ML20216D097

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Rev 3 to WCAP-14780, Prairie Island,Unit 1 Heatup & Cooldown Limit Curves for Normal Operation
ML20216D097
Person / Time
Site: Prairie Island Xcel Energy icon.png
Issue date: 02/28/1998
From: Abbott S, Christopher Boyd, Trombola D
WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20013F498 List:
References
WCAP-14780, WCAP-14780-R03, WCAP-14780-R3, NUDOCS 9803160275
Download: ML20216D097 (30)
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Text

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EXHIBIT E PRAIRIE ISLAND NUCLEAR GENERATING STATION l

License Amendment Request dated March 6,1998 l

WCAP-14780 Rev 3 February 1998

" Prairie Island Unit 1 Heatup and Cooldown Limit Curves for Normal Operation" l

l l

l 9803160275 990306 PDR ADOCK 05000282 P ,

PDR

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-14780, Revision 3 PRAIRIE ISLAND UNIT 1 HEATUP AND COOLDOWN LIMIT CURVES FOR NORMAL OPERATION S. L. Abbott February 1998 Work Performed Under Shop Order NLTP-106 l l

Prepared by Westinghouse Electric Company for Northem States Power Company Approved: ,

C. H. Boyd, ManageF Engineering & Materials Technology l l

Approved: 4 D. M. Trombola, Acting Manager Mechanical Systems Integration WESTINGHOUSE ELECTRIC COMPANY Nuclear Service Division P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355 C 1998 Westinghouse Electric Company All Rights Reserved 2/98

i PREFACE Revision 3:

Added Executive Summary.

Revised Table 1 to include correlation monitor material and additional weld data.

Added Table 2 to include nozzle shell base metal and weld Cu and Nl content data.

Renumbered Tables 3 through 10.

Revised Table 3 by adding nozzle shell and revising chemistry factors per RG 1.99, Rev.  !

2, Position 1.1.  !

Revised Table 4 to revise fluence values and chemistry factor calculations using surveillance capsule data.

Revised Table 5 to update fluence values based upon surveillance capsule data.

Revised Table 6 to show full tr.argins for surveillance data deemed not credible.

Revised Table 7 to show sample ART calculation for nozzle to intermediate shell weld as the limiting reactor vessel material.

Revised ART results in Table 8 based on surveillance capsule data and to including the nozzle shell base and weld materials. I

- Revised Figures 1 and 2 and Tables 9 and 10 in include new heatup and cooldown curves and data listings.

Added References 12 through 17 as sources of material data.

Revision 2:

Revised Initial RT,er and ART values on Table 6.

Revised ART Values on Table 7.

- Revised Tables 8 and 9 based c., revised ART values.

Revised Figures 1 and 2 based on revised ART values.

1 Revision 1:

Revised Tables 4,6 and 7 per updated fluences given in reference 5.

Added Table 9.

l l

Verified By: - - L l T. J. Lauliham Prairie Island Unit 1 Heatup and Cooldown Limit Curves for Normal Operation 2/98

ii EXECUTIVE

SUMMARY

The pressure-temperature limit curves for the Prairie Island Unit 1 plant heatup and cooldown were calculated using the adjusted RTuor (reference nil-ductility transition temperature) corresponding to the limiting beltline region material in the core region of the reactor vessel. The upper shell forging and the upper to intermediate shell weld seam were considered beltline materials along with the intermediate and lower shell forgings and the intermediate to lower shell weld seam because the upper shell extends below the top of the core.

The RTuor values for the materials were adjusted for the effects of exposure to fast-neutron radiation in accordance with NRC Regulatory Guide 1.99, Revision 2. The results of the Prairie Island Unit 1 Surveillance Capsule S analysis were included in the calculation of the Adjusted Reference Temperature (ART) values (initial RTuor + RTuor shift due to neutron exposure) at the 1/4T and 3/4T locations. The most limiting ART values were used in the generation of heatup and cooldown pressure-temperature limit curves at 35 EFPY (EOL). The upper to intermediate shell weld seam was found to be the limiting beltline material. See Figures 1 and 2 of this report for the curves which were generated.

l 1

I Praine Island Unit 1 Heatup and Cooldown Limit Curves for Normal Operation 2/98

iii TABLE OF CONTENTS PREFACE..........................,,,.......,,....,,,,,....,,,,,,,,,.........,,,,,.........,,,,,,,,,,,,... I EXEC UTIVE S U MMA RY . . . . . ,,,, , , . . . . . . . . ,, . . . . . . . . . . . . . .. ,,,,, ,, . . . . . . ... . . . . . . . . . . . . . . . . ,, , , . . . . . ,, , . . ,, , , ,,, ii LI ST O F TA B LE S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .iv .................

LI ST O F FIG U R E S . . . . . . , . . . . . . . . . . . . . . . . . . . . . . . . . . . . , , , , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . v .......

1 I NTR O D U CTI O N . , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , , . . . . . . . . , , , , , , , , , , . . . . . . . . . . . . . . . . . . . . . . . . , , , 1 2 FRACTU R E TOUG H N ESS PROPERTIES ...... ,,,,,........................ ,,,..... .............. .... 2 3 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE R E LATIO N S H I P S , , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , 7 4 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE,......... .... ....... ..., 10 5 HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES .......... 14 6 REFERENCES...................,,,.....,,................................................................. 22 Prairie Island Unit 1 Heatup and Cooldown Umit Curves for Normal Operation 2/98 m

f l iv i

LIST OF TABLES l Table 1 Calculation of Average Cu and Ni Weight Percent Values for Beltline Region Materials. .. . . . . . . . . . . ... . . . ... . . . . . . . ... . . . . . ... . . . .... . . . . . . . . . . .. . . . . . . . . . . . . . . . . .. ... . . . . 3 l

1 Table 2 Calculation of Average Cu and NI Weight Percent Values for Materials l Near Beltline R egion . . . . . . . ... . . . . .. . . . .. .. . .. .. . .. ... .. ... . . . . .. . . . . . . . . , ,,,,, . . . . . . . . . . . . . . . . . .. . . . 4 Table 3 Interpolation of Chemistry Factors from Regulatory Guide 1,99, Revision 2, Position 1.1.. . .. ... . . . . . .. . . . . . . . ... .. . . .. . . .. . . .. . . . . .. . . .. . . . . . . ,,, . . . . . . .. . . . . . 5 Table 4 Calculation of Chemistry Factors Using Surveillance Capsule Data per Regulatory Guide 1,99, Revision 2, Position 2.1.................. .......................... 6 Table 5 Fluence (10" n/cm', E > 1.0 MeV) on the Pressure Vessel Clad / Base Metal Interface for Prairie Island Unit 1,,, , .... . ........ ....... . ... ........... .... 11 Table 6 Margins for Adjusted Reference Temperature (ART) Calculations per Regulatory Guide 1,99, Revision 2 .... ... , ............... ... .... , ,,, .............. .. ,, 12 Table 7 Calculation of ART Values for the Limiting Prairie Island Unit 1 Reactor Vessel Material - Circumferential Weld (using Surveillance Capsule Data)...................................................,,,,,,,,.......................................,...12 l Table 8 Prairie Island Unit i 1/4T and 3/4T ART Calculations at 35 EFPY................. 13 Table 9 35 EFPY Heatup Curve Data Points (Without Instrumentation Error Margins)..................................................................................................18 Table 10 35 EFPY Cooldown Curve Data Points (Without Instrumentation Error Margins) . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . .......................................,19 i l

Prairie Island Unit 1 Heatup and Cooldown Umit Curves for Normal Operation 2/98

i LIST OF FIGURES Figure 1 Prairie Island Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rates up to 100'F/hr) Applicable for the First 35 EFPY (Without Margins i

for Instrumentation Errors)* ,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,16 l

l

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1 Figure 2 Prairie Island Unit 1 Reactor Coolant System Cooldown Umitations l (Cooldown Rates up to 100*F/hr) Applicable for the First 35 EFPY (Without Margins for instrumentation Errors)* ,,,,,,,,, ,,,,,,,,,,,,,,,, , ,,,,,,,,,,,,,,,17 l

Includes Vessel flange requirements of 116 F and 621 psig per 10CFR50, Appendix G.

l Prairie Island Unit 1 Heatup and Cooldown Limit Curves for Normal Operation 2/98 m

l 1 1 1 INTRODUCTION Heatup and cooldown limit curves are calculated using the adjusted RTer (reference nil-ductility l temperature) corresponding to the limiting beltline region material of the reactor vessel. The adjusted RT, of the limiting material in the core region of the reactor vesselis determined by using the unirradiated reactor vessel material fracture toughness properties, estimating the radiation-induced ARTer, and adding a margin. The unirradiated RTuor 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-Ib of impact energy and 35-mil lateral expansion (normal l to the major working direction) minus 60*F.

RTer increases as the materialis exposed to fast-neutron radiation. Therefore, to find the most l limiting RTer at any time period in the reactor's life, ARTer due to the radiation exposure associated with that time period must be added to the unirradiated RTer(IRTer) The extent of the shift in RTer 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""1. Regulatory Guide 1.99, Revision 2, is used for the calculation of Adjusted Reference Temperature (ART) values (IRTer + ARTer + 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.

Prairie island Unit 1 Heatup and Cooldown Limit Curves for Normal Operation 2/98

2 2 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 A . The beltline material properties of the Prairie Island Unit 1 reactor vessel presented in Table 1 are from References 3 through 7 and References 12 through 14. The material properties for the nozzle shell materials near the beltline region in Table 2 are from References 15 through 17.

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 3.) Additionally, surveillance capsule data is available for four capsules (Capsules V, P, R, and S) already removed from the Prairie Island Unit 1 reactor vessel. This surveillance capsule data was used to calculate chemistry factor (CF) values (Table 4) in adoition to those calculated per Tables 1 and 2 of Regulatory Guide 1.99, Revision 2.

i I

I Prairie Island Unit i Heatup and Cooldown Limit Curves for Normal Operation 2/98

3 Table 1 Calculation of Average Cu and Ni Weight Percent Values for Beltline Region Materials Ref. Intermediate Lower Shell InterJLower Shell Correlation Shell Forging Forging D Circumferential Weld (' M Monitor Material C(*3 Cu % Ni % Cu % Ni % Cu % Ni % Cu % Ni %

6 0.06 0.72 6 0.06 0.72 7 0.07 0.66 7 0.065 0.66 3 0.13 -

0.14 0.68 4 0.13 0.09 5 0.078 0.956 0.149 0.138 5 0.138 0.118 5 0.143 0.091 Avg. 0.07 0.80 0.07 0.66 0.14 0.11 0.14 0.68 Vessel Best Estimate Chemistry Average of Surv. 0.14 0.11 Weld Chemistry Heat 1752/1263(4 0.14 0.14 Heat 1752/1230(* 0.14 0.17 Heat 1752/1180(* 0.105 0.11 Best Estimate 0.13 0.13 NOTES:

(a) Surveillance Program base metal material.

(b) The Surveillance Program weld metal was fabricated with Weld Wire Type UM40, Heat No.1752, Flux Type UM89, Lot No.1230 and is identical to the intermediate to lower shell circumferential seam.

(c) Per Reference 12.

(d) Per Reference 13.

(e) Per Reference 14.

Prairie Island Unit 1 Heatup and Cooldown Limit Curves for Normal Operation 2/98

4 Table 2 Calculation of Average Cu and Ni Weight Percent Values for Materials Near Beltline Region Material  % Cu  % Ni Nozzle Shell Forging B" 0.075 0.68 Nozzle to intermediate Shell 0.17 0.15 Circumferential Weld" 0.12 0.14 Weld Material Average 0.15 0.15 NOTES:

(a) Per Reference 17 (b) Weld Seam W2 was fabricated with Weld Wire Type UM40, Heat No. 2269, Flux Type UM89, Lot No.1180 (References 15,16 and 17) l Prairie Island Unit 1 Heatup and Cooldown Limit Curves for Normal Operation 2/98

5 Table 3 Interpolation of Chemistry Factors from Regulatory Guide 1.99, Revision 2, Position 1.1 Ni, wt % Chemistry Factor, 'F Intermediate Shell Foraino C 0.80 44 Given Cu wt% = 0.07 Lower Shell Foraina D 0.66 44 Given Cu wt % = 0.07 Lower to Inter. Shell Cire. Weld Metal 0.00 58 Given Cu wt % = 0.13 0.13 69.7 0.20 76 Surveillance Weld Metal 0.00 61 Given Cu wt % = 0.14 0.11 70.9 0.20 79 Correlation Monitor Material 0.60 100 Given Cu wt % = 0.14 0.68 102 0.80 105 Nozzle Shell Foraina B 0.68 51 Given Cu wt % = 0.08 Nozzle to Inter. Shell Cire. Weld Metal 0.00 66 Given Cu wt% = 0.15 0.15 79.5 0.20 84 Bold Values are the Interpolated Chemistry Factors.

Ratio Procedure:

CFv.wwe = 69.7 CFs wm = 70.9 Ratio = (CFv.wwa + CFs,wa) = 69.7 + 70.9 = 0.98 Prairie Island Unit 1 Heatup and Cooldown Limit Curves for Normal Operation 2/98

6 i-

, Table 4 Calculation of Chemistry Factors Using Surveillance Capsule Data Per Regulatory Guide 1.99, Revision 2, Position 2.1 Material Capsule Capsule FF(* ARTer(* FF*ARTer FF 2 g<o

! Intermediate Shell V 0.6267 0.869 24.07 20.92 0.755 I Forging C 2

l (Axial) # 1.314 1.076 33.98 36.56 1.158 l

l R 4.000 1.356 84.18 114.15 1.839 S 4.338 1.373 74.27 101.97 1.885 Intermediate Shell V 0.6267 0.869 56.36 48.98 0.755 Forging C (Tangential) P 1.318 1.076 23.11 24.87 1.158 R 4.000 1.356 95.85 129.97 1.839 S 4.338 1.373 101.46 139.30 1.885 SUM 616.72 11.27 2

CFw.nn.o sn.a reno = I(FF

  • ARTmT) + I(FF )

= 54.7'F Weld Metal (* V 0.6267 0.869 34.38 33.69(* 29.28 0.755 P 1.314 1.076 45.15 44.25(* 47.61 1.158 R 4.000 1.356 122.47 120.02(* 162.75 1.839 S 4.338 1.373 160.43 157.22(* 215.86 1.885 SUM 455.50 5.64 2

CFw.o um = I(FF

  • ARTer) + I(FF )

= 80.8"F l

NOTES. -

(a) f = fluence (10 n/cm ,2E > 1.0 MeV). All updated fluence values were taken from the Capsule S ,

analysis (WCAP-14779 f

i. I 8 28' " 09

, (b) FF = fluence factor = f l (c) ARTer values were obtained from CVGRAPH Version 4.1 (See WCAP-14779t g, l (d) The reactor vessel intermediate to lower shell circular weld seam was made with the same weld wire and flux as the surveillance weld specimens (Wire UM40, heat number 1752, UM 89 flux, batch no.

1230). The ratio procedure is used since the average Cu and Ni content of the vessel weld differs from that of the surveillance weld material (See Table 3).

(e) Ratio of 0.98 applied (See ' Ratio Procedure' calculation on previous page).

Prairie Island Unit 1 Heatup and Cooldown Limit Curves for Normal Operation 2/98

7 3 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS Appendix G to 10 CFR Part 50, " Fracture Toughness Requirements"M specifies fracture toughness requirements for ferritic materials of pressure-retaining components of the reactor coolant pressure boundary of light water nuclear power reactors to provide adequate margins of safety during any conditionof normal operation, including anticipated operational occurrences and system hydrostatic tests, to which the pressure boundary may be subjected over its service lifetime. The ASME Boiler and Pressure Vessel Code forms the basis for these requirements. Seebon XI, Division 1, " Rules for Inservice Inspection of Nuclear Power Plant Components"M, Vessels, contains the conservative methods of analysis.

L 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, N , for the metal temperature at that time. N is obtained from the reference

! fracture toughness curve, defined in Appendix G of the ASME Code,Section XI. The N l curve is given by the following equation:

l Ku = 26.78 + 1.223

  • etuuswa,+mi (3y where, l N = reference stress intensity factor as a function of the metal temperature T and the metal j reference nil-ductility temperature RTer l

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

1 C

  • Kw+ Ka < Ku (2) where, Km = stress intensity factor caused by membrane (pressure) stress j Ka = stress intensity factor caused by the thermal gradients l N = function of temperature relative to the RTer of the material 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 Prairie Island Unit 1 Heatup and Cooldown Limit Curves for Normal Operation 2/98

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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 3/4T location, the appropriate value for RT,er, 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 pressure-temperature relations are generated for both k 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 wtually 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 reactor coolant temperature, the AT (temperature) 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 I

than the steady-state value.

The above procedures are needed because there is no direct control on temperature at the 1/4T 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 l

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 i 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 l l 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 intemal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the K,, for the 1/4T crack during heatup I

is lower than the K, for the 1/4T crack 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 Ki 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, Prairie Island Unit i Heatup and Cooldown Limit Curves for Normal Ooeration 2/98 L

9 1

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

The second portion of the hertup analysis concems 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 vesselinside 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 at the outside are tensile and increase with increasing heatup rates, each 1 l heatup rate must be analyzed on an individual basis.

i I

Following the generation of pressure-temperature curves for both the steady state and finite heatup rate situations, the final limit curves are produced 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 l conservative heatup limitations because it is possible for conditions to exist wherein, over the l 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 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 RTuor by at least 120'F for normal operation when the pressure exceeds 20 percent of the preservice hydrostatic test pressure (3106 psig), which is 621 psi for Prairie Island Unit 1. j l  :

! The limiting unirradiated RTuor of-4*F occurs in the closurehead/ vessel flange of the Prairie Island Unit i reactor vessel, so the minimum allowable temperature of this region is 116 F at i pressures greater than 621 psi. This limit (where the horizontal line indicates that the pressure l shall not exceed 621 psi for temperatures less than 116 F) is shown as a notch in the curves, presented wherever applicable in Figures 1 and 2.

i l

l Prairie Island Unit 1 Heatup and Cooldown Limit Curves for Normal Operation 2/98

10 4 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 = InitialRTer+ A RTer+ Margin (3)

Initial RTer is the reference temperature for the unirradiated material as defined in paragraph NS-2331 of Section lll of the ASME Boiler and Pressure Vessel CodeD 9. If measured values of initial RTer 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 deviation for the class.

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

A RTer = CF

  • f*"*"* (4)

To calculate ARTer 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,,,w = f ,,

  • e f#" (5) where x inches (vessel beltline thickness is 6.692 inches) is the depth into the vessel wall l measured from the vessel clad / base metal interface. The resultant fluence is then placed in Equation 4 to calculate the ARTer at the specific depth. The fluence (E > 1.0 MeV) values on j the pressure vessel clad / base metal interface for the Prairie Island Unit i reactor vessel are  !

presented in Table 5.  !

l Prairie Island Unit i Heatup and Cooldown Limit Curves for Normal Operation 2S8

11 Table 5 Fluence (10" n/cm 8, E > 1.0 MeV) on the Pressure Vessel -

Clad / Base Metal interface for Prairie Island Unit 1 m EFPY O' 15' 30' 45' Calculated Fluences @ Core Mid-Plane 18.12 2.45 1.55 1.10 0.943 24 2.97 1.93 1.40 1.20 35 3.95 2.62 1.95 1.69 Calculated Fluences @ Top of Core 18.12 1.37 0.865 0.613 0.526 24 1.66 1.07 0.778 0.670 35 2.20 1.46 1.09 0.940 The chemistry factor values obtained from Tables 1 and 2 of Regulatory Guide 1.99, Revision 2, were determined in Table 3 using the copper and nickel content values reported in Tables 1 and 2 of this report. Chemistry factors were also calculated using surveillance capsule data as shown in Table 4.

Margin is calculated as, M = 2 Voi2 + e32 The standard deviation for the initial RTwor margin term, ei, is 0 F when the initial RTwor is a measured value, and 17*F when a generic value is available. The standard deviation for the ARTwor margin term, e3, is 17 F for plates or forgings, and 8.5*F for plates or forgings (half the value) when credible surveillance data is used. For welds, c4 si equal to 28 F when surveillance capsule data is not used, and is 14*F (half the value) when credible surveillance capsule data is used. e3 need not exceed one-half j the mean value of ARTwor. See Table 6. In the case of Prairie Island Unit No.1, the i surveillance capsule forging material data is deemed to be not credible in accordance with the Regulatory Guide 1.99, Revision 2Wcredibility criteria. Since data points for all of the calculated ARTwor values do not fall within the prescribed margins for the surveillance capsule data, NSP has chosen to use the full e3 margin of 17 F for forgings Prairie Island Unit 1 Heatup and Cooldown Limit Curves for Normal Operation 2/98

12 Table 6 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 IRTuor 34 34

  • Generic IRTwar 48 38 WELD METAL Measured IRTsar 56 56
  • l Generic IRTwor 66 44 j NOTE l (a) Full c amargins used since the surveillance capsule data was deemed not credible.

All materials in the beltline region of the Prairie Island Unit 1 reactor vessel were considered in  !

determining the limiting material. Sample calculations to determine the ART values for the j Circumferential Weld are shown in Table 7. The resulting ART values for all beltline materials j at the 1/4T and 3/4T locations are summarized in Table 8. From Table 8, it can be seen that the limiting material is the Circumferential Weld (using surveillance capsule data). Therefore, the 1/4T and 3/4T ART values for the Circumferential Weld will be used in the generation of the heatup and cooldown curves.

Table 7 Calculation of ART Values for the Limiting Prairie Island Unit 1 Reactor Vessel Material - Nozzle to Intermediate Shell Circumferential Wald (Heat 2269)

Parameter Operating Time 35 EFPY Location 1/4T 3/4T Chemistry Factor, CF (*F) 79.5 79.5 2

Fluence, f (10* n/cm )(.' 1.47 0.66 Fluence Factor, FF 1.11 0.884 i 1

ARTuor = CF x FF (*F) 88.2 70.3 i Initial RTuor, I (*F) 0 0 Margin, MW(F) 66 66 Adjusted Reference Temperature" (ART), (*F) 154 136 NOTES.

(a) Fluence, f, is based upon f.,,(10* n/cm'. E>1.0 MeV) = 2.20 at 3s EFPY (for the Top of the Core).

(b) The Praine island Unit 1 reactor vessel wall thickness is 6 692 inches at the bottline region. j (c) The genene margin of 66*F was used since the weid matenal is not included in the surveillance capsule program.

Prairie Island Unit 1 Heatup and Cooldown Limit Curves for Normal Operation 2/98

13 Table 8: Prairie island Unit 11/4T and 3/4T ART Calculations at 35 EFPY Material CF f @35 1/4T f 1/4T FF l M ART ARTsc7 EFPY) 3/4T f 3/4T FF 1/4T Calculations Intermediate Shell 44.0 3.95 2.64 1.26 14 34 55.4 103 Forging C Using S/C Data 54.7 3.95 2.64 1.26 14 34*) 68.9 117 Lower Shell Forging D 44.0 3.95 2.64 1.26 -4 34 55.4 85 Circumferential Weld 69.7 3.95 2.64 1.26 -13 56 87.8 131 Using S/C Data 80.8 3.95 2.64 1.26 -13 56*) 101.8 145 I 1

Nonle to Inter. Shell 79.5 2.20 1.47 1.11 0(* 66 88.2 154 Cire Weld (Heat 2269)

Nonle (Upper) Shell 51 2.20 1.47 1.11 -4 34 56.6 87 Forging B 3/4T Calculations Intermediate Shell 44.0 3.95 1.18 1.05 14 34 46.2 94 Forging C Using S/C Data 54.7 3.95 1.18 1.05 14 34*) 57.4 105 Lower Shell Forging D 44.0 3.95 1.18 1.05 -4 34 46.2 76 Circumferential Weld 69.7 3.95 1.18 1.05 -13 56 73.2 116 Using S/C Data 80.8 3.95 1.18 1.05 -13 56*) 84.8 128 Nonle to Inter. Shell 79.5 2.20 0.660 0.884 0(4 66 70.3 136 l Circ Weld (Heat 2269)

Nonle (Upper) Shell 51 2.20 0.660 0.884 -4 34 45.1 75 i Forging B l NO.IE 2

(a) Fluence values are x 10 n/cm (E > 1.0 MeV). In addrtion, the values used are the calculated values since they are higher than the best-estimate values.

(b) The full e4 rnargin of 17'F for the forging and 28'F for the weld was used since the sury. data was deemed not credble (per Secten 4).

(c) Estrnated per Standard Revow Plan Secton 5.3.2 (See Ref.17).

(d) When two or more credble s'avedlance data sets become available. the data sets rney be used to determsne ART values as desenbed in Regulatory Guide 1.99, Revison 2. Posebon 2.1. If the ART values based on survedlance capsule data are larger than those calculated per Regulatory Guide 1.99. Revisen 2. Poston 1.1, the survallance data should be used. If the survallance capsule data grves lower values, other may be used. In the case of the intermedate to lower shell creumferenbal weld, the survedlance data is ,

deemed to be not credble. However, the full margin term is included in orde* to obtain the most conservatrve result using the survedlance data. Once again, larger ART values based on the surveillance data should be u2d.)

Prairie island Unit 1 Heatup and Cooldown Limit Curves for Normal Operation 2/98

1 14 5 HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES Pressure-temperature limit curves for normal heatup and cooldown of the primary reactor coolant system have been calculated for the pressure and temperature in the reactor vessel beltline region using the methods!"1 discussed in Section 3 and 4 of this report. Since indication of reactor vessel beltline pressure is not available on the plant, the pressure difference between the wide-range pressure transmitter and the limiting beltline region must be accounted for when using the pressure-temperature limits presented in Figures 1 and 2.

Figure 1 presents the heatup curves without margins for possible instrumentation errors using heatup rates up to 100*F/hr applicable for the first 35 EFPY. Figure 2 presents the cooldown curves without margins for possible instrumentation errors using cooldown rates up to 100'F/hr applicable for 35 EFPY. Allowable combinations of temperature and pressure for specific temperature change rates are below and to the right of the limit lines shown in Figures 1 and 2.

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 Figure 1. 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 10 CFR Part 50. The governing equation for the hydrostatic test is defined in Appendix G to Section XI of the ASME Code as follows:

1.S K1.< Ku (6)

where, Ki m is the stress intensity factor covered by membrane (pressure) stress, j Ki,= 26.78 + 1.223 e N ' " * '",

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

The criticality limit curve specifies pressure-temperature limits for core operation to provide additional margin during actual power production as specified in Reference 8. The pressure-temperature 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 minimum temperature required for the inservice hydrostatic test, and at least 40'F higher than the minimum Prairie Island Unit 1 Heatup and Cooldown Limit Curves for Normal Operation 2/98

15 permissible temperature in the corresponding pressure-temperature curve for heatup and cooldown calculated as described in Section 3 of this report. The minimum temperatures for the inservice hydrostatic leak tests for the Prairie Island Unit 1 reactor vessel at 35 EFPY is 273*F.

l The vertical line drawn from these points on the pressure-temperature curve, intersecting a curve 40*F higher than the pressure-temperature limit curve, constitutes the limit for core operation for the reactor vessel.

Figures 1 and 2 define all of the above limits for ensuring prevention of nonductile failure for the

j. Prairie Island Unit 1 reactor vessel.

The data points used for the heatup and cooldown pressure-temperature limit curves shown in Figures 1 and 2 are presented in Tables 9 and 10.

l 4

l l

t Prairie Island Unit 1 Heatup and Cooldown Limit Curves for Normal Operation 2S8

I t 18 MATERIAL PROPERTY BASIS LIMITING MATERIAL: NOZZLE TO INTERMEDIATE SHELL CIRCUMFERENTIAL WELD LIMITING ART VALUES AT 35 EFPY: 1/4T,154'F 3/4T,136*F l

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FIGURE 2 Prairie Island Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100*F/hr) Applicable for the First 35 EFPY (Without Margins for instrumentation Errors)  !

includes Vessel flenge requwements of 116*F and 621 peig per 10CFRSO, Appendet o.

I Praine Island Unit 1 Heatup and Cooldown Limit Curves for Normal Operation 2/98 l

18 TACLE 9 35 EFPY Heatup Curve Data Points (Without Instrumentation Error Margins)

Heatup Curves 60 Critical. Limit 100 Heatup l Critical. Limit Leak Test Limit Heatup T P T P T P T P T P 60 0 273 0 60 0 273 0 251 2000 60 584 273 594 60 560 273 560 273 2485 65 584 273 587 65 560 273 560 85 584 273 584 85 560 273 560 90 584 273 584 90 560 273 560 95 584 273 586 95 560 273 560 100 586 273 591 100 560 273 560 105 591 273 597 105 560 273 560 110 597 273 604 110 560 273 562 115 604 273 613 115 562 273 566 120 613 273 622 120 566 273 571 125 622 273 633 125 571 273 577 130 633 273 645 130 577 273 585 135 645 273 658 135 585 273 594 140 658 273 672 140 594 273 604 145 672 273 687 145 604 273 615 150 687 273 704 150 615 273 627 155 704 273 722 155 627 273 641 160 722 273 741 160 641 273 656 165 741 273 761 165 656 273 672 170 761 273 784 170 672 273 690 175 784 273 808 175 690 273 709 180 808 273 833 180 709 273 730 185 833 273 861 185 730 273 752 190 861 273 891 190 752 273 777 195 891 273 923 195 777 273 802 ,

200 923 273 957 200 802 273 831 205 957 273 994 205 831 273 861 210 994 273 1033 210 861 273 893 215 1033 273 1076 215 893 273 928 220 1076 273 1121 220 928 273 966 225 1121 273 1170 225 966 273 1006 230 1170 275 1223 230 1006 275 1049 235 1223 280 1279 235 1049 280 1096 240 1279 285 1339 240 1096 285 1146 245 1339 290 1404 245 1146 290 1199 250 1404 295 1473 250 1199 295 1257 255 1473 300 1548 255 1257 300 1318 260 1548 305 1628 260 1318 305 1384 265 1628 310 1713 265 1384 310 1455 270 1713 315 1805 270 1455 315 1531 Praine Island Unit 1 Heatup and Cooldown Limit Curves for Normal Operation 2S8

[ .

19 ,

1 1

TABLE 9 (C!ntinued)

, Heatup Curves Data Points Applicable to 35 EFPY l

(without Margins for Instrumentation Errors)

Heatup Curves 50 Heatup Gntical. Limit 100 Heatup Gntical. Limit Leak Test Limit T P T P T P T P T P 275 1805 320 1903 275 1531 320 1612 280 1903 325 2007 280 1612 325 1699 l 285 2007 330 2119 285 1699 330 1792 I 290 2119 335 2231 290 1792 335 1892 295 2231 340 2347 295 1892 340 1998 300 2347 345 2471 300 1998 345 2112 305 2471 305 2112 350 2233 310 2233 355 2363 315 2363 I

I l

l Prairie Island Unit 1 Heatup and Cooldown Limit Curves for Normal Operation 2/98 I

I 20 l TAELE 10 i Cooldown Curves Data Points Applicable to 35 EFPY (without Margins for Instrumentation Errors)

Cooldown Curves Steady State 20*F 40*F 60*F 100'F T P T P T P T P T P 60 0 60 0 60 0 60 0 60 0 60 590 60 563 60 537 60 510 60 455 65 594 65 568 65 542 65 515 65 460 70 599 70 573 70 547 70 520 70 465 75 605 75 579 75 552 75 526 75 471 80 611 80 585 80 558 80 532 80 478 85 617 85 591 85 565 85 539 85 485 90 621 90 598 90 57?. 90 546 90 493 95 621 95 605 95 580 95 554 95 502 103 621 100 613 100 588 100 563 100 511 105 621 105 621 105 597 105 572 105 520 110 621 110 621 110 607 110 582 110 531 115 621 115 621 115 617 115 592 115 543 116 621- 116 621 120 628 120 604 120 555 116 668 116 644 125 640 125 616 125 568 120 676 120 652 130 653 130 630 130 583 125 687 125 664 135 667 135 644 135 599 130 699 130 676 140 682 140 660 140 615 135 712 135 690 145 698 145 676 145 634 140 726 140 704 150 715 150 695 150 653 145 741 145 720 155 734 155 714 155 674 150 757 150 736 160 754 160 735 160 697 155 774 155 754 165 776 165 757 165 722 160 793 160 773 170 799 170 782 170 748 165 813 165 794 175 824 175 808 175 777 170 834 170 816 180 851 180 836 180 808 175 857 175 841 185 880 185 866 185 841 180 882 180 866 190 911 190 899 190 876 185 909 185 894 195 945 195 934 195 915 190 937 190 924 200 981 200 972 200 956 195 968 195 956 205 1019 205 1012 205 1001 200 1001 200 990 210 1061 210 1056 210 1048 205 1036 205 1027 215 1106 215 1102 215 1100 210 1075 210 1067 220 1154 220 1153 220 1155 215 1115 215 1110 225 1205 220 1159 220 1156 225 1206 225 1205 230 1257 235 1311 240 1370 245 1432 I

i Prairie Island Unit i Heatup and Cooldown Limit Curves for Normal Operation 2/98

21 TABLE 10 (Continued)

I Cooldown Curves Data Points Applicable to 35 EFPY (without Margins for Instrumentation Errors)

Cooldown Curves steady State 20*F 40*F 60*F 100*F T P T P T P T P T P l 250 1500 255 1572

< 260 1649 l 265 1732 l 270 1820 275 1915 280 2017 285 2126 290 2243 295 2367 i

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Prairie Island Unit i Heatup and Cooldown Limit Curves for Normal Operation 258 i

22 6- 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 MTEB 5-2, Chapter l 5.3.2 in Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition, NUREG-0800,1981.

l

3. WCAP-8086, "Northem States Power Co. Prairie Island Unit No.1 Reactor Vessel Radiation Surveillance Program", S.E. Yanichko and D.J. Lege, June 1973.

! 4. WCAP-11006, " Analysis of Capsule R from the Northem States Power Company Prairie island Unit 1 Reactor Vessel Radiation Surveillance Program", R.S..Boggs, et.

l al., February 1986.

I 5. WCAP-14779 Rev. 2, " Analysis of Capsule S from the Northern States Power Company Prairie Island Unit i Reactor Vessel Radiation Surveillance Program". S.L Abbott, February 1998.

6. Societe Des Forges et Ateliers Du Creusot usines Schneider, Chemical Analysis l Report No.17-9-2, NSP shell course C, heat 21918/38566.

! 7. Sooete Des Forges et Ateliers Du Creusot usines Schneider, Chemical Analysis Report No.15-8-1, NSP shell course D, heat 21887/38530.

1

! 8. 10 CFR Part 50, Appendix G, " Fracture Toughness Requirements", Federal Register, Volume 60, No. 243, dated December 19,1995.

I 9, 1992 Section XI, Division 1, of the ASME Boiler and Pressure Vessel Code, Appendix G, " Vessels".

L 10. 1989 Section lil, Division 1 of the ASME Boiler and Pressure Vessel Code, Paragraph NB-2331, " Material for Vesseis".

L l 11. WCAP-14040-NP-A, " Methodology Used to Develop Cold Overpressure Mitigating l

System Setpoints and RCS Heatup and Cooldown Limit Curves", J. D. Andrachek, et al., January 1996.

12. Proces Vert >al De Recette De Produits De Soudre, S.A.F., Code No. 118 716 AS4, Spec # PS 308/R, Designation UM 40 (fil)- UM 89 (flux), Lot No. 1752-69 (wire)-1263 (flux), Dated 3/26/70.

l Prairie Island Unit 1 Hestup and Cooldown t.imit Curves for Normal Operation 2/98

)

23

13. Proces Verbal De Recette De Produits De S:udre, S.A.F., Coda N3.118 719 A54, Spec
  1. PS 308/R, Designation UM 40 (fil) - UM 89 (flux), Lot No. 1752-69 (wire)-1230 (flux),

Dated 3/31/70.

14. Proces Verbal De Recette De Produits De Soudre, S.A.F., Code No. 118 716 A54, Spec
  1. PS 308/R, Designation UM 40 (fil) - UM 89 (flux), Lot No. 1752-69 (wire)-1180 (flux),

Dated 10/13/68.

l 15. Proces Verbal De Recette De Produits De Soudre, S.A.F., Code No. 118 716 A54, Spec l # PS 308/R, Designation UM 40 (fil) - UM 89 (flux), Lot Nn. 2269 (wire) -1230 (flux),

l Dated 3/31/70.

I i

16. Proces Verbal De Recette De Produits De Soudre, S.A.F., Code No. 118 716 A54, Spec l
  1. PS 308/R, Designation UM 40 (fil) - UM 89 (flux), Lot No. 2269 (wire) -1180 (flux),

Dated 3/31/70.

17. MM-SME-2925, "NRC Request for Information on Prairie Island Unit No.1 and 2 Reactor Vessel Materials Surveillance Program," S. E. Yanichko,10/19/77.

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Prairie Island Unit 1 Heatup and Cooldown Limit Curves for Normal Operation 2/98

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