ML18052A351

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Proposed Tech Specs Reflecting New Reactor Vessel pressure-temp Limits for Heatup & Cooldown to Account for Effects of Irradiation on Vessel Matls
ML18052A351
Person / Time
Site: Palisades Entergy icon.png
Issue date: 03/17/1986
From:
CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
Shared Package
ML18052A350 List:
References
NUDOCS 8603200201
Download: ML18052A351 (27)
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IC0386-0043-NL04 ATTACHMENT Consumers Power Company Palisades Plant Docket 50-255 PROPOSED TECHNICAL SPECIFICATION PAGE CHANGES March 17, 1986 8603200201 9603 17 ,

  • P,DR *. ADOCK . If . . L-* 9 Pages e-3.1.2 Heatup and Cooldown Rates (Contd) (2) Before the end of the integrated power period for which Figures 3-1, -. . 3-2 and 3-3 apply, the l.imit lines on the figures shall be updated for a new integrated power period. The total integrated reactor thermal power from start-up to the end of the new power period shall be converted to an equivalent integrated fast neutron exposure (E 1 MeV). Such a conversion shall be made consistent with the . . (12) dosimetry evaluation of capsule W-290 * (3) The limit lines in Figures 3-1 *. 3-2 and 3-3 are based on the requirements of Reference 9, Paragraphs IV.A.2 and IV.A.3. These lines reflect a preservice hydrostatic test pressure of 2400 psig and a vessel flange material reference temperature of 60°F(S). Basis All components in the primary coolant system are designed to withstand the effects of cyclic loads due to primary system temperature and . (1) pressure changes. These cyclic loatls are introduced by normal unit load transients, reactor trips and start-up and shutdown operation.

During unit start-up and shutdown, the* rates of temperature and pressure changes are limited. A maximum plant heatup and cooldown rate of 100°F per hour is consistent with the design number of cycles and satisfies stress limits for cyclic operation.

(2) The reactor vessel plate and material opposite the core has been purchased to a specified Charpy V-Notch test result of 30 ft-lb or greater at an NDTT of + 10°F or less. The vessel weld has the highest RTNDT of plate, weld and RAZ materials at the fluence to which the Figures 3-1, 3-2 and 3-3 apply. (lO) The unirradiated RTNDT has been determined to be -56°F. (ll) An RTNDT of -56°F is used as an unirradiated value to which irr.adiation effects are added. In addition, the plate has been 100% volumetrically inspected by ultrasonic test using both 3-5 Proposed . TSP0386-0043-NL04

  • ' 3.1.2 Heatup and Cooldown Rates (Contd) Basis (Contd) longitudinal and shear wave methods. The remaining material in the reactor vessel, and other primary coolant system components, meets the appropriate .design code requirements and specific component function and has a maximum NDTT of +40°F. (S) -As a result of fast neutron irradiation in the region of the core, there will be an increase in the RT with operation.

The techniques used to predict the integrated fast neutron (E > 1 MeV) fluxes of the reactor vessel are described in Section 3.3.2.6 of the FSAR and also in Amendment 13,Section II, to the *FSAR. Since the neutron spectra and the flux measured at the samples and reactor vessel inside radius should be nearly identical, the measured transition shift from a sample can be applied to the adjacent section of the reactor vessel for later stages in plant life equivalent to the difference in calculated flux magnitude.

The maximum exposure of the reactor vessel will be obtained from the measured sample exposure by application of the calculated azimuthal neutron flux variation.

The predicted RTNDT shift for the base metai has been predicted based upon surveillance data and the US NRC Regulatory Guide. (lO) To compensate for any increase in the RT caused by irradiation, limits on the temperature relationship are periodically changed to stay within the stress limits during heatup and cooldown.

Reference 7

a procedure for obtaining the. allowable loadings for ferritic pressure-retaining materials in Class 1 components.

This procedure is based on the principles of linear elastic fracture mechanics and involves a stress intensity factor prediction which is a lower bound of static, dynamic and crack arrest critical valves.* The stress factor computed (7) is a function of -RTNDT' operating temperature, and vessel wall temperature gradients.

3-6 Proposed TSP0386-0043-NL04

.:.: 3 .1. 2 Heatup and Cooldown Rates (Contd) Basis (Contd) Pressure-temperature limit

  • calcul.ational procedures for the reactor coolant pressure boundary are defined in Reference 8 based upon Reference
7. The limit lines of Figures 3-1 through 3-3 consider a .54 psi pressure allowance to account for the fact that pressure is measured in the pressurizer rather than at the vessel beltline.

In addition, for calculational purposes, 5.°F and 30 psi were taken as measurement error allowances for temperature and pressure, respectively.

By Reference 7, reactor vessel wall locations at 1/4 and 3/4 thickness are limiting.

It is at these locations that the crack propagation associated with the hypothetical flaw must be arrested.

At these locations; fluence attenuation and thermal gradients have been evaluated.

During cooldown, the 1/4. thickness location is always more limiting in that the RTNDT is higher than that at the 3/4 thickness location and thermal gradient stresses are tensile there. During heatup, either the 1/4 thickness or 3/4 thickness location may be limiting depending upon heatup rate. Figures 3-1 through 3-3 define stress limitations only from a fracture mechanic's point of view. Other considerations may be more restrictive with respect to temperature limits. For normal operation; other inherent plant characteristics may limit the heatup and cooldown rates which can be achieved.

Pump parameters and pressurizer heating capacity tends to restrict both normal heatup and cooldown rates to less than 60°F per hour. The revised limits are applicable to reactor vessel inner wall fluences of up to 1.8 x 10 19 nvt. The application of appropriate fluence attenuation factors (Reference

10) at the 1/4 and 3/4 thickness locations results in RTNDT shifts of 241°F and 183°F, respectively, for the lim.iting weld material.

The criticality 3-7 Proposed TSP0386-0043-NL04 3.1. 2 Heatup and Cooldowl.ates (Contd) Basis (Contd) condition which defines a temperature below which the core cannot be made critical (strictly based upon fracture mechanics' considerations) is 371°F. The most limiting wall location is at 1/4 thickness.

The minimum criticality temperature, 371°F is the minimum permissible temperature for the inservice system hydrostatic pressure test. That temperature is calculated based upon 2310 psig inservice hydros ta.tic test pressure.

The restriction of heatup and cooldown rates to 100°F/h and the maintenance of a pressure-temperature relationship under the heatup, cooldown and inservice test curves of 3-1, 3-2 and 3-3, respectively, ensures that the requirements of References 6, 7, 8 and 9 are met. The core operational limit applies only when the reactor is critical.

The criticality temperature is determined per Reference 8 and the core operational curves adhere to the requirements of Reference

9. The inservice test curves incorporate allowances for the thermal gradients associated with the heatup curve used to attain inservice test pressure.

These curves differ from heatup curves only with respect to margin for primary membrane stress. (7)

  • Due to the shifts in RTNDT' NDTT requirements .associated with nonreactor vessel materials are, for all practical purposes, no longer limiting.

References (1) FSAR, Section 4.2.2. (2) ASME Boiler and Pressure Vessel Code,Section III, A-2000. (3). Battel,le Columbus Laboratories Report, "Palisades Pressure Vessel Irradiation Capsule Program:*

Unirradiated Mechanical Properties," August 25, 1977. (4) Battelle Columbus Laboratories Report, "Palisades Nuclear Plant Reactor Vessel Surveillance Program: Capsule A-240," March 13, 1979, submitted to the NRC by Consumers Power Company letter dated July 2, 1979. 3-8 Proposed TSP0386-0043-NL04 2600 2400 p 2200 R 2000 E s 1800 s 1600 u R 1400 E 1200 w p I s 1000 '° I G 800 600 400 200 0 I 1' I I I I I I I I I I I I I I I I I *---1---,---1 -i---i---r---r---r---r---r---

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-**--*-_ ,_ --_,_ --_,_ --:-' ---_, _ ---I I ---I----t----t----I----I--I I I . , I I --1----I-I I I I I I --_,_ --_,_ ---1----i -I I I I I *---r ---1* ---r -r ---r ---1-*-r----1---. -,---, ---, -1 I I I I I I I I I I I *I I --i---6-0-V ---:---I I 8 0 FI HR: I _:_ -I --. i----I ---,----,---I *1 00 F /H:R : -'-3 3 3 4 4 2 5 7 0 2 5 0 5 0 5 I -'-4 5 0 I I -,--,----, ----. . I I I I I -----, ---------I I I I I I I __ 1 ____ 1_ -' ----' --*--' ---J . r 4 5 5 5 5 6 7 0 2 5 7 0 5 0 5 0 5 0 TEMPERATURE DEGREES F Palisades Plant Pressure/Temperature for Fluence to 1.8 x 10 nvt Figure 3-2 for Cooldown w I ...... ...... p R E s s u R E p s I *G 2600 2400 2200 2000 1800 1600 1400 I I I -,---,---,-

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F /H-IR I I _ 1 *-L _ 1_ _ __ , ___ J _ 1000 800 600 400 200 0 I I I I I 4---..l---.J--_ .. i -I I I --I I I *---i-I I i I I I .1---,---,-______

__ _ i I I Ii I I I 5 7 0 5 . ' I' I I I J 1 I I I I' I I 1 0 0 1 2 5 1 5 0 I I 1 7 5 I I Ii 2 0 0 I I -' I -1 11 I 1 1 2 2 2 5 5 0 I -l I -L 'Ii Ii' I I 2 3 7 0 5 0 -L --' I I I I I 3 2 5 TEMPERATURE 3 5 0 I I -I-I :---_ .. _ -_, ----1 ---__ , -----1----I ----I I I I :80 : F /l;iR I -t-*I-IOQ I I I I -I_ -I_ *-,_ I --1----1-I I I -r ---r F iHR: I I -,-1 I 1 -1 I -,----:-: _,_ I I I I I --*-L --_1_ ---'--I --1----I --I .1 I --i----i*

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. e

  • 3.1.2 Heatup and Cooldown Rates (Contd) References (Contd) (S) FSAR, Section 4.2.4. (6) US Nuclear Regulatory Commission, Regulator Guide 1.99, "Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials," July, i975. (7) ASME Boiler and Pressure Vessel Code,Section III, Appendix G, "Protection Against Non-Ductile Failure," 1974 Edition. (8) US Atomic Energy Commission Standard Review Directorate of Licensing, Section 5.3.2, "Pressure-Temperature Limits." (9) 10 CFR Part SO, Appendix G, "Fracture Toughness Requirements," May 31, 1983. (10) US Nuclear Regulatory Commission, Regulatory Guide 1.99 Draft Revision 2, April, 1984. (11) Combustion Engineering Report CEN-189, December, 1981. (12) "Analysis of Capsules T-330 and W-290 from the* Consumers Power Company Palisades Reactor Vessel Radiation Surveillanc*e Program," WCAP-10637, September, 1984. 3.1.3 Minimum Conditions for Criticality a) Except during low-power physics test, the reactor shall not be made critical if the primary coolant temperature is below 525°F. b) In no case shall the reactor be made critical if the primary coolant temperature is below 371°F. c) When the primary coolant temperature is below the minimum temperature specified in "a" above, the reactor shall be subcritical by an amount equal to or greater than the potential
  • reactivity insertion due to depressurization.

d) No more than one control rod at a time shall be exercised or withdrawn until after a steam bubble and normal water level are established in the pressurizer.

e) Primary coolant boron concentration shall not be reduced until after a steam bubble and normal water level are established in the pressurizer.

3-12 Proposed TSP0386-0043-NL04

  • .. * ***" Basis At the beginning of life of the initial fuel cycle, the moderator temperature coefficient is expected to be slightly negative at operating temperatures withall control roas withdrawn.

(1) However, the ;uncertainty of the calculation is such that it is possible that a slightly positive coefficient could exist. The moderator coefficient at lower temperatures will be less negative or more positive than at operating temperature. (l, 2) It is, therefore, prudent to restrict the operation of the reactor when primary coolant temperatures are less than normal operating 525°F). Assuming the most pessimistic rods out moderator coeff.f.cient, the. maximum potential reactivity insertion that. could result from depressurizing the coolant from 21oo*psia to saturation pressure at 525°F is 0.1% 6p. During physics tests, special operating precautions will be taken. In addition, the strong negative Doppler coefficient(3) and the small integrated 6p would limit the magnitude of a power excursion resulting from a reduction of moderator density. The requirement that the reactor is not to be made critical below 371°F provides increased assurance that the proper relationship between primary coolant pressure and temperature will be maintained relative to the RTNDT of the primary coolant system pressure boundary material.

Heatup to this temperature will be accomplished by operating the primary coolant pumps. If the shutdown margin required by Specification 3.10.1 is maintained, there is no possibility of an accidental criticality as a result of an increase of moderator temperature or a decrease of coolant pressure.

Normal water level is established in the pressurizer prior to the withdrawal of control rods or the dilution of boron so as to preclude the possible overpressurization of a solid primary coolant system. References (1) FSAR, Table 3-2 . (2) FSAR, Table 3-6.:. * (3) FSAR, Table. 3-3. (next page is 3-13 Proposed TSP0386-0043-NL04 IC0386-0043-NL04 ATTACHMENT I Consumers Power Company Palisades Plant Docket 50-255 PALISADES REACTOR PRESSURE VESSEL TEMPERATURE LIMITS DETERMINATION March 1 7, 1986 16 Pages NUCLEAR OPERATIONS DEPARTMENT Engineering Analysis Work Sheet Attachment I EA -Sheet WJA-85-38 l of 16 Title Palisades Reactor Pressure Vessel Temperature Limits Determination I. Description II. Assumptions and Data III. Equation Development IV. Limit Tables v. Limit Curves I. DESCRIPTION To operate safely, the RPV must be kept in a ductile condition. fore, pressure -temperature limits are defined which ensure the RPV will not be stressed under conditions which promote brittle failure. The basic parameter used in determining these limits is the stress intensity factor, KI, which is a function of the stress state, material ties and flaw configuration.

The minimum K 1 that can cause failure is defined as the critical stress interisity factor or reference stress intensity KIR" Hl0885-0598A-TC01 2 x 1 R is 1 measure of the material's 1bility to 1rrest 1 crack of specified

  • length and depth. The crick used in this analysis is specified in the ASHE Boiler and Pressure Vessel Code (hereafter referred to as The Code),Section III Article G-2000 as the Maximum Postulated Defect (MPD). The MPD is said to be a sharp surface defect normal to the direction of maximum stress. It has a depth one-fourth the section thickness and a length of one-and-a-half times the section thickness.

The basic approach to determining new pressure temperature limits, fore, is to evaluate KIR for the material and determine the operating stresses.

Given the MPD as the limiting flaw, these (KIR & P) are evalu-. ated at the one-quarter and three-quarter wall thicknesses.

Determining KIR KIR depPnds on a materials reference nil ductility temperature (RTNDT) as stated in the Code. This dependancy is given by the .expression:

KIR = 26.78 + L233 exp [0.0145(T-RTNDT

+ 160)] Therefore, knowing RTNDT and the operating temperature (T), XIR can be determined.

Since RTNDT increases with increased neutron exposure, the shift (£1RTNDT) associated with projected fluence must be determined.

This also determines the length of time the new limits are valid. The jected fluence used here is 1.8 x 10 19 n/cm 2 which is equivalent to approximately 9 effective full power years (EFPY's), or until mately mid 1989 based on a capacity factor of 80%. Reg Guide 1.99 provides methodology for determining

£1RTNDT' For this analysis, the draft Revision 2 to Reg Guide 1.99 will be employed.

HI0885-0598A-TC01 3 Determining Qperating Stresses During operation, the RPV is subjected to pressure-induced membrane hoop stresses and stresses resulting from thermal gradients.

The hoop *stress a is calculated using the method of the USNRC Standard Review Plan and m . the thermal stress is covered in Appendix*

G-2000 .. Thermal stresses depend on thermal gradients through the RPV wall. Battelle Columbus Laboratory (BCL) has determined the Palisades RPV thermal gradients for various heatup and cooldown rates as shown in Table 1 (Reference 5). Determining Operating P/T Limits In Section III of this analysis, equations are derived which calculate allowable pressures for selected temperatures.

These temperatures are selected within specific temperature change rates for both heatup and cooldown conditions.

The result is tabulated P/T data from which the most limiting values are chosen and graphed as the Pressure-Temperature Limit Curves in Section V. In addition to operating P/T limits, limits are calculated for the vice hydrostatic test .. This is based on heatup rates only and mines a minimum temperature for the .test. This temperature

{s defined *.as the temperature below which the reactor cannot critical.

II. ASSUMPTIONS AND DATA 1. The temperature limitations of Appendix G, 10CFR50 respect to flange material, critical operation, and non-critical operation will.* be superimposed on the P/T limits contained in this document.

2. The reference temperature of the flange material is defined as 60°F *(Reference
6) .. HI0885-0598A-TC01 4 3. The P/T limits established in this document are based on a projected fluence of 1.8 x 10 19 n/cm 2* This translates to approximately 9 EFPY's. 4. The weld material is considered to be limit1ng.

Both the girth and longitudinal welds need to be considered, and their chemistries have been.established (see Reference

8) as follows:.

Girth Weld 0.21% Cu o;9scx, Ni Longitudinal Weld 0.19% Cu 1.10% Ni 5. The weld material initial reference temperature is -56°F according to Reference

7. 6. Temperature measurement error is +/- S°F. 1. Pressure measurement error is 84 psi due to: 28 psi pressure drop RPV inlet to beltline, 26 psi bead in the pressurizer and 30 psi measurement system error 8. The preservice hydrostatic test pressure is 2400 psi. MI0885-0598A-TC01
9. Miscellaneous Terms and Data ART_= Adjusted Reference Temperature

(°F) a * = RPV Inside Radius (Beltline)

= 86 inches b = RPV Outside Radius (Beltline)

= 94.5 inches CF =Chemistry Factor From R.G. 1.99 Revision 2 f = Fluerice = 1.8 x 10 19 n/cm2 Kim = Membrane Hoop Stress Intensity Factor KSI KIR = Reference Stress Intensity Factor KSI .;-r;-Kit = Thermal Stress Intensity Factor KSI .;-r;-H = Membrane Stress-To-Stress Intensity Factor m Ht = Thermal Stress-To-Stress Intensity Factor p = Operating Pressure = 2100 psig RTNDT = Reference Temperature

(°F)

= Reference Temperature Shift (°F) r = RPV Radius to Location x in Wall x t = RPV Wall Thickness at Beltline = 8.5 inches T = RPV Metal Temperature Actual (0 F) T = Measured RPV Metal Temperature

(°F) m

  • ATHAX = Through Wall Thermal Gradient AT = Thermal Gradient at Location X in Wall x q 1 = Standard Deviation on Initial RTNDT = 17°F = Standard Deviation on (Weld) = 28°F a = Membrane Hoop Stress m x =*Location in Wall @ x = 1/4t = 2.125 inches @ x = 3/4t = 6.375 inches a = Material Yield Strength y HI0885-0598A-TC01 5
  • III. A. P/T LIMIT EQUATIONS
  • DEVELOPMENT (Based on R.G. 1.99 Revision 2) The weld chemistries have been determined to include! Girth Weld 0.21% Cu 0.98% Ni Longitudinal Weld 0.19% Cu 1.10% Ni According to the Draft Reg Guide 1.99 Revision 2 (Reference 3), . adjusted reference temperatures are determined as foll 1 ows: 6 1. Apply the material chemistry to Table 1 of the Draft and obtain the chemistry factor (CF = 226) for the girth weld and CF = 229 for the logitudinal weld. 2. Divide the assumed fluence 1.8 x 10 19 by 10 19 and obtain f = 1.8. 3. The general equation for adjusted reference temperature (ART) is: ART = Initial RTNDT + t.RTNDT + Margin 4. Evaluate the terms: Initial RTNDT = -56 F (from Reference
7) Margin = 2 0 1 2 + 0 6 2 (from Reference 3). a = 17°F I a = 28 for welds (0.28 -0.10 log f) .6RTNDT surface = [CF]f ] -0.067x .6RTNDT attenuated

= [.6RTNDT surface e x = Loc.ation within Wall from ID Ml0885-0598A-TC01

5. Calculate the . Girth Weld .21 Cu .98 Ni CF = 226 Longitudinal Weld 0.19 Cu 1.10 Ni CF = 229 7 Surface=[CF)(0.28-0.10 logf) (226)(1.8)(0.28-o.*10 log 1.8) = 226 (1.161) 6RTND Surface=[CF)f

<0*28-0.lO logf) I c229)(1.8)(0.28-o.10 log 1.8) = 262°F @ 1/4t

= (aRT )e-0.067x

=

= (262)(0.867)

= 227°F = (229)(1.161)

= 266°F

= )e-0.067x

=C 266)e-6:6t1c.2s)(8.5)

= 231°F @ 3/4t

= (243)e-0:067(.75)(8.5)

= (246)e-0.067(.75)(8.5)

= 262(0.652)

= 266(0.652)

= 171 = 173 ART = Initial RT + + Margin @ l/4t ART = -56 + 277 + 2 17 2 + 282 ART = -56 + 231 + 66 ' = -56 + 211*+ 66 = 237 = 241 @ 3/4t ART =.-56 + 171 + 66 = 181 . HI0885-0598A-TC01 ART = -56 + 173 + 66 = 183 8 These ART's *re used to calculate the reference stress }ntensity factors (KIR) ta be. used in determining allowable pressures from Reference 1: KIR = 73 + l.233 exp 0.0145(T-RTNDT

+ 160)), where: KIR =

Intensity Factor T = RPV (Beltline)

Inlet Temperature RTNDT = Reference Temperature Equation (1) must be modified to account for thermal gradients in the vessel wall and temperature measurement error. The adjustment is as follows: T = T m Where: -5 +/- 6Tx T = Measured*

Temperature .m 6Tx = Thermal Gradient At Location X Therefore equation (1) becomes: KIR = 26. 78 + 1. 233 exp [_O. 0145 (Tm + 155 +/- 6Tx -RTNDT)] The thermal gradients are subtracted for heatup calculations and added for cooldown.

BCL determined the thermal gradients for selected heatup/cooldown rates in Reference

5. These are presented in Table 1. Using the ART values for RTNDT' we can calculate KIR for different heatup and cooldown rates at the 1/4 wall and 3/4 wall locations for the weld materials.

Girth Weld (2) Heatup@ 1/4T KIR = 26.78 +. 1.233exp [0.0145(Tm 6T 114)J *(3) Heatup@ 3/4T KIR =.26.78 + 1.233exp [0.0145(Tm 6T 314)] *(4) l/4T KIR = 26.78 + 1.233exp [0.0145(Tm

-82 + ;':(S) Cooldown@

3/4T K.IR = .26. 78 + 1.233exp

-26 + Longitudinal Weld *(2a) Heatup @. 1/4T KIR = 26.78 + 1.233exp [0.0145(Tm AT 114)] (3a) Heatup@ 3/4T KIR = 26.78 + 1.233exp [0.0145(Tm 6T 314)] (4a) Cooldown@

1/4T KIR = 26.78 + 1.233exp [0.0145(Tm

-86 + 6T 114>J *(Sa) Cooldown@

3/4T KIR = 26. 78 + l.233exp [0.0145(Tm

-28 + 6T 314)J *These are not limiting and need not be calculated MI0885-0598A-TC01

  • Equations 2-5 and 2a-5a give 1tress intensities at aelected temperatures and heatup/cooldown rates. What is required is to relate these values to able pressures so the.pressure-temperature limits may be drawn. 9 The RPV is subjected to both pressure-induced membrane hoop stress and thermal stress due to tbrough-wall thermal gradients.

The stress intensity factors for membrane hoop stress and.thermal stress are derived from the following fundamental requirement.

KIR > 2 Kim + Kit (6) Where Kim = Membrane Stress Intensity Factor Kit = Thermal Stress Intensity Factor Kim can be converted to membrane hoop stress (om) with equation 7. From Reference 1, Kim = M 0 (7) m m Where 0 m = Membrane Hoop Stress H =

Intensity Factor m Also, according to Reference 1, Klt can be expressed in terms of a maximum through-wall temperature gradient.

1 It = Mt (S) Where Mt = Stress-To-Stress Intensity Factor 6T = Through Wall Temperature Gradient max . Now equation 6 can be rewritten as: o + M 6T m t max (9) The Mm and Mt terms in eq 9 are obtained from Figures G-2214-1 and G-22,14-2, respectively in Reference

1. Determining Mt is straight Mt = 0.34 To find H we must first evaluate the ratio o /a m m y From Reference 5, the membrane hoop stress due to pressure is: Where: o = Pr m t (10) P = Operating Pressure (psig) r = Average Vessel Radius t = Vessel Thickness (inches) MI0885-0598A*TC01
  • U1ing the values given in Section II a = 22.3 ksi m Given that the yield stress for SA-302-B at 550°F is 44 ksi o = 0.506, M = 2.8 m
  • m oy By combining equations 9 and 10, we have K 1*R > 2HmPr + H (11) t max t Now we can solve equation 11 for P to obtain an expression for allowable pressures at calculated KIR values. t 2M r m x103* (12) *Note Equation 12 is multiplied by 1000 to convert ksi to psi. Substituting in the values for Hm, Mt, t, and r, we obtain the equations for allowable pressures*

P = 17.65 KIR -6.0 -84 (13) . max *Note 84 psi measurement error. 10 Table 2 summarizes the necessary equa_tions to generate P/T limits for various heat up and cooldown rates. Inservice Hydrostatic Testing Since the hydro is performed with the core not critical, heatup rates are low and thermal gradients through the wall are negligible.

Therefore, the thermal heatup condition will govern the test pressure limits. For purposes of scaling up the curves, however, limits for rates greater than 0°F/HR will be calculated.

HI0885-0598A-TC01 Only heatup rates need to be considered for the hydro and according to Reference 1, the equations for XIR during the hydro are: ICIR = 26. 78 + 1.233 exp (0.014S(T-RTNDT

+ 160)) KIR > 1.5 M a +Mt 6T (IS) . m m max. Therefore, using equations 10 and lS p = (XIR-Mt ( t x10 3 ] -84 1.SH r m Substituting the appropriate values P = 23.S3 KIR -8.0 6Tmax -84 Since we have assumed the 0°F/HR heatup rate to be limiting and since PHYDRO = 2310 psi, we have . PHYDRO = 23.10 KIR -8.0 6Tmax -84 KIR = 103.64 Substituting this value into equations 2 and 3 for the Girth Weld, we get @ 1/4t TCrit = 367°F @ 3/4t TCrit = 31t°F Substituting this value into equations 2a and 3a for the Longitudinal Weld, we get @ l/4t @ 3/4t TCrit = 371°F TCrit = 313or 11 From the above discussions, we see that although the RTNDT shifts are nearly equal, the longitudinal weld material is slightly more damaged and is, fore, more limiting.

In addition, .on heatup rates, either the 1/4t or 3/4t wall locations can be limiting.

On cooldown rates, the 1/4t.wall location is always limiting because RTNDT is higher and thermal gradient stresses are tensile there. Based on the tables in Section IV, the 1/4t wall location is found to be limiting for all conditions and the limit curves will all reflect the 1/4t values. For completeness, all equations are given in Table 2. MI088S-0598A-TC01 Temperature Change Rate °F/Hr AT 1/4 0 0 20 6.3 40 12.4 60 18.2 80 24.3 100 . 30.0 HI0885-0598A-TC01 TABLE I THER11AL GRADIENTS Heat up AT 3/4 AT max 0 0 13.3 14.2 26.3 28.0 38.6 41.1 51.4 54.7 63.2 67.3 12 Cool down AT 1/4 AT 3/4 AT max 0 0 0 4.4 9.5 10.1 9.0 19.2 20.5 13.9 29.8 31.8 18.5 39.6 42.2 . 23.6 50.6 54.0 13 WELD CONDITION LOC EQUATION Girth Heat up 1/4 T KIR = 26.78 + 1.233 exp [0.0145(Tm ATl/4)) P = 17.65 KIR -6.0 6T -84 max Girth Heat up 3/4 T KIR = 26.78 + 1.233 exp [0.0145(Tm .AT3/4)) p = 17.65 KIR -6. 0 6T -84 . I '. max Girth Cool down 1/4 T . KIR = 26. 78 + ,1.233 exp [0.0145(Tm

-82 + ATl/4)] P = 17.65 KIR -6.0 AT -84 . max Girth Cool down 3/4 T KIR = 26. 78 + 1.233 exp [0.0145(Tm

-26 + AT3/4)) P = 17.65 KIR -6.0 AT -84 max Girth Hydro.Test 1/4 T . KIR = 26.78 + 1.233 exp [0.014S(Tm ATI/4)) P = 23.53 KIR -8.0 6T -84 max Girth Hydro Test 3/4 T KIR = 26.78 + 1.233 exp [0.0145(TM AT3/4)) P = 23.53 KIR -8.0 AT -84 max

  • Longitudinal.

Heat up 1/4 T KIR = 26.78 + 1.233 exp [0.0145(Tm ATl/4)) P = 17.65 KIR -* 6.0 AT -84 . max Longitudinal Heat up 3/4 T KIR = 26. 78 + 1.233 exp [0.0145(Tht AT3/4)) P = 17.65 KIR -6.0 AT -84 max Longitudinal Cool down 1/4 T KIR = 26.78 + 1.233 exp [0.0145(Tm

-86 + ATl/4)) P = 17.65 KIR -6.0 6T -84 max Longitudinal Cool down 3/4 T KIR = 26.78 + 1.233 exp [0.0145(Tht

-28 + AT3/4)) P = 17.65 KIR -6.0 6T. -84 max Longitudinal Hydro Test 1/4 T KIR = 26.78 + 1.233 exp [0.0145(Tm ATl/4)] P = 23.53 KIR 8.0 AT -84 max Longitudinal Hydro Test 3/4 T KIR = 26. 78 + 1.233 exp [0.0145(Tm AT3/4)] P = 23.53 KIR -8.0 6T -84 max i

. . IV. LIMIT TABLES The following tables contain the results of pressure temperature limit calcul*tions based on Tables 1 and 2 of this document.

In all cases, the 1/4T location was found to be limiting.

MI0885-0598A-TC01 14

' .. 15 HEAT UP -I.ONGITUDINAL WELD Heat Up Rate Wall f:>.T. Allowable Pressure at Noted Temperatures 0 f/HR Loe 1/4 f:>.T (Max)

  • 50°F 100°F '150°F 200°F 250°f 300°F 350°F 400°F 450°F 0 1/4 0 0 402 415 444 sin 623 873 1389 2454 4654 20 1/4 6.3 14.2 315 328 354 407 518 746 1217 2189 4196 40 1/4 12.4 28.0 231 243 267 316 417 625 1056 1946' 3784 60 1/4 18.2 41.1 152 163 184 229 322 514 910 1729 3418 80 1/4 24.3 54. 7 70 79 99 140 225 401 764 1513 3059 100 1/4 30.0 67.3 -7 2 20 58 137 298 632 1322 2745 COOLDOWN -LONGITUDINAL WELD Cooldown Rate Wall f:>.T Allowable Pressure at Noted Temperatures 0 f /HR Loe 1/4 f:>.T (Max) 50°F 100°F 150°F 200°F 250°f 300°F 350°F 400°F 450°F 0 1/4 0 0 402 415 444 502 623 873 1389 2454 4654 20 1/4 4.4 10.1 341 356 387 449 578 845 1394 2530 4874 40 1/4 9.0 20.5 280 296 328 395 533 818 1406 2619 5125 60 1/4 13.9 31.8 214 230 265 337 485 79.1 1422 2725 5415 80 1/4 18.5 42.2 152 170 207 284 442 769 1444 2837 5713 100 1/4 23.6 54.0 83 102 142 225 395 747 1473 2973 6070 HYDRO -LONGITUDINAL WELD Heat Up Rate Wall f:>.T Allowable Pressure at Noted Temperatures

°F/HR Loe 1/4 f:>.T (Max)

  • 50°F 100°F 150°F 200°F 250°F 300°F 350°F 400°F 450°F 0 1/4 0 0 563 582 620 698 859 1192 1880 3300 20 .t/4 6.3 14.2 448 465 500 571 718 1022 1650 2946 5622 40 . 1/4 12.4 28.0 337 352 383 449 584 . 862 1436 2623 5072 60 1/4 18.2 41.1 231 245 274 334 458 713 1242 2332 4584 80 l/4 24.3 54. 7. 121 134 160. 215 328 563 1046 2045 4106 100 1/4 30.0 67.3 19 31 55 106 210 426 871 1790 3688 I : HTORRS-OS98A-TC01 I
  • 16 References
1.
  • ASME Boile_r and Pressure Vessel Code Section III, '77 S78 Appendix G-2000
  • 2. Regulatory Guide 1.99, Revision 1, April 1977 3. Draft Regulatory Guide 1.99, Revision 2 4. Battelle Columbus Laboratories Report, "Palisades Nuclear Plant Reactor Pressure Vessel Surveillance Program: Capsule A-240," March 13, 1979 5. USNRC Standard Review Plan Section 5. 3. 2, "Pressur.e Temperature Limits" 6. Branch Technical Position MTEB No. 5-2, "Fracture Toughness Requirements" 7. Evaluation of Pressurized Thermal Shock Effects Due to Small Break LOCA's With Loss of Feedwater for the Combustion Engineering NSSS," CEN-189, CE Power Systems, December 1981 8. June 14, 1985 submittal from CPCo to the Director of Nuclear Reactor Regulation

.. MI0885-0598A-TC01

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