluence (10" n/cm'); Fluence values were taken from Capsule V analysis (WCAP-14484)

(b) FF = fluence factor = f'"*

(c) ART. values obtained from CVGRAPH Version 4.0"

(d) CF = I(FF ' ART ) + I(FF')

(e) This table has been taken in its entirety from WCAP-14485.

Therefore, the calculated Chemistry Factor for the Intermediate Shell Plate B9004-2 based on surveillance capsule data is 35.3*F. The calculated Chemistry Factor for the Weld Metal based on surveillance capsule data is 15.2'F.

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WCAP-15139 January 1999 c:\\4468. doc 1t>012699 f

6 Table 4 Reactor Vessel Beltline Region Material Properties Used in Calculations" Method Used to Average Average Chemistry Initial RT "

Material Calculate CF" Cu wt %

Ni wt %

Factor (*F)

(*F)

Closure Head Flange"

0.74

-10 Vessel Flange"

0.73 0

Intennediate Shell Plate Position 1.1 0.065 0.55 44.0 60 B9004-1 Intermediate Shell Plate Position 1.1 0.06 0.57 37.0 40 B9004-2 Position 2.1 35 3 40 Lower Shell Plate B9005-1 Position 1.1 0.06 0.58 51.0 28 Lower Shell Plate B9005-2 Position 1.1 0.07 0.57 44.0 33 Weld Metal Position 1.1 0.05 0.07 34.1

-30 (Longitudinal &

Position 2.1 15.2

-30 CircumferentialSeams)

Notes:

(a) Regulatory Guide 1.99, Revision 2, Position.

(b) Initial RT. values of the base metal and weld metal materials are measured values.

(c) This table has been taken in its entirety from WCAP-14485.

WCAP-15139 January 1999 oA4468.docit>012699

7 4

CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE 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, Kg for the metal temperature at that time. K, is obtained from the reference fracture toughness curve, defined in Appendix G of the ASME Code,Section XI".

The K, curve is given by the following equation:

mo2n-w K, = 33.2 + 20.734

  • e (g) i
where, reference stress intensity factor as a function of the metal temperature T and f

K,

=

the metal reference nil-ductility temperature RT, Therefore, the governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows:

j C* K, + K,, < K, (2)

I

where, j

I stress intensity factor caused by membrane (pressure) stress K,

=

stress intensity factor caused by the thermal gradients K,,

=

function of temperature relative to the RT, of the material K,

=

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 C

=

not critical At any time during the heatup or cooldown transient, K, is determined by the metal temperature at the tip of a postulated flaw at the 1/4T and 3/eT location, the appropriate value for RT, 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, K,,, 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 cooldown, the controlling location of the flaw is always at the inside of the wall because the thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates. Allowable pressule-temperature relations are generated for both WCAP-15139 January 1999 c:\\4468. doc 1b-012699

8 steady-state and finite cooldown rate situations. From these relations, composite limit curves are constructed for each cooldown rate of intemst.

The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on the measurement of mactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the 1/4T vessel location is at a higher temperature than the fluid adjacent to the vessel inner diameter. This condition, of course, is not true for the steady-state situation. It follows that, at any given mactor coolant temperature, the AT (temperatum) developed during cooldown results in a higher value of K, at the 1/4T location for finite cooldown rates than for steady-state operation. Furthermore,if conditions exist so that the increase in K, exceeds K,,, the calculated allowable pressure during cooldown will be greater than the steady-state value.

The above procedures are needed because there is no direct control on temperatum at the 1/4T location and, therefore, allowable pressums 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 ensures conservative operation of the system for the entire cooldown 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 conditions assunung the pmsence of a 1/4T defect at the inside of the wall. The heatup results in compressive stresses at the inside surface that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the K, for the 1/4T crack during heatup is lower than the K, for the 1/4T crack during steady-state conditions at the same coolant temperatum. Dunng heatup, especially at the end of the transient, conditions may exist so that the effects of compressive thermal stresses and lower K, values 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 for finite heatup rates when the 1/4T flaw is considered. Therefore, both cases have to be analyzed in order to ensum that at 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 concerns the calculation of the pressure-temperature limitations for the case in which a 1/4T flaw located at the 1/4T location from the outside surface 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 minforce any pressure stresses present. These thermal stresses are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup i

ramp. Since the thermal stresses at 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 produced by constructmg a composite curve based on a point-by-point comparison of the steady-state and finite heatup rate data. At any I

i WCAP-15139 -

January 1999 oA4468.docib-012699

9 given temperature, the allowable pressure is taken to be the lescer 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.

10 CFR Part 50, Appendix G" addresses the metal temperatum 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 pressum exceeds 20 percent of the preservice hydrostatic test pressure (3106 psi), which is 621 psig for Beaver Valley Unit 2.

Table 4 indicates that the limiting unirradiated RT, of 0 F occurs in the vessel flange of the Beaver Valley Unit 2 reactor vessel, so the minimum allowable temperature of this region is 130*F at pressures greater than 561 psi when instrumentation error margms of 10*F and 60 psig are included and 120*F at pressures greater than 621 psi when instrumentation error margins are not included.

WCAP-15139 January 1999 o \\4M8. doc 1b-012699

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'5 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE From Regulatory Guide 1.99, Revision 2, the adjusted reference temperature (ART) for each material in the beltline region is given by the following expression:

ART = Initial RT, + ART, + Margin (3) l Initial RT,is the reference temperature for the unirradiated material as defmed in paragraph i

NB-2331 of Section III of the ASME Boiler and Pressure Vessel Code"". If measured values of initial RT, for the 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 l

deviation for the class.

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

ART, = CF

  • f"""* *

(4) 1

\\

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

f,,, = f,,,,,,,' e '***"'

(5) where x inches (vessel beltline thickness is 7.875 inches) is the depth into the vessel wall measured from the vessel clad / base metal interface. The resultant fluence is then placed in Equation 4 to calculate the ART, at the specific depth. The calculated peak surface fluence for l

the Beaver Valley Unit 2 base metal and circumferential weld (at the 0* azimuthal angle of the reactor vessel) at 15 EFPY is 1.81 x 10" n/cm'. For the longitudinal welds, the surface fluence (at the 45* azimuthal angle of the reactor vessel) at 15 EFPY is 5.66 x 10" n/cm'.

The chemistry factor values obtained from Tables 1 and 2 of Regula' ory Guide 1.99, Revision 2, t

were determined in Table 2 using the copper and nickel content values reyvikd in Table 1.

Chemistry factors were also calculated using credible surveillance capsule data as shown in Table 3.

Margin is calculated as, M = 2 fo' + o*. The standard deviation for the initial RT, margin i

term, is e,0*F when the ir.itial RT, is a measured value, and 17'F when a generic value is

' available. The standard deviation for the ART, margin term, o, is 17'F for plates or forgings, l

and 8.5'F for plates or fr rgings (half the value) when surveillance data is used. For welds, e,is equal to 28'F when sur 'eillance capsule data is not used, and is 14'F (half the value) when T

credible surveillance ct.psule data is used. o, need not exceed 0.5 times the mean value of 2

ART,. See Table 5.

WCAP-15139 January 1999 c:446s. doc.lb-012699 u-

11 Table T, Margins for Adjusted Reference Temperature (ART) Calculations per Regulatory Guide 1.99, Revision 2*

Material Properties Sury. Capsule Data NOT Used Surv. Capsule Data Used Plates or Forgings Measured IRT, 34 17 Generic 1RT.

48 38 Weld Metal Measured IRT, 56 28 Generic IRT, 66 44 (a) This table has been taken in its entirety from WCAP-14485.

i All materials in the beltline region of Beaver Valley Unit 2 reactor vessel were considered in i

determining the limiting material. Sample calculations to determine the ART values for the Intermediate Shell Mate B9004-1 are shown in Table 6. The resulting ART values for all beltline materials at the 1/4T and 3/4T locations are summarized in Table 7. From this table, it can be seen that the limiting material to be used in the generation of the heatup and cooldown curves is Intermediate Shell Plate B9004-1.

(Note: When two or more credible surveillance data sets become available, the data sets may be used to determine ART values as described in Regulatory Guide 1.99, Revision 2, Position 2.1. If the ART values based on surveillance capsule data are larger than those calculated per Regulatory Guide 1.99, Revision 2, Position 1.1, the surveillance data should be used. If the surveillance capsule data gives lower values, either may be used.)

WCAP-15139 January 1999 o:\\4e8. doc 1b-012699

12 Table 6 Calculation of ART Values for the Limiting Beaver Valley Unit 2

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Reactor Vessel Material-Intermediate Shell Plate B9004-1" Parameter Operating Time 15 EFPY Location 1/4T ART 3/4T ART Chemistry Factor, CF (*F) 44.0 44.0 Fluence, f (10" n/cm')'"

1.13 0.439 Fluence Factor, FF 1.03 0.771 ART. = CF x ff ('F) 45.5 33.9 Initial RT., I ('F) 60 60 Margin, M (*F) 34 33.9 Adjusted Reference Temperature (ART),

140 128

('F) per Reg. Guide 1.99, Revision 2 Notes.

(a) Fluence, f,is based upon f,,,,(10" n/cm', E>l.0 MeV) = 1.81 at 15 EFPY.

(b) The Beaver Wiley Unit 2 reactor vessel wall thickness is 7.875 inches at the beltline region.

(c) This table has been taken in its stirety from WCAP-14485.

WCAP-15139 January 1999 o:\\44MLdoc:1b-012699 i

13

)

Table 7 Summary of ART Values at the 1/4T and 3/4T Locations" 15 EFPY ART Method Used to Calculate the CF(Regulatory 1/4T ART 3/4T ART Material Guide 1.99, Revision 2)

(*F)

('F)

Intermediate ShellPlate B9004-1 Position 1.1 140 128 Intermediate ShellPlate B9004-2 Position 1.1 112 97 Positio12.1 94 84 Lower ShellPlate B9005-1 Position 1.1 115 101 Lower Shell Plate B9005-2 Position 1.1 112 101 Longitudinal Welds Position 1.1 19 3

(located at the 45' azimuthalangle) i Position 2.1

-8

-15 Circumferential Weld Position 1.1 41 23 Position 2.1 1

7 Nats:

(a) This table has been taken in its entirety from WCAP 14485.

i l

WCAP-15139 January 1999 0:\\4468. doc 1b-012699

14 6

HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES Pmssure-temperature limit curves for normal heatup and cooldown of the primary mactor coolant system have been calculated for the pressure and temperature in the reactor vessel beltline region using the methods" discussed in Sections 3.0 and 4.0 of this report. The pmssure difference between the wide-range pressure transmitter and the limiting beltline region has not been accounted for in the pressure-temperature limit curves generated for normal operation.

Figum 1 presents the heatup curve with margins for possible instrumentation errors using heatup rates up to 60*F/hr applicable for the first 15 EFPY, Figure 2 present the cooldown curves with margins for possible instrumentation errors using cooldown rates up to 100*F/hr applicable for 15 EFPY. Additionally, Figures 3 and 4 present the heatup and cooldown curves without margins for possible instrumentation errors. Allowable combinations of temperature j

and pmssure for specific temperature change rates are below and to the right of the limit lines

- shown in Figums 1 through 4. This is in addition to other criteria which must be met before the reactor is made critical, as discussed below in the following paragraphs.

The reactor must not be made critical until pressure-temperature combinations are to the right of the criticality limit line shown in Figures 1 and 3. The straight-line portion of the criticality limit is at the minimum permissible temperatum for the 2485 psig inservice hydmstatic test as i

required by Appendix G to 10 CFR Part 50. The governing equation for the hydrostatic test is

)

defined in Code Case N-626'"1(Note: To be implemented in 1999 Addenda to Appendix G to Section XI of the ASME Code) as follows:

1.5 K, <K,

where, K, is the stress intensity factor covered by membrane (pressure) stress, K, = 33.2 + 20.734 e"~"*d, T is the minimum permissible metal temperature, and

. RT,is the metal reference nil-ductility temperature.

2 The criticality limit curve specifies pressure-temperature limits for core operation to pmvide additional margin during actual power production as specified in Reference 16. The pressure-l temperatum limits for core operation (except for low power physics tests) are that the reactor vessel must be at a temperature equal to or higher than the mmimum temperature required for the inservice hydrostatic test, and at least 40*F higher than the mirumum permissible i

temperature in the u,.ispcs. ding pressure-temperatum curve for heatup and cooldown calculated as described in Section 3.0 of this report. For the heatup and cooldown curves with i

margins for instrumentation errors, the mmimum temperature for the in service hydrostatic leak tests for the Beaver Valley Unit 2 reactor vessel at 15 EFPY is 208'F. Without incorporation WCAP-15139 January 1999 a:u46s. doc 1b-012H9

15 of instrumentation error margins, the minimum temperature for the inservice hydrostatic leak tests for the Beaver Valley Unit 2 reactor vessel at 15 EFPY is 196*F. The vertical line drawn

' from these points on the pressum-temperature curve, intersecting a curve 40*F higher than the pressum-temperature limit curve, constitutes the limit for core operation for the reactor vessel.

Figures 1 through 4 define all of the above limits for ensuring prevention of nonductile failure for the Beaver Valley Unit 2 reactor vessel.

The data points used for the heatup and cooldown pressure-temperature limit curves shown in Figures 1 through 4 are presented in Tables 8 and 9. As seen by comparing these results to that fmm Tables 8 and 9 of WCAP-14485, there is a minimum increase in pressure of 165 psig

- (@ lowest temperatum and highest cooldown rate) when K, is used in the calculation of heatup and cooldown limit curves.

I WCAP-15139 January 1999 c:\\4468. doc:1b-012699

7

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MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL PLATE B9004-1 l

LIMITING ART VALUES AT15 EFPY:

1/4T,140*F l

3/4T,128'F 2500

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Figure 1 Beaver Valley Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rates up 60*F/hr) Applicable for the First 15 EFPY (With Margins of 10*F and 60 psig for Instrumentation Errors)

WCAP-15139 January 1999 o:\\4468. doc 1b-012699 l

17 MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL PLATE B9004-1 i

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Figure 2 Beaver Valley Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100*F/hr) Applicable for the First 15 EFPY (With Margins of 10*F and 60 psig for Instrumentation Errors)

Includes vessel flange requirements of 130'F and 561 pelg per 10CFPJ50, Appendix G WCAP-15139 January 1999 c:\\4468. doc 1b-012699

18 Table 8 15 EFPY Heatup and Cooldown Curve Data Points (With Instrumentation Error Margins of 10*F and 60 psig)

Cooldown Curves Heatup Curve Leak Test Steady Criticality State 20F/hr 40F/hr 60F/hr 100F/hr 60F/hr Limit Limit T

P T

P T

P T

P T

P T

P T

P T

P 70 561 70 561 70 561 70 547 9 70 471.77 70 561 208 0

191 2000 75 561 75 561 75 561 75 558.15 75 483 64 75 561 208 668.58 208 2485 80 561 80 561 80 561 80 561 80 496 89 95 561 208 693.1 85 561 85 561 85 M1 85 561 85 5i1.67 100 561 206 687.74 90 M1 90 561 90 561 90 561 90 528 13 105 561 208 686 %

95 561 95 561 95 561 95 561 95 546 46 110 El 208 689 93 100 561 100 561 100 561 100 El 100 561 115 561 208 6 % 41 105 561 105 561 105 561 105 561 105 561 120 561 208 705 98 110 561 110

%1 110 561 110 561 110 561 125 561 208 71857 115 561 115 561 115 Mt 115 561 115 561 130 561 208 734 02 120 561 120 561 120 561 120 561 120 El 130 734 02 208 752 4 125 561 I?5 561 125 561 125 561 125 561 135 752 4 208 773.69 130 561 130 Mt 130 561 130 561 130 561 140 773.69 208 798.06 130 847.2 130 823 94 130 802 130 781.62 130 746.75 145 798.06 208 825.62 135 875.36 135 854.39 135 834 99 135 817.44 135 789 3 150 825 62 208 856 64 140 906.47 140 888 05 140 8715 140 857 13 140 8365 155 856.64 208 89132 145 940.87 145 925.31 145 911.92 145 901.1 145 888.86 160 891.32 208 930 150 978.88 150 966.49 150 956.66 150 949.79 150 946.92 165 930 210 973 155 1020 88 155 1012.06 155 1006.18 155 1003.72 155 1011 29 170 973 215 1020.74 160 1067.31 160 1062 43 160 1060.97 160 1063.42 175 1020.74 220 1073.0 165 1118.62 165 1118.17 165 180

'1073 65 225 113224 170 1175 32 185 113224 230 1197.06 175 1237.99 190 1197.06 235 1268.73 180 1307.25 195 1268.73 240 1347.95 185 1383.79 200 1347.95 245 1435 48 190 1468 38 205 1435 48 250 1532.16 195 1561.87 210 1532.16 255 1638 92 200 1665 19 215 1638.92 260 1756.81 205 1779 38 220 1756.81 265 1886 95 210 190558 225 1886.95 270 2030 62 215 2045 05 230 20% 62 275 2189 19 220 2199 18 235 2189 19 280 2364.21 225 2369 53 240 2364 21 WCAP-15139 January 1999 o:\\4468. doc 1b-022599

19 MATERIAL PROPERTY BASIS 1

LIMITING MATERIAL: INTERMEDIATE SHELL PLATE B9004-1 LIMITING ART VALUES AT 15 EFPY:

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Moderator Temperature (Deg.F)

Figure 3 Beaver Valley Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rates up to 60*F/hr) Applicable for the First 15 EFPY (Without Margins for Instrumentation Errors)

WCAP-15139 January 1999 c:\\4468. doc 1b 012699

20 MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL PLATE B9004-1 LIMITING ART VALUES AT 15 EFPY:

1/4T,140'F 3/4T,128'F 2500 1

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500 - ll "a m '8 250 g Temp. 0 0 5'O 100 150 200 250 300 350 400 450 500 Moderator Temperature (Deg.F) Figure 4 Beaver Valley Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100'F/hr) Applicable for the First 15 EFPY (Without Margins for Instrumentation Errors) includes vessel flange requirements of 120*F and 621 psig per 10CFR50, Appendix G WCAP-15139 I""""U I o:\\4468. doc 1bO12699

21 Table 9 15 EFPY Heatup and Cooldown Curve Data Points (Without Instrumentation Error Margins) Cooldown Curves Heatup Curve Leak Test Steady Criticality State 20F/hr 40F/hr 60F/hr 100F/hr 60F/hr Limit Limit T P T P T P T P T P T P T P T P 60 621 60 621 60 621 60 607,9 60 53177 60 621 1% 0 178 2000 65 621 65 621 65 621 65 618.15 65 54344 65 621 196 66838 1% 2485 70 621 70 621 70 621 70 621 70 556.89 85 621 196 693.1 75 621 75 621 75 621 75 621 75 57147 90 621 1% 68774 80 621 80 621 80 621 80 621 80 548.13 95 621 1% 686 % 85 621 85 621 85 621 85 621 85 60646 100 621 196 689.93 90 621 90 621 90 621 90 621 90 621 105 621 1% 696 41 95 621 95 621 95 62) 95 621 95 621 110 621 196 705.98 100 621 100 622 100 621 100 621 100 621 115 621 196 71837 105 621 105 621 105 621 105 621 105 621 120 621 196 734 02 110 621 110 621 110 621 110 621 110 621 120 794.02 1% 752.4 115 621 115 621 115 621 115 621 115 621 125 812.4 1% 77349 120 621 120 621 120 621 120 621 120 621 130 83349 1% 798M 120 9072 120 883 94 120 862 120 84142 120 806.75 135 858 06 1% 825 62 125 93536 125 91439 125 894.99 125 87744 125 849 3 140 885.62 196 85644 130 96647 130 948 05 130 9313 130 917.13 130 8963 145 91644 1% 89132 135 1000.87 135 985J1 135 971.92 135 %1.1 135 948.86 150 951.32 196 930 140 1038.88 140 1026.49 140 101646 140 1009 19 140 1006.92 155 990 200 973 145 1080.88 145 1072.06 145 1066.18 145 1063.72 145 1071.29 160 1033 205 1020J4 150 1127J1 150 1122.43 150 1120.97 150 1123.42 165 100014 210 1073.65 155 1178.62 155 1178.17 170 1133.65 215 1132.24 160 123532 175 119224 220 11E06 165 1297.99 100 1257.06 225 126813 170 1367.25 185 1328 73 230 1347.95 175 1443.79 190 1407.95 235 143548 180 1528.38 195 1495.48 240 1532 16 185 1621J7 200 1592.16 245 1638.92 190 1725.19 205 1698.92 250 175621 195 1839.38 210 1816.81 255 1886.95 200 1965 38 215 1946.95 260 2030.62 205 2105.05 220 2040 42 265 2189.19 210 2259.18 225 2249.19 270 236421 215 2429 33 230 2424.21 1 WCAP-15139 c:\\4468. doc:1bO22599

l 22 l 7 REFERENCES l 1. Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials," U.S. Nuclear Regulatory Commission, May 1988. 2. " Fracture Toughness Requirements," Branch Technical Position MThB 5-2, Chapter 53.2 in Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power j Plants, LWR Edition, NUREG-0800,1981. 3. Combustion Engineering, Inc., Materials Certification Report, Contract No. 9071, Code No. B-9004-1, J. M. Arnold,7/31/72. 4. Combustion Engmeering, Inc., Materials Certification Report, Contract No. 9071, Code No. B-9004-2, J. M. Arnold,11/17/72. 5. Combustion Engineering, Inc., Materials Certification Report, Contract No. 9071, Code No. B-9005-1, J. M. Arnold,12/4/72. l 6. Combustion Engmeering, Inc., Materials Certification Report, Contract No. 9071, Code No. B-9005-2, J. M. Arnold,12/4/72. i 7. CE Power Systems, Westinghouse Contract No. 9071, " Intermediate and Lower Shells Analyzed to ASTM E-350 Sample Obtained at 1/4 Thickness," W. A. House,8/19/76. 8. Combustion Engineering Check Analysis, Contract 9071, Plate B-9004-2. 9. Metallurgical Research and Development, Chemical Analysis of Wire-flux Test Weld Coupon, Heat No. 83642, Flux 0091, Lot No. 3536, dated 10/12/72. 10. CE Power Systems, Westinghouse Contract No. 9071, Weld Chemical Analysis, W. A. House,8/19/76. 11. BAW-1880, " Analysis of Capsule W-83, Florida Power and Light Company St. Lucie Plant Unit No. 2, Reactor Vessel Material Surveillance Program," A. L Lower, Jr., et al., September 1985. l WCAP-14484, " Analysis of Capsw U from the Duquesne Light Company Beaver Valley 12. Unit 2 Reactor Vessel Radiation Surveillance Program," P. A. Grendys, et al., draft dated November 1995. 13. CVGRAPH, Hyperbolic Tangent Curve-Fitting Program, Version 4.0, developed by ATI Consulting, March 1995. WCAP-15139 January 1999 o:\\4468. doc 1b-012699 l

F-l 23 l 14. WCAP-12406, " Analysis of Capsule U from the Duquesne Light Company Beaver Valley l Unit 2 Reactor Vessel Radiation Surveillance Program," S. E. Yanichko, et al., L September 1989. 15. Section XI of the ASME Boiler and Pressum Vessel Code, Appendix G, " Fracture Toughness Criteria for Protection Against Failure." .16. Code of Federal Regulations,10 CFR Part 50, Appendix G, " Fracture Toughness Requirements," U.S. Nuclear Regulatory Commission, Washington, D.C., Federal Register, Volume 60, No. 243, dated December 19,1995. i 17. 1989 Section III, Division 1 of the ASME Boiler and Pressure Vessel Code, Paragraph NB-2331, " Material for Vessels." i l 18. WCAP-7924-A, " Basis for Heatup and Cooldown Limit Curves," W. S. Hazelton, et al., ) April 1975. 19. WCAP-14485, " Beaver Valley Unit 2 Heatup and Cooldown Limit Curves for Normal Operation," P. A. Grendys, March 1986. 20. Draft Code Case No. 626, " Alternative Reference Fracture Toughness for Development of P-T Limit Curves for Section XI, Division 1," to be published late 1998. 1 l l t I l WCAP-15139 January 1999 o:\\4468. doc 1b-012699

ATTACHMENT D Beaver Valley Power Station, Unit No. 2 License Amendment Request No. 127 UPDATE HEATUP AND COOLDOWN CURVES AND OVERPRESSURE PROTECTION SYSTEM (OPPS) SETPOINT CURVE A Proprietary Information Notice copyright Notice Application for Withholding Proprietary Information from Public Disclosure Low Temperature Overpressure Protection System Setpoint Class 2C Review for Beaver Valley Unit 2 15 EFPY Heatup and Cooldown Curves Low Temperature Overpressure Protection System Setpoint Class 3 Review for Beaver Valley Unit 2 15 EFPY Heatup and Cooldown Curves [

WESTINGHOUSE PROPRIETARY CLASS 3 FINAL REPORT Low Temperature Overpressure Protection System Setpoint Review for Beaver Valley Unit 215 EFPY Heatup and Cooldown Curves d i l-

WESTINGHOUSE PROPRIETARY CLASS 3 FINAL REPORT 1. INTRODUCTION Westinghouse has developed revised Low Temperature Overpressure Protection System (LTOPS) PORV setpoints for Beaver Valley Unit 2, applicable to 15 EFPY heatup and cooldown curves. The setpoints conservatively :ncorporate instrument uncertainties and the delta-P between the wide range pressure transmitter and the reactor vessel limiting beltline region. The methodology and resulting setpoints are discussed below.- 11. BACKGROUND Deaver Valley Unit 2 Technical Specifications, in combination with plant administrative controls, currently require that LTOPS be operable below an RCS temperature of 350F. The applicable design basis transients are as fc!!ows: The first transient is a heat addition scenario in which a reactor coolant pump in a single loop is ) . started when the RCS temperature is as much as SOF lower than the steam generator secondary side temperature. This results in a sudden secondary to primary heat transfer and rapid increase in primary system pressure. The second design basis transient is a mass injection event caused by the failure of the controls for a single charging pump to the full flow condition. The influx of fluid into the relatively inelastic RCS also causes a rapid increase in system pressure. 111. METHODOLOGY / MAJOR INPUT ASSUMPTIONS Major input assumptions are outlined below.

1. Overshoots during the mass injection event and heat addition event from the current analyses of record (Reference 5) will be used. No reanalysis of the design basis mass l

injection or heat addition scenarios will be pedormed. For Beaver Valley Unit 2, analyses of the mass injection and heat addition events have been performed assuming that the RCS is water solid and that the RHRS is isolated (conservative). The RCS temperature assumed for the mass injection event is 70F. The heat addition event has been analyzed for RCS temperatures between 70F and 300F.

2. Pressure overshoots during the design basis events are based on a pressurizer PORV stroke l

open/close time, of 1.65/1.0 seconds.

3. Since the heatup and cooldown curves are generated using the K1C fracture toughness methodology, ASME Code Case N-514 (which permits a 10% relaxation of the Appendix G pressure temperature limits up to the LTOPS enable tamperature) is not applicable.
4. Setpoints are selected so that RCS pressures will not exceed the 15 EFPY Appendix G pressure limits down to the reactor vessel temperature of 60F.
5. In order to preserve the single failure criteria, the overshoots are calculated assuming the availability of one PORV during the design basis mass injection and heat addition events, Page 1

/:. WESTINGHOUSE PROPRIETARY CLASS 3 FINAL REPORT when the RCS is water solid, concurrent with loss of letdown and isolation of the RHRS. The second PORV is assumed to have failed.

6. The maximum allowable setpoints will be derived from steady state heatup and cooldown curves at 15 EFPY. See Section IV-B for more discussion on this topic.
7. The setpoints are appiicable to 30% steam generator tube plugging.
8. The setpoints conservatively account for instrument uncertainties associated with the wide range pressure transmitter (

)** and wide range temperature [ ]**,per Reference 6.

9. The setpoints conservatively account for the pressure difference between the wide-range pressure transmitter and the reactor vessel limiting beltline region, identified in Nuclear Safety Advisory letter NSAL-93-005A.
10. Heat transport effects, which are applied to the heat injection transient results and account for a 50F difference between the wide range temperature sensor and the reactor vessel, have also been incorporated.
11. Nine breakpoints are selected as input into the function generator so that the Appendix G limits are not exceeded during the limiting design basis mass injection or heat injection events. Breakpoints are selected such that the gain in each line segment does not exceed [

)a.b.c

12. A qualitative assessment has been performed with regard to PORV undershoots (margin to the RCP number one seat limit) during water solid operation.

IV. SETPOINT DEVELOPMENT A. Limiting Overshoots for the Mass injection and Heat Addition Events j The limiting peak RCS pressures versus PORV setpoints for the mass injection event are tabulated in Table A, assuming mass injection into a water solid RCS from one charging pump with no RHR relief path available. The analysis of record (Reference 5) modeled the following maximum flows from one charging pump as a function of pressure as follows: l RCS Pressure Mass injection Rate (psig) (gpm) - ~ 350 a,b c 4 400 450 500 550 650 750 Page 2

WESTINGHOUSE PROPRIETARY CLASS 3 FINAL REPORT The limiting peak RCS pressures versus PORV setpoints for the heat addition event, also from j the analysis of record, are tabulated in Table B. B. 15 EFPYf.ppendix G Limits I Reference 3 provides data for the Beaver Valley Unit 2 steady state heatup and cooldown curves at 15 EFPY. Reactor vessel temperatures and corresponding pressure limits are shown respectively in Columns 1 and 2 of Table C. Since LTOPS events are most likely to occur when the reactor vessel is at isothermal conditions, the steady state heatup and cooldown limits have formed the basis for LTOPS setpoint selection for more than a decade and are considered applicable for the current LTOPS setpoint development at 15 EFPY for Beaver Valley Unit 2. C. Adjustments for Pressure Uncenalntles ,,s,e ~ D. Adjustments for Delta-P ~ a,b,c l 1 M Page 3

WESTINGHOUSE PROPRIETARY CLASS 3 FINAL REPORT j E. Temperature Uncertainties and Heat Transport Effects Temperature uncertainty and streaming effects on RCS cold leg wide range temperature indication have been calculated by Westinghouse to be [ ]***. Heat transport effects, which account for reactor vessel being [ J'** colder than the temperature at the wide range temperature sensor are also incorporated into the maximum allowable setpoint development for those setpoints applicable to the heat injection event. F. PORVMaximum Allowable Setpoints a,b c lt should be noted that current Beaver Valley Unit 2 administrative controls restrict the number of RCPs operable when LTOPS is enabled, as follows: Tnes s 100F 0 RCPs running 100F > Tncs s 160F 1 RCP running 160F > Tnes s 190F 2 RCPs running Tnes > 190F 3 RCPs running Page 4 t-

WESTINGHOUSE PROPRIETARY CLASS 3 FINAL REPORT Due to setpoint limitations as a result of the reactor vessel flange requirements, there is no operational benefit achieved by restricting the number of reactor coolant pumps running, to less than two pumps, below an indicated RCS temperature of 190F. Therefore, the PORV setpoints shown in Table D will protect the Appendix G limits for the following combinations of RCPs: i Tncs < 190F 0 - 2 RCPs running Tnes => 190F 3 RCPs running Finally, the recommended LTOPS arming temperature, that will protect both the Appendix G limits and the PORV piping, is 350F. The administrative arming temperature at Beaver Valley Unit 2 is 367F, which is conservative. G. Margin to the Reactor Coolant Pump Number One Seal Limit The upper pressure limit for LTOPS is defined by the Appendix G limit, after consideration of all uncertainties and the delta-P between the wide range pressure transmitter and reactor vessel limiting region. The low limit on pressure during the design basis LTOPS mass injection and heat injection transients is established based on operational consideration for the RCP number one seal which requires a nominal differential pressure across the seal faces for proper film-riding performance. As part of the LTOPS setpoint evaluation, margin to the reactor coolant pump number one seal limit is evaluated. Based on the analyses of record (Reference 5), this limit corresponds to a differential pressure I across the seal of 200psid, which corresponds to the gage pressures shown in Table E. As ) demonstrated in Table E, pressure undershoot below the PORV setpoint dtiring a design basis mass injection or heat injection event can exceed 100 psi. Therefore, with the PORV setpoints developed for the 15 EFPY heatup and cooldown curves, there is the potential for RCS pressure to violate the RCP number one seal limit at the lowest RCS temperatures. While analysis has not been performed that models the simultaneous relief from two PORVs, l undershoots below the PORV setpoint can be significantly higher if both PORVs actuate during an LTOPS event, and it is anticipated that the pump seallimit would be exceeded. However, staggering the PORV setpoints minimizes the likelihood that both PORVs will actuate simultaneously during credible LTOPS events. In addition, WCAP-14040-NP-A (Reference 7) indicates that when there is insufficient range between the upper and lower pressure limits to select PORV setpoints that provide protection against violating both limits, then the setpoint selection that provides protection against the upper limit violation takes precedence. Similarly, since there is insufficient margin at lower temperatures to accommodate independent PORV and RHR relief valve operation, simultaneous PORV and RHR relief valve actuation is possible. Therefore, there is also the concern that the pump seal could be violated if a PORV and RHR relief valve open simultaneously when the RHR is not isolated from the RHRS. H. Peak RCS Pressure versus RCS Temperature A tabulation of peak RCS pressure versus RCS temperature is provided in Table F. The peak i pressures are the maximum RCS pressures that would be expected to occur during either of the design basis mass injection or heat injection events, after consideration for all uncertainties Page 5

l WESTINGHOUSE PROPRIETARY CLASS 3 FINAL REPORT (regardless of whether measured temperature is higher or lower than RCS temperature) and the pressure difference between the wide range pressure transmitter and the reactor vessel limiting beltline region. i 1 V. CONCLUSIONS Maximum allowable LTOPS PORV setpoints have been developed based on the 15 EFPY l heatup and cooldown curves for Beaver Valley Unit 2. The setpoints conservatively incorporate applicable instrument uncertainties and the delta-P between the wide range pressure transmitter and the reactor vessel limiting beltline region. Furthermore, since the 15 EFPY heatup and cooldown curves are based on the K1C fracture toughness methodology, ASME Code Case N-514 has not been incorporated. With the restrictions previously identified on the number of RCPs running, the setpoints shown in Table D and Figure 1 will prevent RCS pressure from exceeding the 15 EFPY Appendix G limits after consideration of all instrument uncertainties and the delta-P between the wide range pressure transmitter and the reactor vessellimiting beltline region. Therefore, during any design basis mass injection or heat injection event, peak RCS pressure can conservatively be considered to be less than or equal to the applicable Appendix G limit over the full range of temperatures when LTOPS is enabled. l i Page6

WESTINGHOUSE PROPRIETARY CLASS 3 FINAL REPORT VI. REFERENCES a,b,c

3. WCAP 15139 " Beaver Valley Unit 2 Heatup and Cooldown Curves for Normal Operation at 15 EFPY using Code Case N-626", T. Laubham, January 1999.
4. Westinghouse Owners Group letter OG-95-54, L Bush,6-16-95.

~

7. WCAP 14040-NP-A, Rev. 2,"Methodologf Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves"J.D. Andracheck et al, January 1996.
8. Beaver Valley Technical Specifications Section 3/4.4.9.

Page 7

WESTINGHOUSE PROPRIETARY CLASS 3 FINAL REPORT TABLE A LIMITING PEAK PRESSURES FOR MASS INJECTION EVENTS (Mass injection from 1 Charging Pump) Temp. = 60F Set Peak j Pressure Pressure a,b,e (psia) (psig) 400 j 450 l 500 600 700 Notes: Overshoots for the mass injection event have conservatively been adjusted to account for steam Generator tube plugging levels up to 30% Page8

WESTINGHOUSE PROPRIETARY CLASS 3 FINAL REPORT TABLE B LIMITING PEAK PRESSURES FOR HEAT ADDITION EVENTS Temp.= 70F Set Peak Pressure Pressure (psig) (psig) abc 400 450 l 500 l 600 i 700 Temp.= 120F " Set Peak Pressure Pressure (psig) (psig) a,b,c 400 450 500 600 700 Temp.= 120.1F " i j Set Peak Pressure Pressure (psig) (psig) a,b,c 400 450 500 600 700 Notr *: The 15 EFPY heatup and cooldown curves are limited by the vessel flange ut arature of 120F. The overshoot resulting from the heat injection event at 'a 's a.servatively taken to be equal to the overshoot analyzed at 150F, However, int - ss injection transient is more limiting at this temperature, there is no adverse i irr e calculated setpoints. Page 9

9 WESTINGHOUSE PROPRIETARY CLASS 3 FINAL REPORT TABLE B, CONT'D LIMITING PEAK PRESSURES FOR HEAT ADDITION EVENTS Temp.= 200F Set Peak Pressure Pressure (psig) (psig) a,b,c 400 450 500 600 700 Temp.= 250F Set Peak Pressure Pressure (psig) (psig) a,b.c 400 450 500 600 700 Temp.= 300F Set - Peak Pressure Pressure i (psig) (psig) a,b c 400 450 500 600 700 Page 10

WESTINGHOUSE PROPRIETARY CLASS 3 FINAL REPORT TABLE C: 15 EFPY STEADY STATE APPENDIX G LIMITS RCS Pressure Limit less Limit with Limit with Limit with 2 Limit with 3 Temperature Limit Uncertainty of 0RCPS 1 RCP RCPS RCPS (degree F) (psig) 99 psi Running Running Running Running (psig) (psig) (psig) (psig) (4) (5) (6) (7) (1) (2) (3) a,b c 60 621 522 65 621 522 70 621 522 75 621 522 80 621 522 85 621 522 90 621 522 95 621 522 100 621 522 105 621 522 110 621 522 115 621 522 120 621 522 120.1 907 808 125 935 836 130 966 867 135' 1001 902 140 1039 940 145 10814 982 150 1127 1028 155 1179 1080 160 1235 1136 165 1298 1199 170 -1367 1268 175 1444 1345 180 1528 1429 185 1622 1523 190 1725 1626 195 1839 1740 200 1966 1867 205 2105 2006 21_0 2259 2160 215 2430 2331 250 2430 2331 300 2430 2331 Note: Maximum Appendix G limit reported at 215F. Appendix G limit taken to be constant at temperatures greater than 215F. Page 11

4 WESTINGHOUSE PROPRIETARY CLASS 3 FINAL REPORT TABLE D: MAXIMUM ALLOWABLE PORV SETPOINT@ 4 Temperature Setpoint PORV #2 Setpoint PORV #1 (degree F) (psig) (psig) 60 455 418 137 455 418 167 465 428 j 177 470 432 190 485 441 200 510 460 210 624 547 367 624 547 425 2335 2335 I (1) The PORV setpoints shown in Table D will protect the Appendix G litnits for the following combinations of RCPs: Tncs < 190F 0 - 2 RCPs running Tees => 190F 3 RCPs running Page 12

WESTINGHOUSE PROPRIETARY CLASS 3 FINAL REPORT TABLE E MARGIN TO THE REACTOR COCLANT PUMP NUMBER ONE SEAL LIMIT a,b,c a,b,c a,b,c RCS Temperature PORV #1 Setpoint (degree F) (psig) 60 418 150 421 200 460 250 547 350 547 Page 13

WESTINGHOUSE PROPRIETARY CLASS 3 FINAL REPORT Figure 1: Beaver Valley Unit 2 Maximum Allowable Setpoints at 15 EFPY Maxim um Allowa ble Nominal PORV Setpoints for the Overpressure Protection System 650 600 g 550 1 8 ~ j < 500 l E ~ ( I / / 450 g a 400 0 100 200 300 400 Auctioneered Low M easured RCS Tem perature (degree F) l l l Page 14 j l i

i WESTINGHOUSE PROPRIETARY CLASS 3 FINAL REPORT TABLE F: MAXIMUM EXPECTED PEAK PRESSURE VERSUS RCS TEMPERATURE RCS Peak Appendix G Temperature Pressure Limit (degree F) (psig) (psig) 60 621 621 120 621 621 150 752 800 200 765 800 250 781 800 300 800 800 350 800 800 1 ) 1 Page 15

ATTACHMENT E Beaver Valley Power Station, Unit No. 2 License Amendment Request No. 127 UPDATE HEATUP AND COOLDOWN CURVES AND OVERPRESSURE PROTECTION SYSTEM (OPPS) SETPOINT CURVE Exemption Request l

l Exemption Request Reauirement for Which Exemotion is Reauested Pursuant ~to 10 CFR 50.12, " Specific Exemptions," enclosed is a request for exemption from. certain requirements of 10 CFR 50.60, " Acceptance' criteria'for fracture prevention measures ror lightwater nuclear power reactors-for normal operation," and 10 CFR 50, Appendix G, " Fracture ~ Toughness Requirements." This exemption is requested to ' allow the application of ASME Code Case N-640 in determining the a' ceptable overpressure protection (OPPS) power c operated relief valve (PORV) setpoints. ASME'Section XI Code Reauirements ASME -Section ^XI, ~ Appendix G,- " Fracture Toughness Criteria for Protection' Against Failure," Article A-2000, provides the service limits : for pressure vessels, and establishes the allowable vessel loading- (internal pressure, external load, thermal stress) versus temperature..The Code requirements are to maintain vessel operating conditions within Article A-2000 requirements. Code Reauirement from Which Exemotion is Recuested Exemption is requested from 10 CFR 50, Appendix G, and ASME Section XI, Appendix G, requirements for reactor vessel pressure limits at low temperatures. Basis for Exemotion Reauest iCurrent OPPS setpoints _ produce operational constraints by limiting j the pressure-temperature (P/T) range available to the operator to ' heat up or cool down the plant. The " operating window" through which the operator 'must heat up and pressurize, or cool down and -depressurize the reactor _ coolant system (RCS) is determined by the -difference 'between-the maximum allowable pressure determined by . Appendix G of ASME Section XI, and the minimum required pressure for the reactor coolant pump (RCP). seals, adjusted for OPPS overshoot and ' instrument uncertainties. Under the P/T requirements of Appendix G fof ASME Section XI, OPPS can have-significant impact on operation by -limiting RCP operation at low temperatures. In

addition, the operating - pressure window imposed by OPPS becomes more and more restrictivs with reactor vessel service.

Reducing this operating window could potentially have an adverse safety impact by increasing I the possibility of inadvertent OPPS actuation due to pressure surges

associated with normal plant evolutions such' as RCP start and swapping operating charging pumps with the RCS in a water-solid i

condition.- ] The. impact on the P/T limits and OPPS setpoints has been evaluated due to increasing the. service period to 15 Effective Full Power Years (EFPY) based on ASME Section XI, Appendix G, requirements. The results indicate OPPS would significantly restrict the ability to perform plant heatup and cooldown, create an unnecessary burden to plant. operations, and challenge control of plant evolutions required with OPPS enabled.

a Exsmption R:quzst '- Paga 2 Procosed Alternative t , The use of ASME. Code Case N-640 requirements for. reactor vessel pressure limits at low temperatures is proposed as an alternative to ~ 10 CFR 50, Appendix G,fand ASME Section XI, Appendix G, requirements. Justification for Grantina Relief - Pursuant to : 10 CFR* 50.12, the Commission. may grant exemptions from the requirements-of 10 CFR 50 when:

1) the exemptions'are authorized' by law, will not present an undue risk to public health or safety, and are' consistent with the common defense and seenrity, and 2) when special circumstances are present... Special circumstances are present whenever, according to 10 CFR 50.12 (a) ( 2 ) (ii), " Application of the regulation in the particular. circumstances would not serve the underlying purpose of.the rule or is not necessary to achieve the underlying purpose of the rule."

~ The underlying purpose of 10 CFR 50.60, Appendix G, is to establish fracture toughness requirements for ferritic materials of pressure retaining. components of.the reactor coolant pressure boundary. These requirements provide adequate margins of safety during any condition of normal-operation, including anticipated operational occurrences, to which the pressure boundary may be subjected to over its service lifetime. Section IV.A.2 of this appendix requires that the reactor vessel be operated with P/T limits at least as conservative as those obtained by following the methods of analysis and the required margins of safety of Appendix G of ASME Section XI. Appendix G of Section XI of the ASME Code requires the P/T limits be calculated: a) using a safety factor of 2 on the principal membrane (pressure) stresses, b) margin added to the reactor vessel RTNDT in accordance with Regulatory Guide _1.99, Rev. 2, " Radiation Embrittlement of Reactor Vessel Materials," c) assuming a flaw at the -surface with a depth of 1/4 of the vessel wall thickness and a length of 6 times its depth, and d) using a conservative fracture toughness curve that is based on_the lower bound of static, dynamic, and crack arrest - fracture toughness tests on material similar to the reactor vessel material. In determining the setpoints for OPPS, ASME Code Case N-640 is applied. The lower bound fracture toughness curves available in Appendix G to Section XI uses the reference stress intensity factor Kra and is the basis for the current heatup and cooldown curves. Code. Case ~N-640 provides for normal operation within the P/T limits determined in accordance with ASME Section XI, Appendix G, but allows determination of setpoints for OPPS events.using the reference stress intensity ~ factor K c. Kra is a fracture toughness curve which is a I lower bound on all static, dynamic and arrest fracture toughness, and K c is a fracture toughness curve which is a lower bound on static I fracture toughness only. All other factors, including assumed flaw . size; and fracture toughness, remain the same as ASME Section XI, ' Appendix G, methodology. The basis of ASME Code Case N-640 indicated

Exrmption RIquset Page 3 that, due to the isothermal nature of the OPPS events, the margin with respect to toughness for an OPPS transient is within the range provided by ASME Section XI, Appendix G, for normal heatup and cooldown in the low temperature range. Thus, applying code Case N-640 will satisfy the intent of 10 CFR 50.60 for fracture toughness requirements. Further, application of Code Case N-640 will relieve operational restrictions. This will reduce the potential for inadvertent RCS pressure relief

events, thereby improving plant
safety, and will reduce unnecessary burdens on operators during important plant evolutions.

Imolementation Schedule The exemption request will be implemented within 60 days of approval of the associated license amendment request, and prior to reaching 10.0 EFPY. I t a}}

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