ML20058A478

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Final Rept on Reactor Vessel Pressure-Temp Limits for Calvert Cliffs Unit 2 for 12 Efpys
ML20058A478
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Site: Calvert Cliffs Constellation icon.png
Issue date: 06/30/1990
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ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY
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ML20058A468 List:
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NUDOCS 9010260233
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,537. doc (9033)/kr-1 Attachment (2)

FINAL REPORT c

ON REACTOR VESSEL PRESSURE-TEMPERATURE LIMITS FOR CALVERT CLIFFS UNIT 2 FOR 12 EFFECTIVE FULL POWER YEARS Prepared For:

BALTIMORE GAS & ELECTRIC COMPANY I

CALVERT CLIFFS NUCLEAR GENERATING STATION L

LUSBY, MARYLAND 20657 I

I By-ABB COMBUSTION ENGINEERING NUCLEAR POWER COMBUSTION ENGINEERING, INC.

i.

REACTOR VESSEL INTEGRITY GROUP l

1000 PROSPECT HILL ROAD WINDSOR, CONNECTICUT 06095-0500 l

h June 1990 9010260233 901022 DR ADOCK 050 0

537. doc (9033;/kr-2 TABLE OF CONTENTS SECTION lilLE EAG.E

1.0 INTRODUCTION

5 2.0 ADJUSTED REFERENCE TEMPERATURE 6

PROJECTIONS 3.0 PRESSURE - TEMPERATURE LIMITS 11 3,1 GENERAL APPROACH FOR CALCULATING 11 PRESSURE-TEMPERATURE LIMITS 3.2 THERMAL ANALYSIS METHODOLOGY 15 3.3 COOLDOWN LIMIT ANALYSIS 16 3.4 HEATUP LIMIT ANALYSIS 18 3.5 HYDROSTATIC TEST AND CORE CRITICAL 20 LIMIT ANALYSIS 3.0 LOWEST SERVICE TEMPERATURE, MINIMUM BOLTUP 21 TEMPERATURE, AND MINIMUM PRESSURE LIMITS 4.0 LTOP ENABLE TEMPERATURES 23 5.0 DAIA 24

6.0 REFERENCES

25 Page 2

,537.dtc(9033)/kr-3 e

LIST Of TABLES liq 11RE 11GE i

1 Calvert Cliffs Unit 2 Reactor Vessel Beltline 27 Materials 2

Calvert Cliffs Unit 2 Beltline ART Calculations 28 3

BG&E Calvert Cliffs Unit 2 P-Allowable (Ksi) vs. RCS 29 Temperature (Deg. F) for 12 EFPY, Normal Operation 4

BG&E Calvert Cliffs Unit 2 P-Allowable (Ksi) vs. RCS 30 Temperature (Deg. F) for 12 EFPY, Hydrostatic i

Operation 5

Calvert Cliffs Unit 2 LTOP Enable Temperatures 31 12 EFPY i

+

k i

Page 3-

1

.537. doc (9033)/kr-4 i

LIST OF FIGURES HQ 1111I IEE 1

'0 i t alvert Cliffs Unit 2 Beltline P-T 32 o. ' 5, 12 EFPY Heatup 2

si Calvert Cliffs Unit 2 Beltline P-T 33 e

Limits, 12 EFPY Heatup 3

BGLE Calvert Cliffs Unit 2 Beltline P-T 34 Limits, 12 EFPY Cooldown 4

BG&E Calvert Cliffs Unit 2 Beltline P-T 35 Limits, 12 EFPY Cooldown 5

BG&E Calvert Cliffs Unit 2 Beltline P-T 36 Limits, 12 EFPY Hydrostatic 6

BG&E Calvert Cliffs Unit 2 Beltline P-T 37 Limits, 12 EFPY Heatup 7

BG&E Calvert Cliffs Unit 2 Beltline P-T 38 Limits, 12 EFPY Heatup 8

BG&E Calvert Cliffs Unit 2 Beltline P-T 39 Limits, 12 EFPY Cooldown 9

BG&E Calvert Cliffs Unit 2 Beltline P-T 40 Limits 12 EFPY Cooldown I

L L

Page 4

,537. doc (9033)/kr-5 4

1.0 INTRODUCTION

The following sections describe the basis for development of reactor vessel beltline pressure-temperature limitations for the Calvert Cliffs Unit 2 Nuclear Generating Station.

These limits are calculated to meet the regulations of 10 CFR Part 50 Appendix A,(I)

Design Criterion 14 and Design Criterion 31. These design criteria required that the reactor coolant pressure boundary be designed, fabricated, erected, and tested in order to have an extremely low i

probability of abnormal leakage, of rapid failure, and of gross rupture. The criteria also require that the reactor coolant pressure boundary be designed with sufficient margin to assure that when stressed under operating, maintenance, and testing the boundary behaves in a non-brittle manner and the probability of rapidly propagating fracture is minimized.

The pressure-temperature limits are developed using the requirements of 10 CFR 50 Appendix G(2)

This appendix describes the requirements for developing the pressure-temperature limits and provides the general basis for these limitations.

The margins of safety against fracture provided by the pressure-temperature limits using the requirements of 10 CFR Part 50 Appendix G are equivalent to those recommended in the ASME Boiler and Pressure Vessel Code Section III, Appendix G, " Protection Against Nonductile Failure."(3)

The general guidance provided in those procedures has been utilized to develop the Calvert tilffs Unit 2 pressure-temperature limits with the requisite margins of safety for the heatup and cooldown conditions.

The-Reactor Pressure Vessel beltline pressure-temperature limits are based upon tiie irradiation damage prediction methods of Regulatory Guide l.99 Revision 02(4)

This methodology has been used to c-calculate the limiting material Adjusted Reference Temperatures for Calvert Cliffs Unit 2 and have utilized fluence values for 12 Effective Full Power Years (EFPY).

I Page 5 m

2

537. doc (9033)/kr-6 t

This report provides reactor vessel beltline pressure-temperature limits in accordance with 10 CFR 50 Appendix G for 12 EFPY.

The events analyzed for cooldown are the isothermal, 10, 20, 30, 40, 50, 75 and 100'F/hr conditions and for heatup are the 10, 20, 30, 40, 50, 60, 70 and 75'F/hr conditions. These conditions were analyzed to provide a data base of reactor vessel P-T limits for use ir, establishing low Temperature Overpressure Protection requirements.

Low Temperature Overpressure Protection (LTOP) enable temperatures are calculated based upon the guidance provided in USNRC Standard Review Plan (SRP) 5.2.2.(5) Using this guidance the temperatures at which the LTOP system must be aligned to the RCS under heatup and cooldown conditions are established for automatic protection of the Appendix G P-T limits.

2.0 ADJUSTED REFERENCE TEMPERATURE PROJECTIONS In order to develop pressure-temperature limits over the design life of the reactor vessel, adjusted reference temperatures (ART) for the controlling beltline material need to be determined. The adjusted reference temperatures of reactor vessel beltline materials for Calvert Cliffs Unit 2 have been calculated at the 1/4t and 3/4t locations after 12 EFPY of operation.

By comparing ART data for each material, the controlling material for Calvert Cliffs Unit 2 has been determined.

The adjusted reference temperatures (ART) have been calculated using the procedures in Regulatory Position 1.1 of Regulatory Guide 1.99 Revision 02(4)

The calculative procedure for the ART values for each material in the beltline is given by the following expression:

ART - Initial RTNDT + ARTNDT + Margin i

Page 6 i

537. doc (9033)/kr-7 i

Initial RT is the reference temperature for the unirradiated NDT material. ART is the mean value of the adjustment in the NDT reference temperature caused by irradiation and is given by the following expression:

NDT - (CF) f(0.28 - 0.10 log f)

ART CF is the chemistry factor for the beltline materials which is a function of residual element content, i.e., weight percent copper and nickel.

Regulatory Guide 1.99 Revision 02 provides values for the chemistry factors for welds and for base metal plates and forgings.

The term f is the neutron fluence at any depth in the vessel.

The neutron fluence at any depth is given by the following expression:

f-fsurf (*

)

The term f is the calculated value of the neutron fluence surf I

2 (10 'n/cm, E> IMeV) at the inner wetted surface of the vessel at

(

th'e location of the postulated defect (1/4t or 3/4t), and x is the l

depth into the vessel wall from the inner wetted surface in inches.

l Margin is the quantity that is added to obtain a conservative upper bound value of ART. The margin term is given by the following expression:

Margin - 2/ 2,7 i

1

  1. a The terms og and o represent the standard deviation for initial 3

RT and the standard deviation of the mean value for the reference NDT i

temperature shift.

l The following information provides the basis for the calculated ART

- values for Calvert Cliffs Unit 2:

Page 7 i

i

537. doc (9033)/kre8

/

I 1.

Material data were obtained from Reference 6, including copper content, nickel content and initial reference temperature (RTNDT). These data are summarized in Table 1 for Calvert Cliffs Unit 2.

2.

Peak neutron fluence for the Calvert Cliffs Unit 2 beltline region was determined to be 1,69 x 10 n/cm2 (E>l MeV) at 12 I9 EFPY (Reference 7).

3.

Shell course minimum reference thickness is 8.625 in. for both the lower and intermediate shells (Reference 9).

4.

Calculations were based on the procedures in Regulatory Position 1.1 of NRC Regulatory Guide 1.99, Rev. 2 (Reference 4). Uncertainty in initial RT was taken as O'F for measured NDT values and 17'F for welds without measured values (Reference 10).

Adjusted reference temperatures for all beltline materials at the 1/4t and 3/4t locations after 12 EFPY were calculated using Regulatory Guide 1.99 Revision 02 and the results of the calculation are listed in Table 2 for Calvert Cliffs Unit 2.

The controlling i

materials-are shown in Table 2; the term " controlling" means having the highest ART for a given time and position within the vessel wall. The highest, or limiting, ARTS are then used to develop the pressure-temperature limits for the corresponding time period.

.In the case of Calvert Cliffs Unit 2, the intormediate shell longitudinal welds are controlling at the 1/4t and 3/4t locations after 12 EFPY based on the predicted ART values of 171'F and 125'F respectively.

According to Position 1.1 of Regulatory Guide 1.99, Revision 2(4),

the uncertainty in the value of initial RT is to be estimated NDT from the precision of test method when a " measured" value of initial j

1 Page 8

537. doc (9033)/kr-9 RT is available.

RT is derived in accordance with NB2300 of NDT NDT the ASME Boiler and Pressure Vessel Code,Section III.

It involves both a series of drop weight (ASTM E208) and Charpy impact (ASTM E23) tests on the material. The RT resulting from this two test NDT method evaluation is conservatively biased. The elements of this conservatism include:

1)

Choice for RT is the higher of NDT or TCV -60'F.

The NDT drop-weight test is performed to obtain NDT and a full Charpy for a given material.

impact curve is developed to obtain TCV The combination of the two test methods gives protection against the possibility of errors in conducting either test and, with the full Charpy curve, demonstrates that the material is typical of reactor pressure vessel steel.

Choice of the more conservative of the two (i.e., the higher of NDTT or TCV ~

60*F) assures that tests at temperatures above the reference temperature will Jield increasing values of toughness, and verifies the temperature dependence of the fracture toughness curve (ASME Code, Section 111, Appendix G).

implicit in the KIR 2)

Selection of the most adverse Charpy results for T I"

CV' accordance with NB2300, a temperature, TCV, is established at which three Charpy specimens exhibit at least 35 mils lateral expansion and not less than 50 ft-lb absorbed energy.

The l

three specimens will typically exhibit a range of lateral expansion and absorbed energy consistent with the variables f

inherent in the test: Specimen temperature, testing equipment, operator, and test specimen (e.g., dimensional tolerance and I

material homogeneity).

All of these variables are controlled using process and procedural controls, calibration and operator l

training, and they are conservatively bounded by using the L

lowest measurement of the three specimens.

Furthermore, two related criteria are used, lateral expansion and absorbed energy, where consistency between the two measurements provides-1 Page 9

537. doc (9033)/kr-10 l

c.

further assurance that they are realistic and the material will exhibit the intended strength, ductility and toughness implicit in the KIR curve.

3)

Inherent conservatism in the protocol used in performing the drop-weight test.

The drop-weight test procedure was carefully designed to assure attainment of explicit values of deflection and stress concentration, eliminating a specific need to account for below nominal test conditions and thereby guaranteeing a conservative direction of these uncertainty components.

In addition, the test protocol calls for decreasing temperature until the first failure is encountered, followed by increasing the test temperature 10*F above the paint where the int failure is encountered.

This in fact assures that one has biased the resulting estimate toward a low failure probability region of the temperature versus failure rate function diagrammed below. The effect of this protocol is to conservatively accommodate the integrated uncertainty components, g

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i Given the three elements of conservatism described above, values of initial RT obtained in accordance with NB2300 will NDT result in a conservative measure of the reference temperature.

The conservative bias of the NB2300 methodology and the i

l i

Page.10

537. doc (9033)/kr-11 a

drop-weight test protocol essentially eliminate the uncertainty which might result from the precision of an individual drop-weight or Charpy impact test. Therefore, when measured values of RT are available, the estimate of uncertainty in NDT initial RT is taken as zero.

NDT 3.0 PRESSURE - TEMPERATURE LIMITS 3.1 GENERAL APPROACH FOR CALCULATING PRESSURE-TEMPERATURE LIMITS The analytical procedure for developing reactor vessel pressure-temperature limits utilizes the methods of Linear Elastic Fracture Mechanics (LEFM) found in the ASME Boiler and Pressure Vessel Code Section III, Appendix G (Reference 3) in accordance with the requirements of 10 CFR Part 50 Appendix G (Reference 2).

For these analyses, the Mode I (opening mode) stress intensity factors are used for the solution basis.

The general method utilizes Linear E ustic Fracture Mechanics procedures.

Linear Elastic Fracture Mechanics relates the size of a flaw with the allowable loading which precludes crack initiation.

This relation is based upon a mathematical stress analysis of the beltline material fracture toughness properties as prescribed in Appendix G to Section 111 of the ASME Code.

t e

The reactor vessel beltline region is analyzed assuming a semi-elliptical surface flaw oriented in se axial direction with a depth of one quarter of the reactor vessel beltline thickness and an aspect ratio of one to six.

This postulated flaw is analyzed at both the inside diameter location (referred to as the 1/4t location) and the outside diameter location (reierred to as the 3/4t location) to assure.the most limiting condition is achieved.

The above flaw geometry and orientation is the maximum postulated defect size (reference flaw) described in Appendix G to Section III of the ASME Code.

Page 11

537. doc (9033)/kr-12 j

t At each of the postulated flaw locations the Mode I stress intensity factor, K, produced by each of the specified loadings is calculated g

and the summation of the Kg values is compared to a reference stress intensity, KIR, which is the critical value of Kg for the material and temperature involved.

The result of this method is a relation of pressure versus temperature for each reactor vessel operating period which precludes brittle fracture. K is btained from a IR reference fracture toughness curve for low alloy reactor pressure vessel steels as defined in Appendix G to Section III of the ASME Code.

This governing curve is defined by the following expression:

(.0145(T-ART +160))

KIR = 26.78 + 1.223 e

where, reference stress intensity factor, Ksi /in K

IR temperature at the postulated crack tip, 'F T

adjusted reference nil ductility temperature at AR1 the postulated crack tip, 'F For any instant during the postulated heatup or cooldown, K i8 IR calculated using the metal temperature at the tip of the flaw, and using the value of adjusted reference temperature at that flaw location.- Also for any instant during the heatup or cooldown the temperature gradients across the reactor vessel wall are calculated (see Section 2.3) and the corresponding thermal stress intensity factor, KIT, is determined. Through the use of superposition, the thermal-stress intensity is subtracted from the available KIR to determine the allowable pressure stress intensity factor and-t consequently the allowable pressure..

Page 12

.537. doc (9033)/kr-13 e

i In accordance with the ASME Code Section III Appendix G J

requirements, the general equations for determining the allowable pressure for any assumed rate of temperature change during Service Level A and B operation are:

2Kgg + KIT < KIR 1.5Kgg + KIT < EIR (Inservice Hydrostatic Test)

where, Allowable pressure stress intensity factor, Ksi/in K

gg Thermal stress intensity factor, Ksi/in K

=

IT Reference stress intensity, Ksi/in K

IR The pressure-temperature limits provided in this report account for the temperature differential between the reactor vessel base metal and the reactor coolant bulk fluid temperature.

Correction for elevation and RCS flow induced pressure differences between the reactor vessel beltline and pressurizer, are provided by Reference 8 and are included in the development of.the pressure-temperature limits.

Consequently, the P-T limits are-provided on coordinates of pressurizer pessure versus indicated RCS temperature.

I The pressure correction factors are based upon the dii sential i

pressure due to the elevation difference between the reactor vessel beltline wall and the pressurizer..This term of the pressure l'

correction factor is equal.to -15.0 psi.

The pressure correction-factors are als: based upon flow induced pressure drops across the reactor core through the hot leg pipe up to the surge line nozzle.

This term of the pressure correction factor has two values which are I

dependent upon the Reactor Coolant. Pump (RCP) combination utilized during operation. During cooldown at temperatures of Tc > 150*F, Page 13

537 doc (9033)/kr-14 i

the flow induced plus static pressure drop is based upon the RCS flow rates resulting from four operating RCPs and elevation differences and is equal to -52.0 psi. During cooldown at i

temperatures of T 5 150'F, the flow induced pressure drop is zero e

based upon no operating RCPs.

During heatup for all RCS temperatures, the flow induced pressure drop is assumed to be that associated with operation of four reactor coolant pumps.

Instrument uncertainties are provided by Reference 7 and have been included in the pressure-temperature limits.

Consequently, two i

pressure correction factors are utilized in correcting the reactor vessel beltline region pressure to pressurizer pressure depending upon the cold leg temperature. The uncertainty associated with the pressure indication instrument loop is -48 psi and the uncertainty l

associated with the temperature indication instrument loop is +10*F.

The previous information was combined to develop the following pressure correction factors which have been utilized in the development of the P-T curves:

COOLDOWN HEATUP Tc ('F)

PRESSURE CORRECTION Tc ( F)

PRESSURE CORRECTION FACTOR (PSI)

FACTOR (ISI)

> 150*F

-100 psi All RCS

-100 psi s 150'F

-63 psi

Temp, i

l By explicitly accounting for the temperature differential between the flaw tip base metal temperature and the reactor coolant bulk fluid temperature, and t'ne pressure differential between the beltline region of the reactor vessel and the pressurizer, the P-T limits are correctly represented on coordinates of pressurizer pressure and cold leg temperature.

Page 14 i:

~

537. doc (9033)/kr-15 t

3.2 THERMAL ANALYSIS METHODOLOGY The Mode I thermal stress intensity factor is obtained through a detailed thermal analysis of the reactor vessel beltline wall using a computer code.

In this code a one dimensional three noded isoparametric finite element is used for performing the radial conduction-convection heat transfer analysis.

The vessel wall is divided into 24 elements and an accurate distribution of temperature as a function of radial location and transient time is calculated.

The code utilizes convective boundary condition on the inside wall of the vessel and an insulative boundary on the outside wall of the vessel.

Variation of material properties through the vessel wall are permitted allowing for the change in material thermal properties between the cladding and the base metal.

In general, the temperature distribution through the reactor vessel wall is governed by a partial differential equation, pC EI - K (A 1 + I EI]

at 2

r ar ar subject to the following boundary conditions at the Inside and outside wall surface locations:

At r = rj

-K h = h (T-T )

c At r = r, h

= 0-i

where, 3

density,1b/ft p

j --

specific heat, btu /lb 'F C

=

thermal conductivity, btu /hr-ft 'F K

=

vessel wall temperature, 'F T

=

radius, ft r-

=

time, hr t

=

Page 15 i

,537. doc (9033)/kr-16 a

l i

2 h

convective heat transfer coefficient, btu /hr-ft _.7 RCS coolant teroperature, 'F T

=

c rg,r,=

inside and outside.adii of vessel wall, ft The above is solved numerically using a finite element model to determine wall temperature as a function of radius, time, and thermal rate.

Thermal stress intensity factors are determined by the calculated temperature difference through the beltline wall using thermal influence coefficients specifically generated for this purpose.

The influence coefficients depend upon geometrical parameters associated with the maximum postulated defect, and the geometry of the reactor vessel beltline region (i.e., r /r, a/c, g j a/t), along with the assumed unit loading.

The thermal stress intensity factors are determined by the temperature difference and temperature profile through the beltline wall using thermal influence coefficients and superposition. ASME code Section !!! Appendix G recognizes the limitations of the method it provides for calculating K because of the assumed temperature g7 profile.

Since a detailed heat transfer analysis results in varying

. temperature profiles (and consequently varying thermal stresses), an alternate method for calculating K was employed as. required'by IT Article G-2214.3 of Reference 3.

The alternate method employed used a polynomial fit of the temperature profile and superposition using influence coefficients to calculate K The influence coefficients IT.

were calculated using a 2-dimensional finite element model of the reactor vessel.

The influence coefficients were corrected for 3 dimensional effects using ASTM Special Technical Publication 677 (Reference 11).

I L

3.3-COOLDOWN LIMIT ANALYSIS L

During cooldown, membrane and thermal bending stresses act together in tension at the reactor vessel inside wall.

This results in the pressure stress intensity factor, Kgg, and the thermal stress intensity factor, KIT, acting in unison creating a high stress intensity.

At the reactor vessel outside wall the tensile pressure Page 16

. 537 doc (9033)/kr-17 i

t J

l stress and the compressive thermal stress act in opposition resulting in a lower total stress than at the inside wall location.

Also neutron embrittlement, the shift in RT and the associated NDT reduction in fracture toughness are less severe at the outside wall compared to the inside wall location.

Consequently, the inside flaw location is more limiting and is analyzed for the cooldown event.

Utilizing the material metal temperature and adjusted reference temperature at the 1/4t location, the reference stress intensity is determined.

From the method provided in Section 2.3, the through wall temperature gradient is calculated for the assumed cooldown rate to determine the thermal stress intensity factor.

In general, the thermal stress intensity factors are found using the temperature difference through the wall as a function of transient time as described in Section 2.3.

They are then subtracted from the available K value to find the allowable pressure stress intensity IR factor and consequently the allowable pressure.

The cooldown pressure temperature curves are thus generated by calculating the allowable pressure on the reference flaw at the 1/4t location based upon 9

Kip, KIT K

jg 2

where, t

Allowable pressure stress intensity as a function of K

gg coolant temperature, Ksi/in Reference stress intensity as a function of coolant K

IR temperature, Ksi/in 1

Thermal stress intensity as a function of coolant K

=

IT L

temperature, Ksi/in To develop a composite pressure-temperature limit for the cooldown event, the isothermal pressure-temperature limit must be calculated.

l Page.17

537. doc (9033)/kr-18 The isothermal pressure-temperature limit is then compared to the l

pressure-temperature limit associated with a cooling rate and the mori restrictive allowable pressure-temperature limit is chosen resulting in a composite limit curve for the reactor vessel

beltline, i

1u i provides the results for the isothermal, 10, 20, 30, 40, 50, 75 and 100'F/hr cooidown pressure-temperature limits.

These tables provide the allowable pressure versus reactor coolant temperature for the various cooldown conditions. The allowable pressure is in units of ksi while the temperature is in units of 'F.

Figures 3, 4, 8 and 9 provide a graphical presentation of the cooldown pressure-temperature limits found in Table 3.

It is permissible to linearly l

interpolate between the cooldown pressure-temperature limits.

l 3.4 HEATUP LIMIT ANALYSIS During a heatup transient, the thermal bending stress is compressive at the reactor vessel inside wall and is tensile at the reactor.

vessel outside wall.

Internal pressure creates a tensile stress at the inside wall as well as the outside wall locations.

Consequently, the outside wall location has the larger total stress I

when compared to the inside wall.

However, neutron embrittlement (the shift in material RT and the associated reduction in NDT fracture toughness) is greater at the inside location than the outside.

Therefore, both the inside and outside flaw locations must be analyzed to assure that the most limiting condition is achieved.

l As described ~in the cooldown case, the reference stress intensity l-factor is calculated by the metal temperature at the tip of the flaw and by the adjusted reference temperature at the flaw location.

For heatup the reference stress intensity is calculated for both the 1/4t and 3/4t locations.

Using the finite element method described in Section 2.3, the temperature profile through the wall and the Page 18

.537. doc (9033)/kr-19 i

metal temperatures at the tip of the flaw are calculated for the transient history. This information is used to calculate the thermal stress intensity factor at the 1/4t and 3/'d locations using the calculated wall gradient and thermal influence coefficients.

The allowable pressure stress intensity is then determined by superposition of the thermal stress intensity factor with the available reference stress intensity at the flaw tip.

The allowable pressure is then derived from the calculated allowable pressure stress intensity factor.

It is interesting to note that a sign change occurs in the thermal stress through the reactor vessel beltline wall. Assuming a reference flaw at the 1/4t location the thermal stress tends to alleviate the pressure stress indicating the isothermal steady state condition would represent the limiting P-T limit.

However, the isothermal condition may not always provide the limiting i

pressure-temperature limit for the 1/4t location during a heatup transient.

This is due to the correction of the base metal temperature to the Reactor Coolant System (RCS) fluid temperature at the inside wall by accounting for clad and film temperature differentials, for a given heatup rate (non-isothermal), the differential tempera-ture through the clad and film increases as a function of thermal rate resulting in a higher RCS fluid temperature at the inside wall than the isothermal condition for the same flaw tip temperature and pressure.

Therefore to ensure the accurate representation of the 1/4t pressure-temperature limit during heatup, both the isothermal and heatup rate dependent pressure-temperature limits are calculated to ensure.the limiting condition was achieved. These limits account for clad and film differential temperatures and for the gradual buildup of wall differential temperatures with time, as do the cooldown limits.

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.53{. doc (9033)/kr-20 t

4 At the 3/4t location the pressure stress and thermal stresses are tensile resulting in the maximum stress at that location.

Pressure-temperature limits were calculated for the 3/4t location accounting for clad and film differential temperature and the buildup of wall temperature gradients with time using the method described in Section 2.3.

The allowable pressure was derived based upon a flaw j

at the 3/4t location by superposition of the thermal stress intensity with the available reference stress intensity for the metal temperature and adjusted reference temperature at that position.

To develop composite pressure-temperature limits for the heatup transient, the isothermal, 1/4t heatup, and 3/4t heatup pressure-temperature limits are compared for a given thermal rate.

Then the most restrictive pressure-temperature limits are combined over the complete temperature interval resulting in a composite limit curve for the reactor vessel beltline for the heatup event.

Table 3 provides the results for the 10, 20, 30,- 40, 50, 60, 70 and 75'F/hr heatup pressure-temperature limits. These tables provide the allowable pressure versus reactor coolant temperature for the various heatup conditions. The allowable pressure is in units of

-ksi while the temperature is in units of 'F.

Figures 1, 2, 6 and 7, provide a graphical presentation of the heatup pressure-temperature limits found in Table 3.. It is permissible to linearly interpolate between the heatup pressure-temperature limits.

3.5 HYDROSTATIC TEST AND CORE CRITICAL LIMIT ANALYSIS Both 10 CFR Part 50 Appendix G and the ASME Code Appendix G require the development of pressure-temperature limits which are applicable

)

7' to inservice hydrostatic tests.

For hydrostatic tests performed subsequent to loading fuel into the reactor vessel, the minimum test temperature is determined by evaluating K, the mode I stress g

intensity factors. The evaluation of K is performed in the same g

Page 20

537. doc (9033)/kr-21 j

manner as that for normal operation heatup and cooldown conditions except the factor of safety applied to the pressure stress intensity factor is 1.5 versus 2.0.

From this evaluation, a pressure-temperature limit which is applicable to inservice hydrostatic tests is established.

The minimum temperature for the inservice hydrostatic test pressure can be determined by entering the curve at the test pressure (1.1 times normal operating pressure) and locating the corresponding temperature.

The inservice hydrostatic test limit is provided for 12 EFPY in Table 4 and is shown in Figure 5.

Appendix G to 10 CFR Part 50, specifies pressure-temperature limits for core critical operation to provide additional margin during actual power operation. The pressure-temperature limit for core critical operation is based upon two criteria.

These criteria are that the reactor vessel must be at a temperature equal to or greater than the minimum temperature required for the inservice hydrostatic test, and be at least 40*F higher than the minimum pressure-temperature curve for normal operation heatup or cooldown.

Note, that the core critical limits established above are solely based upon fracture mechanics considerations, and do not consider l

core reactivity safety analyses which can control the temperature at l

which the core can be brought critical.

L 3.6

_ LOWEST SERVICE TEMPERATURE, MINIMVM BOLTUP TEMPERATURE, AND MINIMVM PRESSURE LIMITS

)

In addition to the computation'of the reactor vessel beltline P-T g

l limits, additional limits have been provided for reference.

These L

additional limits are the Lowest Service Temperature, Minimum Boltup Temperature, and Minimum Pressure Limits. These limits are described below.

2

(

a Page.21

.537. doc (9033)/kr-22 i.

t The Lowest Service Temperature is the minimum allowable temperature at pressures above 20% of the pre-operational system hydrostatic test pressure (625 psia). This temperature is defined as equal to the most limiting RT for the balance of Reactor Coolant System NDT (RCS) components plus 100'F, per Article NB 2332 of Section III of the ASME Boiler and Pressure Vessel Code.

The maximum RT for the balance of the RCS components was NDT conservatively estimated as 50'F.

Therefore, the Lowest Service Temperature is equal to 100'F + 50'F + 10'F - 160'F.

The minimum pressure limit is the break point between the minimum boltup temperature and the lowest Service Temperature.

Defined by the ASME Boiler and Pressure Vessel Code as 20% of the pre-operational hydrostatic test pressure, the minimum pressure is as follows when pressure correction factors for elevation and flow are taken into account:

Cooldown Heatuo l

'562 psia T s 150'F 525 psia All RCS e

l 525 psia Tc > 150'F Temperatures r

The minimum boltuo temperature is the minimum allowable temperature at pressures below the 20% of the pre-operational system hydrostatic test pressure. The minimum is defined as the initial RT for the NDT material of the higher stressed region of the reactor vessel plus.

any effects for_1rradiation per Article G-2222 of Section 111 of the U

ASME Boiler and Pressure Vessel Code.

The init's1 reference temperature'of the reactor vessel and closure head flanges was determined using the certified n.aterial test reports and Branch Technical Position MTEB 5-2.

The maximum initial RT associated NDT with the-stressed region of the closure head flange is 30'F. The Page 22-

537. doc (9033)/kr-23 I

minimum boltup temperature including temperature instrument uncertainty is 30*F + 10'F 40'F.

However, for conservatism a minimum boltup temperature of 70*F is utilized.

1 4.0 LTOP ENABLE TEMPERATURES Standard Review Plan 5.2.2, Overpressure Protection (5), has defined the temperature at which the Low Temperature Overpressure Protection (LTOP) system should be operable during startup and shutdown conditions.

This temperature known as the LTOP enable temperature is defined as the water temperature corresponding to a metal temperature of at least RTNDT + 90*F at the beltline location (1/4t or 3/4t) that is controlling in the Appendix G calculations.

Below the LTOP enable temperaturis the LTOP system must be aligned to the RCS to prevent exceeding ths; applicable technical specification and Appendix G limits in the event of a transient.

Based upon this definition and upon the results of the Appendix G l

calculations, LTOP enable temperatures for cooldown and heatup l'

conditions have been calculated for Calvert Cliffs Unit 2.

In l

addition, LTOP enable temperatures have been calculated using l

specific heatup transients with changing thermal rates. The LTOP l

enable temperatures associated with the heatup transients which utilize changing thermal rates credit soak times for reducing thermal stress and meet the criteria for the LTOP enable temperatures as defined in Standard Review Plan 5.2.2.

L The LTOP enable temperature for a cooldown is based upon the isotiermal P-T limit.

Consequently the LTOP enable temperature is equel to the 1/4t adjusted reference temperature + 90'F.

Therefore for cooldown the LTOP enable temperatures equal 271'F for 12 EFPY whrn the temperature instrumentation uncertainty of 10*F is in:1uded.

The LTOP enable temperatures for heatup are provided in Table 5.

Page 23

537. doc (9033)/kr-24 5.0 DAIA i

Reactor Vessel Data Reference Design Pressure 2500 psia 8

Design Temperature 650'F 8

Operating Pressure 2250 psia 8

Beltline Thickness 8.625 in 8

6 Inside Radius (to wetted surface) 86.9028 in 8

=

Cladding Thickness

.3125 in 8

=

Material-SA 533 Grade B Class 1 Reference Thermal Conductivity 23.8 BTU /hr-ft 'F 11 6

Youngs Modulus 28 x 10 psi 11

=

Coefficient of Thermal 7.77 x 10-61n/in/'F 11

+

Expansion

.122 BTU /lb *F Specific Heat

=

3

.283 lb/in Density

=

Stainless Steel Claddirig Thermal Conductivity 10.1 BTU /hr-ft 'F 11 0.126 B+W/16 'F Specific Heat

=

3 0.285 lb/in Density Adiusted Reference Temoerature Values 12 EFPY Ref.trence 1/4t 171.0'F 3/4t.

125.0*F II 2

r Fast Neutron Fluence.= 1,69 x 10 n/cm 7

2 Film coefficient on inside surface = 1000 BTU /hr-ft,.p.

r Page 24

.537. doc (9033)/kr-25 i

1 Pressure Correction Factors For Elevation and Flow (Reference 7)

Cooldown Heatuo RCS temperature s 150'F dp = -63 psia All RCS dp = -100 psia RCS temperature > 150'F dp = -100 psia Temperatures Temoerature Instrument Correction (Reference 7) dt = +10*F l.

6.0 REFERENCES

1.

Code of Federal Regulations,10 CFR Part 50, Appendix A,

" General Design Criteria for Nuclear Power Plants",

January 1988.

i 2.

Code of Federal Regulations, 10 CFR Part 50, Appendix G

" Fracture Toughness Requirements", January 1988.

3.

ASME Boiler and Pressure Vessel Code Section !!!,

Appendix G, " Protection Against Nonductile Failure", 1986 Edition.

4.

' Regulatory Guide 1.99, " Radiation Embrittlement of Reactor Vessel Materials", U.5 Nuclear Regulatory Commission, Revision 2, May 1938.

5.

U. S. Nuclear Regulatory Cammission Standard Review Plan

]

(SRP) 5.2.2, Overpressure Protection, Revision 2, November 1988.

at End of j

6.

D. A. Wright, " Assumptions and Calculations of RTPTS l

License for Materials in Calvert Cliffs Units 1 and 2 Reactor Vessel Beltline Region," Baltimore Gas & Electric Co., ME&Au Job Number MR-85-170, December 23, 1985.

Page 25

537. doc (9033)/kr-26

.o i

7.

W. R. Boyd to Paul Hijeck, " Transmittal of Data for PT Curve Calculation", BGLE letter NEU-90-427, May 24,1990.

8.

W. R. Boyd to Paul Hijeck, " Transmittal of Data for P-T Curve Calculation," BG&E letter NEU-90-442, June 5,1990, 9.

Instruction Manual, Reactor Vessel Assembly, Calvert Cliffs Station, Baltimore Gas & Electric Co., C-E Book No. 72167/73167 June 1972.

10.

Evaluation of Pressurized Thermal Shock Effects Due to Small Break LOCA with loss of feedwater for the C-E NSSS, CEN-189, December 1981.

11.

" Semi-Elliptical Cracks in a Cylinder Subjected to Stress Gradients", J. Heliot, R. C. Labbens, and Pellisser-Tanon, ASTM Special Technical Publication 677, August 1979.

12. ASME Boiler and Pressure Vessel Code Section !!!, Appendix 1,

" Design Stress Intensity Values Allowable Stresses, Material L

Properties, and Design Fatigue Curves", 1986 Edition.

i i

l f

Page 26

.537. doc (9033)/kr-27 t

Table 1 CALVERT CLIFFS UNIT 2 REACTOR VESSEL BELTLINE MATERIALS Identification RT Component Number (giX1 M1111 NDT Intermediate Shell 2-203-A,B,C 0.12 1.01

-56'Fil) long. Welds lower Shell 3-203-A,B,C 0.23 0.23

-80*F Long. Welds Inter./ Lower Shell 9-203 0.22 0.05

-60'F Girth Weld Intermediate Shell D-8906-1 0.15 0.56 10'F Places D-8906-2 0.11 0.56 10'F 0-8906-3 0.14 0.55 5'F Lower Shell D-8907-1 0.15 0.60

-8'F Platet D-8907-2 0.14 0.66 10*F D-8907-3 0.11 0.74 16'F (1)

Generic value for submerged are welds 4

i Page 27

537. doc (9033)/kr-28 Table 2 CALVERT CLIFFS UNIT 2 BELTLINE ART CALCULATION Identification Chemistry Initial 1/4 Thickness 3/4 Thickness RTNDT, F A RTET, F ART.*F RTET' I MT.

  • F Number

' Factor 203-A,B,C 161

-56 17 28 161 171 115 125 3-203-A,B,C 120.

-80 0

28 121 97 86 62 9-203 101

-60 0

28 101 97 72 68 D-8906-1 108 10 0

17 108 152 77 121 D-8906-2 74 10 0

17 74 118 53 97 i

D-8906-3 98 5

0 17 98 137 70 109 0-8907-1 110

-8 0

17 l'.3 136 79 105 D-8907-2 102 10 0

17 102 146 73 117 D-8907-3 76

-16 0

17 76 94 55 73 5

E M-t

...__~1__._

m__,

I

1 TABLE 3 8 gee CAL 9ERT CL1FFS 1511T 2 P-ALLthstBLE (KSI) WS. ats TEleERATURE (OEG. F)

FWt 12 EFPT, WINWWil IFER6f!ON CEOL9ebSI NEATUP P-ALLthsteLE (KSI)

P-ALLtRetSLE (KSI)

RCS

=

TEM ISO 10 F/

20 F/

30 F/

40 F/

50 F/

60 F/

70 F/

75 F/

ISO 10 F/

20 F/

30 F/

40 F/

50 FI 75 F/ 1es F/

DEG F TNEAMAL - Netst NOUR WOUR NOUR NOUR Neut 40lst notat THEIBett WOUR WOUR neur Webst neut usuR meut 0.4M8 0.4356 0.3973 0.3592 0.3213 0.2857 8.1968 0.9999 0.4803 8.4422 9.4864 8.3668 8.3295 0.2*M 8.2911 S.1122 60 0.4370 9.4875 0.4499 S.4126 0.37 % 0.3399 0.3025 0.2130 8.1264

70. 0.4433 90 0.4588 0.4588 0.4588 0.4588 0.4588 0.4588 0.4588 6.4588 0.4588 0.4958 0.4588 8.4222 0.3858 8.3498 0.3141 0.2268 8.142T 80 0.4505 100 0.4486 0.4686 0.4686 0.4686 0.4686 0.4686 0.4686 0.4686 8.4686 0.5054 8.4691 9.4332 0.3976 0.M23 0.3276 S.2627 8.M15 110 0.4796 0.4796 0.4796 8.4796 0.4796 0.4719 0.4667 0.4639 0.4631 0.5166 0.4810 8.4459 0.4112 0.3769 0.3431 s.2610 0.1831 120 0.4924 0.4924 0.4926 0.49M 0.4782 0.4635 0.4532 9.4463 0.4438 0.5296 9.4968 0.4686 0.42e9 0.3937 0.3611 S.28M e.2985 130 0.5073 0.5073 0.5073 0.5073 8.4863 0.4630 0.4473 0.4359 0.4311 0.5443 8.5107 0.4776 0.4451 0.4131 0.38t9 0.3869 0.2378 140 0.5245 0.5245 0.5245 0.5265 0.4973 0.4696 0.4482 0.4319 0.4254 8.M15 0.5291 8.4972 0.4661 0.43 % 0.4059 8.33 % G.2n4

-150 0.5444 0.5444 9.5444 0.5444 8.5165 0.4825 0.4554 0.4339 0.4252 0.5814 0.5503 8.5208 0.4906 0.46 4 0.433T 0.3683 8.3102 160 9.M73 0.M 73 0.%T3

0. % 73 0.5418 0.5014 0.4685 0.4420 0.4308 0.MT3 0.5379 0.5882 0.4814 0.4546 0.6287 8.3600 8.3177 170 0.5939 0.5939 0.5939 0.5939 8.5731 0.5264 0.4875 8.4557 0.4421 0.5939 0.5663 8.5306 0.5139 0.4893 0.4668 9.4132 8.3708 180 0.6266 0.6246 0.6246 0.6266 0.6108 0.5574 0.5126 0.4749 0.4586 0.62 %

0.5991 0.574T 0.5514 0.5296 0.50e8 c.4638 e.4382 190 0.6601 0.6601 0.6601 0.6601 0.65 % 0.5954 0.5439 0.5005 0.48 4 0.6601 8.63T1 s.6153 0.5968 0.5758 0.5585 0.5225 0.4995 200 0.7012 0.7012 0.7912. 0.7012 0.7012 0.6406 0.5822 0.5323 0.5105 0.7912 0.6810 e.6622 0.6450 0.4296 0.6159 0.5905 S.5792 210 0.7486 0.M86 0.M86 8.M86 0.7486 0.6939 0.6288 8.5711 e.5462 0.M86 9.731T 3.7164 0.7030 0.6914 0.6829 0.6683 9.6708 220 0.8035 0.8035 0.8835 0.8835 0.8035 9.7565 0.6822 9.6175 8.5896 8.8035 0.7983 0.T791 0.7700 0.7631 0.7588 8.7595 8.7783 230 0.8669 0.8669 0.8669 0.8669 8.8689 0.8292 0.7459 8.6730 9.6485 0.86ee S.8581 8.85 4 0.8675 0.8660 8.86M 0.8639 8.8609 240 0.9602 0.9602 0.9602 0.9402 0.9602 0.9143 0.8203 0.7382 0.7017 9.9602 8.9MS 8.9354 0.93T1 8.9602 0.96el 0.9602 9.9582 250 1.0249 1.0249 1.8249 1.8249 1.0249 1.0126 8.9971 8.8141 S.7730 1.0269 1.0249 1.9249 1.0249 1.8249 1.8249 1.8269 1.0249 260 1.1229 1.1172 1.1145 1.1147 1.11M 1.1226 1.0000 0.9031 0.8564 1.1229 1.1229 1.1229 1.1229 1.1229 1.1229 1.1229 f.1229 270 1.2361 1.2231 1.21M 1,2073 1.2062 1.2060 1.1252 1.0069 0.9548 f.2361 1.2361 1.2361 1.2M1 1.2M1 1.2M1 1.2361 1.2361 280 1.M70 1.3655 1.3281 1.3145 1.3064 1.2976 1.2609 1.1269 1.e664 1.M79 1.3670 1.M79 1.3679 1.M79 1.M79 1.M79 1.3679 290 1.5186 1.4871 1.4605 1.4383 ~1.4202 1.4864. 1.3954 1.2658 1.1987 1.5186 1.5186 1.5186 1.5186 1.5186 1.5186 1.5186 1.5106 300 1.6933 1.6507 1.61M 1.5815 1.5541 1.53 %

1.512T 1.4282 1.3588 1.6933 1.8933 1.6933 1.8933 1.8953 1.6933 1.8953 1.6833 310 1.8956' 1.8399 1.7905 1.M70 1.7089 1.6764 1.6483 1.61 %

1.5274 1.89M 1.3956 1.89M 1.89%

1.8956 1.9956 1.89M 1.8956 320 2.1296 2.0586 1.9951 1.9383 1.8879 1.8648 1.8069 1.7T19 1.7325 2.1296 2.1296 2.1296 2.1296 2.1296 2.1296 2.1296 2.1296 330 2.3997 2.3114 2.2315 2.1595 2.0967 2.9370 1.9959 1.96 M 1.9212 2.3MT 2.3997 2.3997 2.3997 2.3997 2.399T 2.399T 2.399T 340 2.7122 2.603T 2.5869 2.4152 2.3339 2.26M 2.1952 2.1380 2.1121 2.7122 2.7122 2.7122 7.7122 2.7122 2.7122 2.7122 2.7122 350 2.9000 2.9000 2.8209 2.7107 2.6106 2.5198 2.4379 2.3660 2.3387 2.9000 2.9000 2.9000 2.9006 2.9000 2.9000 2.9000 2.9000 360 2.9000 2.9000 2.9000 2.8188 2.7166 2.6247 2.5839 2.9000 2.9800 2.9000 2.8773 2.9000 370 380 CXMAECTION FAC10RS:

g CINtRECTION FACTORS:

COOL 90bSI: TEW ERAllpE 10*F to ip NEATUP: TEWERATURE 10*F PRESSURE Tc 5 150*F DELTA-P * - 63 PSIA PHES5URE Tc < 558*F DELTA-P

-100 PSIA Tc > 150*F SELTA-P

  • -100 PSIA m

to 1

1

.______.u__

i

e.

.l

  • F TA8LE 4 8G&E CALVERT CLIFFS UNIT 2 P ALLmdA8LE (KSI) VS. RCS TEMP (DEG. F)

FOR 12 EFPT, NYDRDSTATIC OPERATION 1

RCS TEMP MYDR0$fAfic DEG F (KSI) 60 0.6161 70 0.6244 B0 0.6340 90 0.6451 1 00 0.657?

110 0.6727 120 0.6899 130 0.7097

'140 0.7327 150 0.7592 160 0.7898 1 70 0.8252 160 0.8662 190 0.9135 200 0.9682 l

210 1.0315 220 1.1046 L

230 1.1892 240 1.2869 lt 250 1.3999 260 1.5305 270 1.6815-280 1.8560 290 2.0578 300 2.2911.

310 2.5608 320 2.8726 330

.2.9000.

i t

  • j i

CORRECT 10W FACTORS:

J TEMPERATURE 10'F h

PRES 5URE Tc:< 550*F DELTA P e 100 PSIA

/

c

);

q 5 't. i f

"di

  • !i'; s.

's :

i f.

i

'ilI

(

j ?. ).

?!-

f 4

(

'Page ??'

I-Wl1 o

,+

537. doc (9033)/kr-31 Table 5 t

CALVERT CLIFFS 2 LTOP ENABLE TEMPERATURES (INCLUDING TEMPERATURE INSTRUMENT UNCERTAINTY) 12 EFPY Cooldown'LTOP Enable Temperature equals 271'F for all rates.

Enable Enable Heatup Limiting Temp Limiting Temp Rate _

Location 1/4t Location 3/4t 10'F/hr 1/4t 276'F 1/4t 233*F 20'F/hr 1/4t-280*F 1/4t 241*F 30'F/hr 1/4t 285'F 1/4t 249'F 40'F/hr 1/4t 289'F 1/4t 257'F 50*F/hr 1/4t 294*F 1/4t 265'F 60*F/hr 1/4t 299'F 1/4t-272'F 70'F/hr=

1/4t.-

303*F 1/4t 280'F-75'F/hr 1/4t 305'F 1/4t 283'F i

i

  1. s 4

,i,'

p;;

i

\\

~\\;

f l,'

a.'

.n

, ;\\;

-l

i. ' " -

4 t

.i.

?

~~.

g gt s

$ l1 Ph,.,

. Page 31 w j,gn n

. /[ {0j. M ',

s 1 1

.,a I-F!GURE 1 BG&E CALVERT CLIFFS UNIT 2 BELTLINE P-T LIMITS,12 EFPY HEATUP 2'500 107N Nf [

307/HR 507/HR U) i 707/HR N1 Iffl 2,000 r

i

/

y 1,500 g

I 1,000)

~

t

~

i:

l, J80,107/HR,307/HR

. V

/

JA 500-

-T n

507/HR 707/HR b'

j L

?

0~

a p

'50 100.-

1150 200 250

'300 350 400-

-450 l

f INDICATED RCS TEMPERATURE Tc, DEG.' F' f'

Tc < 86&F d.=

100 pele l

j4 '

AT = + 10.0T '

,g, g,g, 3/44 = 125.1F 1

'ka'

.Page 32 m

. l w'-

1

,4

.i i

si I :" i u

)

z'

'.~

'J

_.__;2.____

FIGURE 2 BGAE CALVERT CLIFF 8 UNIT 2 BELTLINE P-T LIMITS,12 EFPY

.)

HEATUP 2'500 spN 40TMR C

007MR IIU 7s7MR 2,000 1,500 f

[

l :1,000

/

ISO. 207HR -

~

3

,/

E

/TL%

007MR 40'rMR --

o 757MR o

a of;

"-(L. '

10 l.J,

50.

...100 - 150-200'

.250 300-350

'400-

'450 3

INDICATED RCS TEMPERATURE Tc, DEG. F -

I E-'

s>

t

.. ~,

Tc < 86# # = '.100 pela g g, 979,g.,

~ AT = + 10.07 =

3/4t = 125.07 K'

i f

. l, Page 33-l;p if

@%3g

>;n o

'q.

Uiit jijyc in i

I a'

. l

\\

FIGURE 3 BO&E CALVERT CLIFFS UNIT 2 BELTLINE P-T LIMITS,12 EFPY COOLDOWN 2,500 2,000 1,500 1,000 iso 1FFm srFm serm kG

/

500 g y

  1. 1 4

/y g

-d a

./

0 50; 100:

150 200 250-

-300' 350

-400-450 x

. INDICATED RCS TEMPERATURE Tc, DEG. F :

Tc.s 18dF d =..Se pela m.

~

Tc > 1seF e =.100 pela jj,, 373,yp u

tT = +10,fP<

a/4 = 1as.vP

.m

+

.Page 34 2

W-l',y:{ ',' ' _ j_ i Y I.

I,'I E _.

_ y

>i

. I FIGURE 4 BGAE CALVERT CLIFFS UNIT 2 BELTLINE P-T LIMITS,12 EFPY COOLDOWN 2,500 2,000 1,500 l

1,000 l;

lSO l

207/HR 407/HR f

1007/HR i

~

y j

500

,'. /

/~.-sr 4

/

J f

u

q-

>>ii

.,ii

,e ii ii

,i

,, i,-

j +g g

i>

50 100 150 200.

-250 300 350 400 450-INDICATED RCSTEMPERATURE Tc, DEG. F s

Tc.s 1818 9 =.63 pela g

Tc > 15# # =.100 pela p,,73,3 AT = +10.W 3/4t = 125.07 l

.hg

. ;it' r

,4 i

-: y,,

.Page 35 f,

  • la;r o,; ;,

mn n

.60

.1-

.I FIGURE 5 BG&E CALVERT CLIFFS UNIT 2 BELTLINE P-T LIMITS i

12 EFPY, HYDROSTATIC 2,500 2,000 l

g E

1,500

[

I g

E 1,000 u.

/

p a

500 m

y

't

,c, f

- 0 m

l.,,

> 50

'100 150 200~

250 500

'350 400 INDICATED FCS TEMPERATURE Tc, DEG. F-m.

Tc < aser s - 100 pela 3j,, iyi,e,

AT - +1a.0 aim - tas. O Page 36' H'

s

' 4. 2

(
y 4,

is

',,I }

4',,'

. j

\\

1 FIGURE 6 BG&E CALVERT CLIFFS UNIT 2 1

BELTLINE P-T LIMITS,12 EFPY l

HEATUP 1

1,000

//

~

950

//

~

/

900 :

850 r

j 800. !

I J

/

l L

750 :

L L

700 :

I600

~

650 i

// /

r 1

r 55n j

iso 1ove.seym 500.

n 450

/

5

_y s

y

.400- :

L e

350 o:

s.

300:

J 50:

756 :100. 125 150 ;175 200,2251 250 275 300 l

. INDICATED RCS TEMPERATURE Tc, DEG, F Tc < 88# aP = 100 pela AT = +10.#

g,

,p 4

3/44 = 125.#

.y }

Page 37

.i

,,9q.

, i

..k [

. Ji E

' ' I,5 f,' ' ~

,,3

f FIGURE 7 BG&E CALVERT CLIFFS UNIT 2 BELTLINE P-T LIMITS,12 EFPY HEATUP 1,000 _

950 l ll

/ //

i i

/ / /

850

=

g

/ / /

2 g

800 7 j)g l

/

750

/ //

700

/ / /

i

'650-

/ / /

E i

/)-/ /

}

ISO,27FNR

[

r i ew 450 _.g x

y L400.

L

350 2 1
3 00 c

50

-751 -100 -125 150 175 200 225 250: 275-300

< INDICATED RCS TEMPERATURE Tc, DEG. F fu 1/4t=

1,#

p.

3/44 = 128,W tip

,l Page 38 w;, ' 1:

.c-.

i it y

t FIGURE 8 BG&E CALVERT CLIFFS UNIT 2 BELTLINE P-T LIMITS,12 EFPY COOLDOWN 900

/

=

850 :

[

800 :

4

=

750 :

EN 700 5

Yf i

l

/////

i H//h L

550 i

//

/g/,/

v 500 _-

450 ;_ snw

/g,/

7 p

/

/

/s/

e F

'400 y

_m m7 7

j 350 300 somy

/

5 q

250 i"Y Lm

.200

~

i

,+

-150 :

o 1

m 100 :

^

,:r '

E 50 :

1 1

b/ '

-Oi 50' 75

100 125 150. 175 200.225 250 275 300 94

' ' ~

lND!CATED RCS TEMPERATURE Tc, DEG. F

,J

}(;'? %

Te s ts e e =.ss pela

_i Tc > ts# # =.i00 pela g, 37g,p K

AT = + 10.#

aMt=125.3 n_ :

<t l.',.'%'y,g.

Page.39 (M(; }iff, * '

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a-

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FIGURE 9 BG&E CALVERT CLIFFS UNIT 2 BELTLINE P-T LIMITS,12 EPPY COOLDOWN 900 _

}

=

850 :

l'{

800 :

=

//

750 :

[:

i

///

//9 8"

////

l*****

[

M//l l500 i

/

d//

. go,,e 7

/

450' y

l l

E mi m

'400. : '

e

/

/

o

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  1. "g'

<350. p

  • u 300: :

/

250- :

/

200 !

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=

.150 ;,ony' 100. !#

g; JJ 50 I 1

-01

~ ' ' ' ' '

i

) 50' 75 100 125 150- 175' 200' 225 250 275-300 L

INDICATED RCS TEMPERATURE Tc, DEG. F.

Tc.g ISO d = 63 pela g

+

Tc > iSH # = 100 pela W=m.@

Mi AT = +10.@ ~

3/44 = 125.#

- > _ i lll li I N'

Page 40

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