ML20081K982

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Final Rept on Reactor Vessel App G Pressure-Temp Limits for Arkansas Nuclear One Unit 2 for 21 Efpys
ML20081K982
Person / Time
Site: Arkansas Nuclear Entergy icon.png
Issue date: 05/31/1991
From:
ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY
To:
Shared Package
ML20081K976 List:
References
A-MPS-ER-002, A-MPS-ER-2, NUDOCS 9107020257
Download: ML20081K982 (53)
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ARKANSAS NUCLEAR ONE UNIT 2 RUSSELLVILLE, ARKANSAS Dy ABB COMBUSTION ENGINEERING NUCLEAR POWER COMBUSTION ENGINEERING, INC.

REACTOR VESSEL INTEGRITY GROUP 1000 PROSPECT HILL ROAD WINDSOR, CONNECTICUT 06095-0500 MAY 1991 A-MPS-ER-002 COMBUSTIONbENGINEERING '-

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TABLE OF CONTENTS SECTION- TITLE PAGE

1.0 INTRODUCTION

5 2.0 ADJUSTED REFERENCE TEMPERATURE 6 PROJECTIONS 3.0 GENERAL APPROACH FOR CALCULATING 11 PRESSURE-TEMPERATURE LIMITS 3.1 Thermal Analysis Methodology 14 3.2 Heatup Limit Analysis 16 3.3 Cooldown Limit Analysis 19 3.4 Hydrostatic Tert and Core Critical 21 Limit Analysis 3.5 Lowest Service Temperature, Minimum Boltup 22 Temperature, and Minimum Pressure Limits 3.6 Development of Technical Specification Figures 23 4.0 DATA 24

5.0 REFERENCES

26 A-MPS-ER-002 Page 2

LIST OF TABLES EQi TITLE PAGE 1 Arkansas Nuclear One - Unit 2 Reactor Vessel 28 Beltline Materials 2 Arkansmo Nuclear One - Unit 2 Reactor Vessel 29 Beltline Art palculations 3 Arkansas Nuclear One P-Allowable (Kei) vs. RCS 30 Temperature (Deg. F) for 21 EFPY, Normal Operation .

1 4 Arkansas Nuclear One Unit 2, Step Change 31 l Transients Table l l

5 Arkansas Nuclear One Unit 2 Ascending Order 32-Step Change Transients Table 6 Arkansas Nuclear One Unit 2 P-Allowable (Kei) vs. 33 RCS Temperature (Deg. F) for 21 EFPY, Hydrostatic Operation 7 Arkansas Nuclear One Unit 2, 21 EFPY 34 Technical Specification Pressuro-Temperature Limits, Limiting Pressure Values A-MPS-ER-002 Page 3

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LIST OF FIGURES HQt DESCRIPTION PAGE 1 Arkansas Nuclear One Unit 2 Beltline P-T 35 Limit s , 21 EFPY Heatup 2 Arkansas Nuclear One Unit 2 Beltline P-T 36 Limit s , 21 EFPY Heatup 3 Arkansas Nuclear One Unit 2 Beltline P-T 37 Limits, 21 EFPY Cooldown 4 Arkansas Nuclear One Unit 2 Beltline P-T 38 Limits, 21 EFPY Cooldown 5 Arkansas Nuclear One Unit 2 Beltline P-T 39 Limits, 21 EFPY Cooldown 6 Arkansas Nuclear One Unit 2 Beltline P-T 40 Limits, 21 EFPY Hydrostatic 7 Arkansas Nuclear One Unit 2 Beltline P-T 41 Limits, 21 EFPY Heatup 8 Arkansas Nuclear One Unit 2 Beltline P-T 42 Limits , 21 EFPY Heatup 9 Arkansas Nuclear One Unit 2 Beltline P-T 43 Limits, 21 EFPY Cooldown

-10 Arkansas Nuclear One Unit 2 Beltline P-T 44 Limits. 21 EFPY Cooldown 11 Arkansas Nuclear One Unit 2 Beltline P-T 45 Limits, 21 EFPY Cooldown 12 Arkansas Nuclear One Unit 2, Technical 46 Specification Heatup Curve, 21 EFPY, Reactor Coolant System Pressure Temperature Limits 13 Arkansas Nuclear One Unit 2, Technical 47 Specification Cooldown Curve, 21 EFPY, Reacter Coolant System Pressure Temperature Limits A-MDS-ER-OO2 Page 4

1.0 INTRODUCTION

The following sections describe the basis for development of reactor vessel beltline pressure-temperature limitations for the Arkansas Nuclear One Unit 2 (ANO2) Nuclear Power Plant. These limits are calculated to meet the regulations of 10 CFR Part 50 Appendix A,I I Design Criterion 14 and Design Criterion 31. These design criteria require that the reactor coolant pressure boundary be designed, fabricated, erected, and tested in order to have an extremely low 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( . This appendix describes the requirements for developing the pressure-temperature limits and provices 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 ASHE Boiler and Pressure Vessel Code Section III, Appendix G, " Protection Against Nonductile Failure."

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

The Reactor Pressure Vessel beltline pr"asure-temperature limits are based upon the irradiation damage predic; ion methods of Regulatory Guide 1.99 Revision 02( } . This methodology has been used to calculate the limiting material Adjusted Reference Temperatures for ANO Unit 2 and have utilized fluence values for 21 Effective full Power Years (EFPY).

A-MPS-ER-002 Page 5

This report provides reactor vesse: beltline pressure-temperature limits in accordance with 10 CFR 50 Appendix G for 21 EFPY. The events analyzed for cooldown were the isothermal condition and 25, 60 and 100*F/hr based on linear rates of temperature change. In addition, the linear cooldown rates were also approximated as a series of step changes. The step change transient profiles were based on a step decrease in temperature equal to half the linear rate followed by a thirty (30) minute hold period. The events analyzed for heatup included 50, 60, 70, 80, 90 and 100'F/hr condition also based on linear rates, l

2.0 ADJUSTED REFERENCE TEMPERATURE PPOJECTIONS In order to develop pressure-temperature limits over the design lifo )

of the reactor vessel, adjusted reference temperatures (ART) for the l

controlling beltlino material need to be determined. The adjusted reference temperatures of reactor vessel beltline materials for Arkansas Nuclear One Unit 2 have been calculated at the 1/4t and 3/4t locations after 21 EFPY of operation. By comparing ART data I for each material, the controlling materials for ANO Unit 2 have 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 I . The calculative procedure for the ART values for each material in the beltline is given by the following expression:

ART = Initial RT 9" NDT NDT Initial RT is the reference temperature for the unieradiated DT material. ART is the mean value of the adjustment in the reference temperature caused by irradiation and is given by the following expression:

ART , I 7) f(0.28 - 0.10 log f)

NDT I

A-t'.PS-ER-002 Page 6

, , . _ ~ . . . . . - - . . - - . - ~- .

Cr 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 expressions ,

~'

f=f, f (e *)

The term f, f is the calculated value of the neutron fluence 19 2 (10 n/cm , E.> 1MeV) at the inner wetted surface of the vessel and x is the depth into the vessel wall from the inner wetted surface in inches.

Margin-is ths quantity that is added to obtain a conservative upper

-bound value of ART. The margin term is given by the.following expression:

Margin = TVh 2 + 2 (4)

I d The terms a and a represent the standard deviation for initial d

RT # ** " * " * ***" * "" # * #*

  • NDT temperature shift.

The following information provides the basis for the calculated ART values for ANO Unit 2:

1. Material data were obtained from Reference 5, including copper content, nickel content and initial reference temperature (RTNDT). These data-are summarized in Table 1 for ANO Unit 2.

A-MPS-ER-OO2 Page 7

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.2. Peak neutron fluence for the ANO Unit 2 beltline region was determined to be 3.74 x 10 n/cm (E > 1MeV) at 21 EFPY (Reference 6).

3. Shell course minimum reference thickness is 7.875 in, for both the lower and intermediate shells (Reference 7).
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 0*F for measured NDT values and 17*F for welds'without measured values (Reference 8).

Adjusted ~ reference temperatures for all bnitline materials at the 1/4t and 3/4t locations through 21 EFPY were calculated using Regulatory Guide 1.99 Revision 02 and the results of the calculations are listed in. Table 2 for ANO Unit 2. The controlling materials can be established from the results of tLa material evaluation 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 ILmits for the corresponding time period.

In the case of ANO Unit 2, the plates C-8009-1 and C-8010-1 are controlling at the 1/4t'and 3/4t locations through 21 EFPY based on the predicted ART values of 111*F and 96'F respectively.

According to Position 1.1 of Regulatory cuide 1.99, Revision 2 ,

the uncertainty in the value of initial RT * * "" * *

  • NDT from the precision of test method when a " measured" value of initial-RT is available. RT e e e - cc r ance with NB3200 of NDT E NDT the ASME Boiler and Pressure Vessel Code,Section III. It involves both a series of drop weight l ASTM E208) and Charpy impact (ASTM E23) tests on the material. The RT NDT
      • " "9 # * *** * **

methods of evaluation are conservatively biased. The elements of this conservatism include:

A-MPS-ER-002 Page 8

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1) Choica for RT in the higher of NDT or T y - 60*F. The DT drop-aeight test is performed to obtain NDT and a full Charpy impact curve is developed to obtain T # " 9 * * " "" I ' S
  • CV The combination of the two test methods gives protection again st 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 T -

60*F) assures that tests at temperaturos above the reference temperature will yield increasing values of toughness, and verifies the temperature dependence of tne fracture toughness implicit in the K p curve (ASME Code,Section III, Appendix G) .

2) Selection of the most adverse Charpy results for T I" CV' accordance with NB2300, a temperature, T gy, is established at which three Charpy specimens exhibit at least 35 mile lateral expansion and not less than 50 ft-lb absorbed energy. The three specimens will typically exhibit a range of lateral expansion and absorbed energy consistent with the variables inherent in the test specimen temperature, testing equipment, operator, and test specimen (e.g., dimensional tolerance and material homogeneity). All of these variablos are controlled using process and procedural controls, calibration and operator training, and they are conservatively bounded by using the lowest measurement of the three specimena. Furthermore, two related criteria are used, lateral expansion and absorbed energy, where consistency between the two measurementa provides further assurance that they are realistic and the material will exhibit the intended strength, ductility and toughness implicit in the K 7g curve.
3) Inherent conservatism in the protccol used in performing the drop-weight test. The drop-weight test procedure was carefully designed to assure attainment of explicit values of deflection A-MPS-ER-002 Page 9 I

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and stress concentration, eliminating a specific need to account for below nominal test conditions and thereby I guaranteeing a conservative direction of these uncertainty componente. In addition, the test protocol calls for decreasing temperature until the first fallure is encountered, followed by increasing the test temperature 10*r above the point where the last 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.

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  • NDT js conservative measure of the reference temperature. The conservative f- bias of the NB2300 methodology and the drop-weight test protocci i

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l A-MPS-ER-002 Page 10

'.d'* *,i.-.14-.og.c--oc -n e.w-v-w=tg px,9i,-7 e.m.ev.r.m-np.-- ey -.'Wvy.'-. g -g+qwg.g-7---.ryvwf.eNlid*f.-Bh TC*e-r rv e g--ymm.,-W9 wwve WT yvW Q.--eq t t(e 4+.='prqtop 'T ----=T*"pTT-g i-rg T'F'= t Y T'M

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precision of an individual drop-weight or Charpy Lmpact test.

I Therefore, when measured values of RT are available, the estimate NDT of uncertainty in initial RT is taken as zero.  !

NDT

-3.0 GENERAL APPROACH FOR CALCULATING PRESSURE-TEMPERATURE LIMITS )

i The analytical procedure for developing reactor vessel pressure-temperature limits utilizes the methods of Linear Elastic Fracture Mechanics (LETM) 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 C (Reference 2). For l

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

The general method utilizes Linear Elastic Fracture Mechanics procedures. Linear Elagtic Fracture Mechanics relates the size of a flaw with the allowable loading which precludes crack initiation.

This relation is based upon a mathematical strega analysis of the beltline material fracture toughness properties as prescribed in Appendix C to Section III of the ASKE Code.

The reactor vessel beltline region is analyzed assuming a l semi-elliptical surface flaw oriented in the axial direction with a i depth of cne quarter of the reactor vessel beltline thicknesc 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 (referred 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.

A-MPS-ER-002 Page 11

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l At each'of the postulated flaw locations the Mode I stress intensity factor, K g , produced by each of the specified loadings is calculated. Then the summation of the Kg values is compared to a reference stress intensity, r.IR, which is the critical value of K l for the material and temperature involved. The result of thia '

l method is a relation of pressere versus temperature for each l condition ana.lyzed providing reactor vessel operating limits which preclude brittle fracture. K is btained from a reference IR I fracture toughness curve for low alloy reactor pressure vessel steels as defined in Appendix 0 to section III of the AsMt codo.

, This governing curve is defined by the following expression:

I l

r = 26.78 + 1.23 e[0.0145(T-ART + 160))

where, l

K = reference stress intensity f actor, Kei1((n i

l l T = teaporature at the postulated crack tip, *r ART = adjusted reference nil ductility temperature at the postulated crack tip, 'F l

l For &ny instant during the postulated heatup or cooldown, K "

l IR j calculated at the metal temperature at the tip of the flaw, and using the value of adjusted reference temperature at that flaw location. Also for any instent during the heatup or cooldown the temperature gradients across the reactor vessel wall are calculated (see section 3.1) and the corresponding thermal stress intensity factor, K is determined. Through the use of superposition, the IT, thermal strees intensity is subtracted from the available K to 1

f determine the allowable pressure stress intensity factor and consequently the allowable pressure.

l A-MPS-ER-002 Page 12 l

In accordance with the ASME Code Section III Appendix 0 requirements, the general equations for determining the allowable pressure for any assumed rate of temperature change during Service j Level A and B operation are: l I

l IM IT IR 4

1.5Kgg + KIT <K IR (Inservice Hydrostatic Test) where, K

gg

= Allowable pressure strens intensity f actor, KoiN /in K

T

= Thermal stress intensity factor, Ks D/I~n K = Reference stress intensity, KsD/in The pressure-temperature limits provided in this report account for the temperatura differential between the reactor vessel base metal and the reactor coolant bulk fluid temperature. Corrections to account for pressure differentials between the locatioa of concern and the location of measurement due to elevation differences and RCS flow are included in the development of the pressure-temperature limits. Uncertainties for the temperature and pressure instrumentation loops associated with control room indications are also included. Consequently, the P-T limits are provided on cocedinates of indicated pressurizer pressure versus indicated RCS temperature.

The reactor coolant system pressure measurement in taken from the pressurizer. The differential pressure due to the elevation difference between the reactor vessel beltline wall and the pressurizer was conservatively established and ei.al to -27.2 psia.

The pressure differential due to the flow indutoo pressure drop between the reactor vessel inlet nozzle and the pressurizer surge line nozzle was established based on three (3) reactor coolant pumps A-MPS-ER-002 Page 13 I

. - .- ,,---,--,.,--...r - - - * - -~r

. t operating and is equal to -58.8 psi. The uncertainty associated with the pressure indication instrument loop is -85.0 poi and was provided by Reference 9. The unegrtainty associated with flow induced pressure drop is ~8.5 pai. The total uncertah.nty propagation depends on the squares of the independent uncertainties, and is determined by taking the root-mean-square summation of the l

!.ndependent un, certainties. Therefore, the total uncertainty is calculated to be -85.4 pai. This information wam combined to i determine the following pressure correction factor utilized in the development of the P-T curves:

l Cooldown and Heatun l Pressure Correction T ('r) Pactor (psi)

All RCS -171.4 Temperatures In addition, the uncertainty associated with the teacperature indication instrument loop was included. The value of r.his uncertainty la +30*P and was provided by Reference 9. .

By explicitly accounting for tho temperature differential between the flaw tip base metal temperature and the reactor coolant bulk l fluid temperature, and the pressure differentials between the beltline region of the reactor vessel and the pressurizer including the uncertainties associated with the indication loops, the P-T limits are correctly represented on coordinates of indicated pressurizer pressure and indicated cold leg temperature.

3.1 THERHAL ANALYSIS METHODOLQQX, 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 l

i A-MPS-ER-002 Page 14

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 in calculated.

The code utilizes a convective boundary condition on the inside wall of the vessel and an insulated boundary on the outside wall of the vessel. Variation of material properties through the vessel uall 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, 2

dI dI

  • 137
  1. p Ot " g 2 rde subject to the following boundary conditions at the inside and outside wall surface locations At r = r -K dI g

a h (T-T )

At r = r dI =0 g

where, 3

p = density, lb/ft C = specific heat, btu /lb *F K = thermal conductivity, btu /hr-ft *F T = vessel wall temperature, *F r = radius, it t = t ime , hr h = convective heat transfer coefficient, btu /hr-ft *F T = RCS coolant temperature, *F rg ro = inside and outside radii of vessel wall, it i

A-KPS-ER-OO2 Page 1+

l l

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The above is solved numerically using a firite element model to determine wall temperature as a function of radius, tLme, and

. 'l rate. Thermal stress intensity factors are determined by the culated temperature dif f erence and profile through the welt no wall using thermal influence coefficients specifically generated for this purpose. The influence coefficients depend upon qsometrical parameters associated with the maximum postulated defect, and the geometry of the reactor vessel beltline region (i.e., r /r g , a/c, a/t where a= crack depth, c= crack half length, and t-vessel wall thickness), along with the assumed unit loading.

ASKE code Section III Appendix G recognizes the limitations of the method it provides for calculating K because of the assumed T

temperature profile. Since a detailed heat transfer analysis results in varying temperature profiles (and consequently varying thermal stresses), an alternate method for calculating K w""

IT employed as required by Article C-2214.3 of Reference 3. The alternate method employed used a polynomial fit of the temperature profile and superposition using influence coefficioets to calculate K The influence coefficients were calculated using a T.

2-dimensional finite element model of the reactor vessel. The influence coefficients were corrected for 3 dLmensional effects using ASTM Special Technical Publication 677 (Reference 10).

3.2 HEATUP LIMIT ANALYSIS During a heatup transient, the thermal bending stress is compressive at the reactor vessel inside vall and is tensile at the reactor vessel outside wall. Internal pressure creates a tensile stress at the inside wall as well as the octaide wall locations.

Consequently, the outside wall location has the larger total stress when compared to the inside wall. However, neutron embrittlement (the shift in material RT NDT "" * *** ** #" "" '" "

fractare 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.

A-MPS-ER-002 Page 16

I l

As described in the cooldown case, the reference stress intensity f actor is calculated for the metal temperature at the tip of the flaw and the adjustoa reference temperature at the flaw location.

For hastup the reference stress intensity is calculated for both the 1/4t and 3/4t locations. Using the finite element method described ,

1 in section 3.1, the temperature profile through the wall and the  ;

1 metal temperatures at the tip of the flaw are calculated for the transient history. This information is used to calculato the thermal stress intensity factor at the 1/4t and 3/4t locations using the calculated wall gradient and thermal influence coefficients.

The allowable precoure stress intensity is then determined by superposition of the thermal wtress intensity f actor 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 ilmiting P-T limit. However, the isothermal condition may not always provide the limiting pressure-temperature limit for the 1/4t location during a heatup transient. This is due to the correction of the base m6tal 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 A-MPS-ER-002 Page 17

, , l for clad and film differential temperatures and for the gradual buildup of wall dif f erential temperatures with time, as do the cooldown 1 Laits.

At the 3/4t location the pressure stress and thermal stresses are I

tensile resulting in the maximum stress at that location. Pressure- l 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 3.1. The allowable pressure was derived based upon a flaw at the 3/4t location by superposition of the thermal stress intensity with the available reference stress intensity for the metal temperature and adjusted ref erence temperature at that position.

To develop composite pressure-temperature limits for the heati.p 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 50, 60, 70, 80, 90 and 100'r/hr heatup pressure-temperature limits. This table provides the allowable pressure versus reactor coolant temperature for the various heatup conditions. The allowable pressure is in units of pai while the temperature is in units of *F. Figures 1, 2, 7 and 8 provide a graphical presentation of the heatup pressure-temperature 1Luits found in Table 3. It is permissible to linearly interpolate between the heatup pressure-temperature limits.

A-MPS-ER-002 Page 18

1 3.3 COOLDOWN LIMIT ANALYSIS i

During cooldown, membrane and charmal bending stresses act together l in tension at the reactor vessel inside wall. This results in the pressure stress intensity factor, Kyg, and the thermal stress intensity factor, K , acting in unison creating a high stress intensity. At the reactor vessel outside wall the tensile pressure 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 * "'" '"

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 cocidown event.

Utilizing the material metal temperature and adjusted reference temperature at the 1/4t location, the reference stress intensity is determined. From che method provided in Section 3.1, 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 3.1. 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:

IR - IT where, K = Allowable pressure stress intensity as a function of g

coolant temperature, Kai d A-MPS-ER-002 Page 19

. ~ _

l I

J K = Reference stress intensity as a function of coolant temperature, KaiVTn i l

I J

K = Thermal stress intensity an a function of coolant j temperature, KaiVTn To develop a composite pressure-temperature limit for a specific i

cooldown event, the isothermal pressure-temperature limit must be calculated. The isothermal pressure-temperature limit is then compared to the pressure-temperature limit $ssociated with a specific cooling rate and the more restrictive allowable pressure-temperature limit is chosen resulting in a composite limit curve for the reactor vessel beltline.

Table 3 provides the results for the isothermal, 25, 60, and 100'F/hr linear cooldown pressure-temperature limits. Table 4 provides the results for the cooldown analysis corresponding to the step-change transient profiles of 12.5'F/ hour, 30'F/ hour, and  ;

50*F/h hour. In order to obtain the allowable pressures at the specified cooldown rate transition temperatures (150'F and 230*F),

the step-change transients were initiated at an appropriate temperature to include the temperature transition point of interest.

Therefore, multiple transients were defined for each step-change rate to specifically address the break points between the various cooldown rates ao shown in Table 4. These tables provide the allowable pressure versus reactor coolant temperature for the various cooldown conditions. Tha allowable pressure is in units of psi while the temperature is in units of 'F. Figures 3 through 5 and 9 through 11 provide a graphical presentation of the cooldown pressure- temperature limits found in Tables 3 and 4. Table 5 tabulates in ascending order the minimum and maximum pressures resulting from each step-change rate. This table lists the results shown graphically in Figures 3 through 5 and 9 through 11. It is permissible to linearly interpolate between the linear cooldown pressure-temperature limits.

A-MPS-ER-002 Page 20

l r

3.4 HYDROETATIC TEST AND CORE CRITICAL tfMIT ANALYSIS Both 10 CFR Part 50 Appendix G and the ASHE Code Appendix G require the development of pressure-temperature limits which are applicable to inservice hyocostatic tests. For hydrostatic tests performed subsequent to load'ng fuel into the reactor vessel, the minimum test tamperature 1s determined by evaluating K g, the mode I stress intensity factors. The evaluation of Ky is performed in the sLme 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 21 EFPY in Table 6 and to shown in Figure 6.

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 core reactivity safecy analyses which can control the temperature at

.which the-core can be brought critical.

A-MPS-ER-002 Page 21

3.5 LOWEST SERVICE TEMPERATURE. MINIMUM BOLTUP TEMPERATURE. AND MINIMUM PRESSURE LIMITE In addition to the computation of the reactor vessel beltline P-T ILmits, additional ILmits have been provided for reference. These additional limits are the Lowest Service Temperature, Minimum Boltup Temperature, and Minimum Pressure Limits. These limits are described below.

The Lowest Service Temperature is the minimum allowable temperature at pressures above 20% of the pre-operational system hydrostatic ter4t pressure (625 psia). This temperature is defined an equal to the most limiting RT f r the balance of Reactor Coolant System T

(RCS) components plus 100*F, per Article HB 2332 of Section III of the ASME Boiler and Pressure Vessel Code.

The maximum RT f r the balance of the RCS components was NDT conservatively estimated ao 50*F. Therefore, the Lowest Service Temperature is equal to 100'F + 50*F + 30'F = 180'F.

The minimum pressure limit in 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 preneure is as follows when pressure correction factors for elevation, flow and uncertainties are taken into account Cooldown and Heatuo 453.6 pela All RCS Temperatures The minimum boltup temperature i s the mintmum allowable temperature at pressures below the 20% of too pre-operational system hydrostatic test pressure. The minimum i , defined as the initial RT NDT material of the higher strersed region of the reactor vessel plus any effects for irradiation oer Article C-2222 of Section III of the I

A-MPS-ER-002 Page 22 l

_ _ . ,, __ - .- -- .- - .- .. _- . - ~ - _ . - - - .

l ASME Boiler and Pressure Vessel Code. The initial reference temperatures of the reactor vessel and closure head flanges were determined using the certified material test reports JReferonce 5).

The maximum initial RT ass ciated with the stressed region NDT pertains to the closure head flange and is 30'F. The minimum boltup temperature including temperature instrument uncertainty is 30'r + l 30'r = 60*F. ,However, for additional conservatism a minimum boltup temperature of 70'F has been utilized.

3.6 QEVELOPMENT OF TECHNICAL SPECIFICATION FIGURES The Technical Specification figures shown in Figures 12 and 13 are established based upon the beltline P-T limits and are corrected for pressure drops due to flow and elevation as well as pressure and temperature instrumentation uncertainties. They also include the minimum boltup temperature, lowest service temperature and consider the flange limit. Core critical limits are also specified as is the inservice hydrostatic test curve. The maximum pressure for shutdown cooling operation is included per Reference 5.13.

Table 7 provides a listing of the values used to define the cooldown and heatup Technical Specification figures. The applicable heatup and cooldown rates for Arkansas Nuclear One Unit 2 at 21 EFPY are as follows:

Gooldown Heatuo Temoerature Pate Temocrature Rate

>230'r $100*F/hr All 230*F to 150*F s 60'F/hr RCS 80'F/hr

<l50*F s 25'F/hr Temperatures A-MPS-ER-002 Page 23

. _ ~ _ _ _ _ , _ , _ . . . _ _ _ _ _

These temperature ranges and associated rates were requested by ANO2. The Technical Specification figures represent the most ilmiting pressure for the specified composite heatup or cooldown rate. In the case of cooldown, the composite linear cooldown transient resulte were compared to the step-change transient results to determine the more limiting pressure values within each specified traperature range. The more limiting value from the transient analyses was compared to the isothermal P-T limit values within each specified temperature range and the most limiting value was chosen for the Technical Specification requirements. In the case of heatup, the composite linear heatup transient results were compared to the icothermal P-T limit values and the most limiting value chosen for the Technical Specification requirement.

4.0 DATA Reactor Vessel Data Eeference Design Pressure = 2500 psia 7 Design Temperature = 650'r 7 operating Pressure = 2250 psia 7 Beltline Thickness = 7.875 in 7 Inside Radius (to wetted surface) = 79.436 in 7 Cladding Thickness = .2188 in 7 Meterial-SA 533 Grade B Class 1 Beference Thermal Conductivity =

-23.8 BTU /hr-ft 'T 11 Youngs Modulus = 28 x 10 psi 11

~

Coefficient of Thermal = 7.77 x 10 in/in/'F 11 Expansion Specific Heat. = .122 BTU /lb 'T Density =

.284 lb/in stainless Steel Claddina Thermal Conductivity =

10.1 BTU /hr-ft *F 11 Specific Heat =

0.126 BTU /lb *F 12 Density =

0.285 lb/in 12 l

'A-KPS-ER-002 Page 24 q i

.- . - , - . . - _ . . _ . ~ - .. - - . , , . - . . . , - , , . _ , .. .- -- , , , , - , , . , . - , . - . , . . , , , - . , - , -

Adiusted Reference Termerature Values 21 EFPY Peference 1/4t 111*F 3/4t 96'T j ramt Neutron Fluence =

3.74 x 10 ' n/cm 6 Film coefficient on inside surface = 1000 BTU /hr-ft 'r Pressure Correction Factors for Elevation and Flow httuo and Cooldown All RCs Temperatures dP = -171.4 psia 1

1 Temperature Instrument correction (Reference 9) dT = +30*r A-MPS-ER-002 Page 25

__._L_ _ _ _ . , _ . _ _ . __ _ _... _ _.___,_..-__ _ . _ _ _ _ .

. _ _ _ . - _ - - . . - . . . ~ . - .-- - -

5.0 REFERENCES

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

" General Design Criteria for Nuclear Power Plants", January 1988.

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

" Fracture Toughness Requirements", January 1988.

l

3. ASME Boiler and Pressure Vessel Code Section III, Appendix G,

" Protection Against Nonductile Failure", 1986 Edition.

4. Regulatory Guide 1.99, " Radiation Embrittlement of Reactor Vessel Materials", U.S. Nuclear Regulatory Commission, Revision 2, May 1988. 1 1

, 5. Combustion Engineering, Inc. Material Certification Reporte.

6. C. H. Turk to C. D. Stewart, " Development of ANO-2's Pressure-Temperature and LTOP Limitations Under Contract

, C-100-G," Letter No. ANO-91-2-OOl64, dated February 5, 1991.

7 Instruction Manual, Reactor Vessel Assembly, ANO Unit 2, C-E Book No. 73170, November 1974.

8. " Evaluation of Pressurized Thermal Shock Effects Due to Small Break LOCA's with Loss of Feedwater for the Combustion Engineering NSSS", Combustion Engineering Report CEN-189, December 1981.
9. C. H. Turi to C. D. Stewart, " Development of ANO-2's Pressure-Temperature and LTOP Limitations," Letter No.

ANO-91-2-OO192, dated March 1, 1991.

A-MPS-ER-OO2 Page 26

- _ _ __ _ _ _ _ _ _ _ . _ _ _ _ . _ . .. , . . . ~

i l

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

l

11. ASME Boiler and Pressure Vessel Code Section III, Appendix I, )

i

" Design Stress Intensity values, Allowable Stresses, Material i Properties, and Design Fatigue Curves", 1986 Edition.

12. " Fundamentals of Heat and Mass Transfer," F. P. Incropera and D. P. DeWitt, p. 757, 2nd Edition, Copyright John Wiley and Sons, Inc., 1985.
13. AP&L Arkansas Nuclear Ono Information Request Form No. 5063, Unit EPO Setpoint Document, February 28, 1991.

A-MPS-ER-002 Page 27

\ j t

)

TABLE 1 ARKANSAS NUCLEAR ONE - UNIT 2 REACTOR VESSEL BELTLINE KATERI ALS l 1

I Identification f

component . Number Cu f%) Ni f%) NDT l

)

l III l Intermediate Shell 2-203-A,B,C O.05 0.18 -56'F '

Long. Wolds Lower Shell 3-203-A,B,C 0.05 0.18 -56'F I Long. Welds l

l Inter./ Lower Shell 9-203 0.05 0.08 -10*F Girth Wold Intermediate Shell C-8009-1 0.12 0.63 -26'F Plates C-8009-2 0.08 0.59 0'F C-8009-3 0.08 0.60 0'F Lower Shell C-8010-1 0.08 0.59 12*F Plates C-8010-2 0.07 0.66 -28'F

{

C-8010-3 0.07 0.65 -30*F 1

(1) Generic value for submerged arc wolds.

I l

A-MPS-ER-002 Page 28 )

)

i TABLE 2 ,

B

)

ARKANSAS NUCLEAR ONE - UNIT 2 i Y BELTLINE ART CALCULATION 4

T ,

$ Identification Chemistry Initial 1/4 Thickness 3/4 Thickness b

RT a a ART ART. *F

' ART NDT. 'F ART. *F o Number Factor NDT. *F l _A NDT. *F n

i 2-203-A,B,C 47- -56 17 28* 57 67 45 46 e

i 3-203-A,B,C 47 -56 17 28* 57 67. 45 46 ,

9-203 35 -10 0 28* 43 76 34 58 i

C-8009-1 83 -26 0 17 103 111 81 89 C-8009-2 51 0 0 17 63 97 50 84 l

C-8009-3 51 0 0 17 63 97 50 84 t

!- C-8010-1 51 12 0 17 63 109 50 96 ,

l C-8010-2 44 -28 0 17 54 60 43 49 C-8010-3 44 -30 0 17 54 58 43 47 ,

i l

a 2

us t

  • I
  • Value of 2a is less than or equal to ART 1 m A NDT.

o I

i 4

'l -

. . _ - - ~ . _ _ . . ._ - . - - - _ - _ - - ~ _ _ . - - - . . -- . -

IABLE 3 ARKANSAS NUCLEAR ONE UNII 2 21 EFPY APPEN01X 0 PRESSURE-1EMPERAiutt L]Ml1$

COOLDOWN P ALLOWASLE (P$l A) HEATUP (LINEAR RATES) P*ALLOWA8tt (PSIA)

RCS - - - - - - - * - - - - - * - - - - .

TEMP 150 25 F/ 60 F/ 100 F/ ISO 50 r/ 60 f/ 70 F/ 80 F/ 90 f/ 100 F/

DEG F THERMAL HOUR HOUR HOUR IFUMAL HUUR HOUR HOJR HOUR HOUR NGJE 70 435.2 361.3 260.9 151.5 435.2 435.2 435.2 435.2 435.2 435.2 435.2 80 448.4 376.6 2 79.4 1 74 .2 448.4 448.4 448.4 448.4 448.4 448.4 448.4 90 464.0 394.4 300.8 200.8 464.0 449.3 445.0 444.5 444.0 444.7 445.5 100 482.0 414.9 325.5 231.5 482.0 436.6 427.4 422.9 420.0 417.6 417.0 110 502.8 438.6 354.0 266.8 502.8 433.8 419.7 410.6 403.6 398.9 395.1 120 526.8 466.1 387.0 307.5 526.8 439.2 420.2 406.5 395.6 387.0 379.8 130 554.6 497.7 425.2 354.3 554.6 452.6 427.9 409.3 393.8 381.4 371.7 140 586.7 534.5 469.3 409.3 586.7 472.1 442.1 418.4 399.1 382.8 369.3 150 623.9 576.9 520.3 4 72.6 623.9 497.8 462.6 434.1 409.7 389.7 3 72.4 160 666.8 625.9 579.2 545.5 666.8 529.5 489.2 455.7 427.0 401.9 351.1 170 716.4 682.6 647.4 629.5 716.4 567.7 522.4 483.5 449.7 420.9 395.2 180 773.8 747.9 726.2 726.1 773.8 613.9 562.4 518.1 479.5

  • 445.5 415.8 190 840.2 823.9 817.3 839.6 840.2 668.0 610.1 560,1 515.4 476.3 442.5 200 916.9 911.3 916.9 916.9 916.9 731.0 666.3 609.8 559.6 515.0 4 75.6 210 1005.5 1005.5 1005.5 1005.5 1005.5 803.9 732.1 668.2 611.3 561.1 515.9 220 1108.0 1108.0 1108.0 1108.0 1108.0 838.2 808.7 73 7.4 6 73.4 615.4 564.1 230 1226.3 1226.5 1226.5 1226.5 1226.5 987.6 897.8 817.9 744.9 680.7 622.4 240 *363.4 1363.4 1363.4 1363.4 1363.4 1102.3 1001.2 911.1 830.0 75 6. 6 690.9 250 1521.8 1521.8 1521.8 1521.8 1521.8 1234.4 1121.1 1019.7 927.3 844.3 770.8 260 1704.8 1704.8 1704.8 1704.8 1704.8 1386.7 1257.9 1146.1 1042.4 948.5 863.7 270 1916.5 1916.5 1916.5 1916.5 1916.5 1562.1 1420.5 1291.9 1173.7 1068.5 9 71.4 280 2161.1 2161.1 2161.1 2161.1 2161.1 1768.1 1606.4 1460.0 1328.4 1206.5 1098.7 290 2443.9 2443.9 2443.9 2443.9 2443.9 2005.6 1821.3 1656.6 1504.7 1369.4 1246.0 300 2770.8 2770.8 2770.8 2770.8 277).8 2279.1 2070.0 1883.2 1712.2 1556.6 1416.1 310 ~~ - -- - -

2594.0 2357.5 2144.3 1948.5 1771.2 1612.3 320 *- - - - -- --

2690.0 2447.0 2226.3 2024.3 1838.5 330 " -- " - " " "

2798.3 2542.5 2314.5 2104.7 340 - " " ' " " - " ' 2647.1 2411.7 350 " - " " " - " " " "

2765.4 CORRECT!0W FACTOR $3 (HEATUP AND C00LDOWN)

TEMPERATURE +30 F PRESSURE ALL RC$ TEMPS DELTA.P = 171.4 PSIA T EMFERATU RE LIMIT NG RATE RANGE LOCATION 150 THERMAL ALL RCS 114t TEMPERATURES COOLDOWN ALL RCS 1/41 TEMPERATURES HEATUP 70 < T < so F 1/48 to < 7 < 350 F 1'41 A-MPS-ER-002 Page 30

_ _.. ___ _ _ _ _ .~ _ _ , _._.-_ _ ,__ ___

, . _ . . - _ - . . - - ~ _ - _ --. -- . . - . . . . .. ._ .. . - - - _ .

e i I

l l

TABLE 4 ARKANSAS WUCLEAR ONE UNIT 2 21 EFPT APPENDIX G PRESSURE TEMPERATURE LIMITS

($1EP CMANCE RATES) - C00LDOWN (12.5 F / 1/2 HR) (30 F / 1/2 HR) (50 F / 1/2 HR)

RCS Pat $$ 8 RCS PRESS S RCS PRES $ 3 TEMP 25 F/HR TEMP 60 F/NR TEMP 100 F/NR (DEG F) (PSIA) (DEG F) (P11A) (DEG F) (PSIA) 70 388.6 160 588.6 230 1308.6 70 348.6 1W 558.6 230 1330.6 70 378.6 '

160 718.6 230 1368.6 82.5 398.4 190 798.6 230 1418.6 '

82.5 368.6 190 788.6 230 1518.6 82.5 398.6 190 1028.6 230 1748.6 )

95 428.6 220 1118.6 280 2588.6 1 95 388.6 220 1508.6 '

1 95 418.6 107.5 448.6 150 538.6 l 107.5 418.6 150 498.6 l 107.5 458.6 150 638.6 '

180 718.6 80 398.6 .180 698.6 80 358.6 180 908.6 80 388.6 210 998.6 92.5 418.6 210-- 1328.6 92.5 378.6 92.5 418.6 170 648.6 105 448.6 170 628.6 105- 408.6 170 808.6 105 448.6 200 888.6 117.5 478.6 200 1168.6 117.5 448.6 230 1268.6 117.5 488.6 230 1718.6 130 108.6 130 488.6 130 528.6 142.5 558.6 142.5 528.6 142.5 578.6 112.5 ~ ~ 458.6 112.5 428.6 112.5 468.6 125 498.6-125 468.6-125' 508.6 137.5 538.6 137.5 508.6 137.5 558.6 150 588.6 150 568.6 150 618.6 CORRECTION FACTORS: (HEATUP AND C00LOOWN)

. TEMPERATURE +30 F PRESSURE ALL RCS TEMPS - DELTA P * -171.4 PSIA NOTE:

IN ORDER TO 08TAIN THE ALLOWA8LE PRESSURES AT THE $PECIFIED C00LDOWN RATE TRAN$1 TION TEMPERATURES (150 F AND 230 F), THE STEP-CHANGE TRANSIENTS UERE INITI ATED AT AN APPROPRI ATE TEMPERATURE 10 INCLUDE THE TEMPERATURE PolNT OF INTEREST. THEREFONE, WLTIPLE TRANSIENTS WERE DEFlWED FOR EACH RATE OF TEMPERATURE CMANGE TO SPECIFICALLY ADDRESS THE 8REAK POINTS 8ETWEEN THE VARIOUS COOLDOWN RATES. THE A80VE TA8LE NAS THE RESULTS OF EACH STEP CHANGE RATE ANALTZED. SupetAtlZED UNDER EACM RATE ARE THE INDIVIDUAL TRANSIENT PRE $5URE TEMPERATURE RESULTS.

A-MPS-ER-002 Page 31

.a. . = - - - = .. . _ _ - . . = .

- . - .- - . . - - - - - - _ - - - - . . . . .-_.--- - .~.~.~. .-- .-.-. -

o

  • TASLE 5 ARKANSAS NUCLEAR ONE UNIT 2 21 E f PY ,

APPEbolX G PRESSURE

  • TEMPERATURE L1 HIT 5 '

(5TEP CHANGE RATES)

  • COOLDOWN (ASCENDlhG ORDER)

(52.5 F / 1/2 HR) (30 t / 1/2 HR) (50 F / 1/2 HR)

RCS PRESS S RCS PRES $ 8 RCS PRESS S TEMP 25 f/HR TEMP 60 f/HR TEMP 100 f/MR (DEG f) (P$l A)- (DEG f). (PSIA) (DEG f) (PSIA) 70 388.6 150 498.6 230 1308.6 70 348.6 -

150 638.6 230 1748.6 60 398.6 160 558.6 2 50 25B8.6 J0 358.6 160 718.6 B2.5 398.6 1 70 628.6 82.5 3 68.6 1 70 808.6 92.5 418.6 180 698.6

-92.5 378.6 180 908.6 95 388.6 190 788.6 95 428.6 190 1028.6 105 448.6 200 888.6 105 408.6 200 1168.6 107.5 418.6 210 998.6 )

107.5 458.6 210 1328.6

-112.5 428.6 220 1118.6 112.5 4 68.6 220 15D8.6 117.5 448.6 230 1268.6 117.5 688.6 230 1718.6 125- 468.6 125 508.6 130 Ah8.6 130 528.6 137.5 508.6 137.5 558.6 142.5 528.6 i 142.5 5 78.6 150 568.6 150 618.6 - CORRECTION F ACTOR $s (HEATUP AND COOLDOWN)

TEMPERATURE +30 F PRESSURE ALL RC$ TEMPS DELTA *P e 171.4 PSIA NOTE:

1HIS TABLE REPRESENTS THE CONEGLIDATION of THE MULTIPLE TRANSIENTS (SEE TABLE 4)

ANALYZED FOR EACH $PEClfic RATE. $PEClflCALLY IT TABULATES IN ASCENDING ORDER THE MIN 10RM AND MAXIMLM PRES $URES RESULT lWG FROM EACH $TEP* CHANGE IN TEMPERATURE. THE EXACT PRES $URE TRAN$lT!DN PR0flLE 15 NOT. INCLUDEO SINCE THE LIMITING VALUES ARE SHOWN.

THit TABLE IS UTILI2ED TO PRESENT THE RESULTS GRAPHICALLT. THESE RESULTS ARE $NOWN ALONG WITH THEIR RESPECTIVE LINEAR RATES IN FIGURE $ 3 THROUGH 5, AND 9 THROUGH 11 r

A-MPS-ER-002 Page 32

. e TABLE 6 ARKAWSAS NUCL( AR ONE OWlf 2 21 (FPT Arr[wDix C PRESSURE 1EMPERATURE LIM 115 HTDROSTAi!C RCS 1(MP HYDROSTATIC CEG F (PSI) 70 637.5 80 655.0

  • 90 675.8 100 699.8 110 727.5 120 75 9.5 130 796.6 140 839.4 150 689.0 160 946.2 170 1012.4 180 1083.9 190 1177.4 200 1279.6 210 1397.8 220 1528.2 230 1678.0 240 1851.3 250 2051.5 260 2283.0 270 2550.7 CORRECilDN FAC10RS:

itMPERATURE +30 F PRESSURE ALL RCS TEMPS DELTA P

  • 171.4 PSIA TEMPERATURE LIMITING RATE RANGE LOC ATION ISOTHERM AL ALL RCS 1141 TEMPERATURES A-MPS-ER-002 Page 33

e s I

l TABLE 7 ARKANSAS huCLEAR ONE UNIT 2 l 21 EfPY I TECHNICAL SPECIFICATION P8E55URE TEMPERATURE LIMIT $

COOLDOWW HEATUP BELTL!hE BELTLibE RCS COMP 0511E CURVE RC$ COMPO$li[ CURVE TEMPERATURE P ALLOWABLE T E MPE RATURE P ALLDWABLE DEG. F PSIA DEG. F PSIA 70 348.6 TO 393.8 1 80 358.6 80 393.8 '

82.5 168.6 90 393.8 4 i

92.5 3 78.6 100 393.8 '

95 388.6 110 393.8 105 408.6 120 393.8

'37.5 418.6 130 393.8 112.5 428.6 140 3 99.1 117.5 448.6 150 409.7 130 458.6 160 427.0 142.5 528.6 170 449.7 150 498.6 180 479.5 1 60 558.6 190 515.4 170 628.6 200 559.6 180 698.6 210 611.3 190 788.6 220 673.4 200 888.6 230 744.9 210 998.6 240 830.0 220 1108.0 250 927.3 230 1226.5 260 1042.4 240 1363.4 270 1173.7 250 1521.8 260 1328.4 260 1704.8 290 15D4.7 2 70 1916.5 300 1712.2 280 2161.1 310 1948.5 290 2443.9 320 2226.5 300 2770.8 330 2542.5 A-MPS-ER-04)2 Page 34

l l

FIGURE 1 ARKANSAS NUCLEAR ONE UNIT 2

^

APPENDIX G PRESSURE TEMPERATURE LIMITS 21 EFPY, HEATUP ISO 50*FIHR =

I 70*FIHR 90*F/HR

/ / /!=/

.g 2,000 D

$ ' /

1,500 cr y ..

if . .

D

@ 1,000

a. 7 O

9

$ 500 xxy 0

50 100 150 200 250 300 350 400 INDICATED RCS TEMPERATURE-Tc, DEG F ART AP = 171.4 PSIA 1/4t = 111.0*F AT = +30*F 3/4t = 96.0*F A-MPS-ER-002 Page 35

1

)

)

FIGURE 2 ARKANSAS NUCLEAR ONE UNIT 2 APPENDIX G PRESSURE-TEMPERATURE LIMITS l 21 EFPY, HEATUP l 2,500 -

Iso -

60'F/HR  :=.

80*.",MRl-  !

!  ! /

~

10CF/Hh

2,000 -

/ / / =/

g ,7 -l, y .

g .. /

I l w

a- I

$ ~

$ 1,500 ' -

o_. f 1

~

N ct .

D

~

/

$ 1.000 7 7

c. _

O w _.

/

~

N 500 ,,< -

. Lp fjp 0

50. 100 150 200 250 300 350 400 INDICATED RCS TEMPERATURE-Tc, DEG. F ART AP = -171.4 PSIA 1/4t = 111.0*F AT = +30*F 3/41 = 96.0*F A-MPS-ER-002 Page 36 .

. --..._,,.,..m. . , _ . , _ _ _ , ,,,,...,,._._..-m. . . - , , . _ . _ ~ ~..m, . ~ . . _ _ . . . . _ - . , _ , ,

FIGURE 3 ARKANSAS NUCLEAR ONE UNIT 2 APPENDIX G F RESSURE-TEMPERATURE LIMITS 21 EFPY, COOLDOWN 2,500 4

4 2,000

. lb . -

Q.

. $ 1,500 '

O. ~

T k -

rr .

D

-@ 1,000 g *

! TEMPERATURE fWAP

' RANGE RATE STEP

  • W -

T < 150 F 25 FMR  !!.6 F /1/2 HR l$Q

$

  • REPRESENTS THE SPECIFIED INSTANTANEOUS O -

DECREASEIN TEMPO 4ATL AE AND A1HIRTY

-Z MINUTE HotD.

500 p --

' ' [a 257. HR (128FJiHR 3.F/HR '

"^

0 ' ' ' ' ' ' ' ' ' ' ' ' ' '

50 100 150- -200 250 300 350 400 INDICATED RCS TEMPERATURE-Tc, DEG. F AP = 171.4 PSIA '

AT = +307 1/4t = 111.0'F 3/41 = 96.0*F 1

A-MPS-ER-002 Page 37 l

l

L l

l FIGURE 4 ARKANSAS NUCLEAR ONE UNIT 2 APPENDlX G PRESSURE-TEMPERATURE LIMITS 21 EFPY, COOLDOWN 4 2,000 M -

G.

Ld C - i D 60 7/HR (W F/ iHR --

Q -

STEPCHANGE)

$ 1,500 '

a- .

C LU ,

CS C -

D in -

m

@ 1,000 /

a. '

/ TcupeRArune RAue g RANGE Rs* step

  • isof T3:3o r so riuR 20 e i si2 nR ISO

//

O.__

O . 60 F/HR -f /

/

'HEPRESENTS THE SPECIFIED INSTANTANEOUS oEcacAseis TodPERATURE AND A THIRTY Z / umure noto.

500 V

/

0 ' ' ' ' ' ' ' ' ' '

50 100 150 200 250 300 350 400 INDICATED RCS TEMPERATURE-Tc, DEG, F AP = -171.4 PSIA ART AT = +307 1/4t = 111.0*F 3/41 = 96.0*F A-MPS-ER-002 Page 38

FIGURE 5 ARKANSAS NUCLEAR ONE UNIT 2 APPENDIX G PRESSURE-TEMPERATURE LIMITS 21 EFPY, COOLDOWN 2,500 l

^

. J 4 2,000 .)

1

55. .

0.

g. -

100"F/HR (5 1F/ i HR STEP CHANG?.)

to en w

~

l r

c 1,500

. a. ~

T w _

N ,

[ .

D en . -

en

@ 1,000 '

'A TEMPERATURE RAMP RANGE RATE STEP

  • g y ~

iso T > 230 F 100 F/HR 50 Fi1/2 HR 9 -100*F/HR -

REeREsENTs THE specerico NSTANTANECUS O -

-II / OECREASE IN TEMPERATURE AND A THIRTY Z .

'/ MINUTE N -

500

/

Y

~

/

0 50 100 150 200 250 300 350 400 INDICATD TS TEMPERATURE-Tc, DEG. F ART AP = 171.4 PSIA 1/4t = 111.0*F AT = +30*F 3/4t = 96.0 F A-MPS-ER-002 Page 39

l l r FIGURE G ARKANSAS NUCLEAR ONE UNIT 2 APPENDIX G PRESSURE-TEMPERATURE LIMITS 21 EFPY, HYDROSTATIC _

I 2,000 4

G5 O. -

~

Ld C _

D <

cn ..

m 1,500 0- _

E N

  • e _

D

~

W 1 000

-E Q

[

w -

D ~

Q /

O -

500 0

50 100 150 200 250 300 350 400 INDICATED RCS TEMPERATURE-Tc, DEG. F ART aP = 171.4 PStA 1/4t = 111.07 AT = +307 3/4t = 96.0T A-MPS-ER-002 Page 40

FIGURE 7 ARKANSAS NUCLEAR ONE UNIT 2 APPENDIX G PRESSURE-TEMPERATURE LIMITS '

21 EFPY, HEATUP I 1,000 950 i / ,/ / /  !

900 )

< 850 i / / /

/ /

e ui T

800

- l

/ / l D -

750 )

u) # 7 e

Q-~ -700

-. / f

/

650 -

l m 600 l / <

/ //

550 '

9

< 500 i 'ey / / /

/

- '450

~

/ J b / /

400 _

E Myog f

/

350  :

~

u.i' , ' ' , ' , ' ' '

300 50 75 100 125 150 175 200 225 250 275 300 INDICATED RCS TEMPERATURE-Tc, DEG. F ART oP = -171.4 PSIA 1/4t = 111.0*F

! oT = +30*F 3/4t = 96.0*F A-MPS-ER-002 Page 41 l

t. .._ . _ ._ _ _ .

FIGURE 8 ARKANSAS NUCLEAR ONE UNIT 2 APPENDIX G PRESSURE-TEMPERATURE LIMITS 21 EFPY, HEATUP 1,000 950

/ / / /

[ / I 900

/ / /

< 850 i / / / /

ui 800 _

i / / / /

W 750

~

! '!/

1 e

700

/

/ // '

/ / ,

r tu -

N

~

/

G50

///

m 600 i / ,

550 500 9  : / / / /

a z 450

~

l '

/

f j 400 i

~

K8sv 350 300 ~' '

50 75 100 125 150 175 200 225 250 275 300 INDICATED RCS TEMPERATURE-Tc, DEG F AP = -171.4 PStA '

AT = +30 F 3/4t = 960'F A-MPS-ER-002 Page 42

FIGURE S ARKANSAS NUCLEAR ONE UNIT 2

] APPENDIX G P9 ESSURE-TEMPERATURE LIMITS 21 EFPY, COOLDOWN 1,000 .

950 ~

900 ~

< 850 i /

uf 800 [

[ -

a  : .',

@ 750 _ j w -

/

/ j/

[  :

0- 700

[  : /

w -

/

y

~

650 4 7

a  : 4 to -

/

ca 600 .

7 I 7DAPERATURE RAMP a- ' "** * "*** "

550 /

T < 150 F 25 F/HR 12.5 F i 1/2 HR F<- 500

- ISO /

7 j .ncenestNTs Tue setCtFIQ INSTANTANEOUS DECREASEIN TEMPERATURE AND ATHIRTY a -

~ / uinuTE souo.

Z 450

'- /
/ 4 400 -
/ '

257/H R (12.S' F/ i HR

~

, / STEP CHANGE) h 350 .

300 50 75 100 125 150 175 200 225 250 275 300 INDCATED RCS TEMPERATURE-Tc, DEG. F ART AP = -171.4 PSIA 1/4t = 111.07 AT = +307 3/41 = 96.0*F Page 43 A-MPS-ER-002

f FIGURE 10 ARKANSAS NUCLEAR ONE UNIT 2 APPENDIX G PRESSURE-TEMPERATURE LIMITS 21 EFPY, COOLDOWN 1,000 950 i /

5

~

[

900 850 i /

@ 800 }

/7//

d  :

C 750 .. } fl 3 l, m -

700

~

  1. }

r m  ; 60'F /HR (30'iF / i HR > [ /

U- -

STEP CHANGE) /

m 650 /

w  : /

, j) y 600

~

j f a  :

@ 550 iso I / /

/

w }_ I/

m -

/

n_ 500 _

y

/

b g 450 5

_ y

/ I/

,L,,,,,,,,, g,,,

I RANGE RATE S*EP*

5 400 _

,,, , , , , , , ,yn , ,, ,y gn

~ 60 F/4R -

~-

f~

350 ntPnEsEnis TsE sPEciPiEo insTAuTAncous DECREASE IN TEMPERATURE AND A THIRTY 300 'k MINUTE HOLD.

/
/

250 .

200 ' ' ' ' ' '

50 75 100 125 150 175 200 225 250 275 300 INDICATED RCS TEMPERATURE-Tc, DEG. F LP = -171.4 PSIA ART oT = +30*F 1/4t = 111.0*F 3/4t = 96.0 F 1

A-MPS-ER-002 Page 44 '

, , ,Y n , , , , , m, , ,.

l FIGURE 11 ARKANSAS NUCLEAR ONE UNIT 2 APPENDIX G PRESSURE-TEMPERATURE LIMITS 21 EFPY, COOLDOWN 1 ,0 0 0 ..

950 _

/[

900 :

=

/

850 :

800 :

=

1 u

4-

$ 750 ;

//

a 700 5

= /j w 650 :

m w

m 600 _

i ISO

,/

/

f

/  !

a.

m 550

/ /

it! 500 y -

/

g f f

/ /

5 m

450 5 ,

W 400

/

[ ~3 100*F/HR- [

a 350 : f w  :

y/

Q 300  :

o 5 250 i

f

/

200

^

150

/

100  :

50  :

0~'

50 75 100 125 150 175 200 225 250 275 300 INDICATED RCS TEMPERATURE-Tc, DEG. F ART AP = -171.4 PSIA 1/4t = 111.0*F AT = +307 3/41= 96.0 F 1

A-MPS-ER-002 Page 45

FIGURE 12 ARK ANSAS NUCLE AR ONE UNIT 2 HE ATUP CURVE,21 EFPY REACTOR COOL ANT SYSTEM PRESSURE TEMPER ATURE LIMITS 2500 . n = f. =. == nl _. . ..; r ; I: = l .. . . !=... b u L .. . :r i rz. r.q=_c-l INSERVICE " {~ ' '

~ ^7]2 3--+--- :~E - ----

EE :19=! T:Tc l' IT i

! ~- -

l-JE HYDROSTATICE EI"**E ci'.':1-E - E

---in .

_.___y -- n y..

____ :r =n : := =_ n..._t.

=: ;=  := Tt EST == q:;=- -- -r--e -- Er l =r: -). .=-dr ;.:- =t g

.  : = _ ; ; .. . , .

i= -i==r;=: g- _jr +- p __ _ r==g:n=.=-;r ==_. =.: = :- ;. t- . : :.  :: :x:;

_ = 2-

- r-1

- :L __ 1_.__. _c r + f___ ____gj h ^

w.

' " - P - = +l~ i= -

c_f ____ _j j r--" HE ATUP OED F =~E 4----l--tf:lr.=- -- - - - l - -

2000

' fp i t r- -s :=

l!

  • 4 ".~~'.=T__._

C"

~

Y f$. !f-- j'- - *- m:4 ;5@5 3-

<t

/

4 =;= ;. =

4 4.___/ _ _._  ; = ._ _a = =

4- - fr: -

"n. --==LOWEST  : ____. I_-- -# - _ NCORE CRITICdLU;r ==

l- __u .c=c-g EESERVICE - 2 =:=/_ 0 - - =

== TEMPER ATURE ;+H-M

~f:

1 x ---- d - '- _i, a .

f- 1 t.n 1500 == 180 F - I- I- _..y-,

c_ _. r_;_ ._

m *-/-___. /_-__ Y_

-4 w

m ==;

Jf---l, x __ . -

n. 1 g _. ,

E- '- # I I --~'--- + . . h. .i~-i. =.- _' -__

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-q l_____. +- _.l - _+ - . ..

f -*- -

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g I m

p (: .r-l -

-L_. _ - _ _ i _t m

t /- - - -

_. ._ . . . -?? ---

w 1000 .__-___;--

g n.

t -' .i _ -

f -+-  :

RCS TEMP. Hlu RATE E o

/ -

w -/ ^

E N ~~' - , ~ ~ ~ ~.

l

, - + - -

(-. + ALL RCS  :

O - . - __, - z - ~-__._

e " /* -- TEMPERATURES 80 F/HR  :

5 .

. . _ . _ . ___ ._3 j  : - .

500 i

' ~

t= -2 :=r  ; ;__ _ b = = =J = =i=-  : = i: = '_ f i_ -

f.=_1=_

__ . ._4 =

/ . . _ . .

  • --"~ -

>-+ -

4=5--t - -- =

m 'H---~' MIN. BOLT.UP, TEMP.- 70*F

_ _- , ---,_- _-.. y ._

... _. 5_ ._. _d==_2t =:= tr = .. . .j

- t._.- -_.

_ 4__..____ . _ t: =1. _ . ___.

23:2 5~

~~

!r_il.._ c._ MAXIMUM PRESSURE P25

~

==l =3 -

=a,=4. : = =:2 -.

JFOR.S_DC. == :*1 = t = =.=:n  :-u

= t _= t.: t- r:-t - - COPER ATI.ON ~=}=" ..

= t . _. .-! _ _. L _. .__;__. ..__3 =. : === ; ==: r= p a : . _

0 100 200 300 400 500 600 INDICATED REACTOR COOLANT TEMPERATURECT ' I Page 46 )

A-MPS-ER-002

l 1

1 1

)

FIGURE 13 ARKANSAS NUCLE AR ONE UNIT'2 COOLDOWN CURVE,21 EFPY REACTOR COOLANT SYSTEM PRESSURE TEMPERATURE LIMITS l

1 2500

.. j.. =r . .p.

c n.j..:.l ..._...:j. ,

+ . . . . . . -.I- ...j. f t.. j - .,

.;.-1;--;._..g ,j .. , -:

_ {. . _... . .

=;r 11NSERVICE ;a;.. .g. ._ . _ ....._ _.g ;.,. ... ..

=t=: HYDROSTATIC C: 7 r:q:-. 2:

.:=== ~ . . } =-: n =r=inu: =lre  : 1

= n: 2' *  !

[=}.= TES.T ._._

21=2=l:===t

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=== f=:=:=={&= =:;: u 2 l

= =r:  ::=- = n. .=. f: ==;= = = =.._

=p-- -~ -- --- : -

28, _ -t =_p==:-~~ --+=-~ =] :=u 1l=* =

~

_4._.p=:=:{=.=.=

.=--e----  :

- " st = t: =j = :; = ; = n Or.= _ = ===; --- p-- ----i r:= .^ ~;-- :: ===~~ 1

'~ * - - ~ ~

2000

4

~ ~ ~ ~

_ . . - . _ _ _ . = _ "~-._ _. . !~2-~ __- "-7COOLDOWN--+ =3~ ~*--- i-- --

mi __,

_ _. .._._.L._ _ . _.._:.

-- ~ =r

__i . _- g ::---+

_r

=q=,

j __ __

. . . ._._3

.g .-.--_. - . _+-. -

_4._..-p._._.._q_.__..__.-_

4

- = LOW, EST _- - _.. _ _ . ..

=-
SERVICE, -.- - _ - ,.-_-- -. .-

m -

t. __ 3 ,,_

cc - _ . .._

3 = TEMPER ATURE M._.-,_ '

m -

~-'_--__~_f

~ ~ ' ~ ' - "

-:180T m 1500 w -

._.j _. I:=-1 ___.,

cc c.

_+_ /=-/. 1

~

a W

._  ?? _':f '

y g -

~ 7:-~ = __i

_,.__f

-.+. __

p- ,

.f.t - f _

m _ f._i_ _ / ,

.-- RCS TEMP. C/D RATE STEP *

.m~

@_ ,4-./ r .

c. 1000 _.__- __.,..

.o-

=

T < 150T 25T/HR 12.5T/1/2 HR W

g _- - ,, .-,- .

150T 5 T s 230T 60T/HR 30T/1/2 HR --

T > 230T 100T/HR *50T/1/2 HR

O._ _. c=  :

(' 1

.e-. ._ -

Represents the specified instantaneous decrease ECOOLDOWN .

~

.- in temperature and a thirty minute hold.

,,  :- --i

._ e 500 ___, , p _. - g . ___;.__..,_ ;_._. _ ,_

,L

= v._ w

.. 4 _, ----r-----

- " - - " - ~

XFZ- MIN.- - _ . . _ _

BOLTUP TEMP. 70T w -

i - "_- ~* .y

_ _ . = :-- _. 2- p -_- ;=7_

,_,._ .__.;r_ g. .=.t: n:__ _ . , =

_  :=t= i ._ <

.= := .' _.._ _n t=1---" . - - _ .

-_____y--:___g- :.

L' 1 = -: ..

_ . - . . . __j'M AXIMUM PRESSURE _ ~ ~ " ~ ' .:2-u_ ;- __

q=:-- ---2==-4

FOR SDC OPERATION ~ =

. .g-.. 2 =  :. in= = =

=,t - E=;=- 21 ==== rl = =:'  : . . =n= = trr =.i = . =:: = r

-t= =>=r i :.j=: :t r ==

-- n =l - - -i - - 3 r *== =: =

0 100 200 300 400- 500 600 INDICATED REACTOR COOLANT TEMPERATURECT ' Y i

A-MPS-ER-002 Page 47

.l l=

. - - . . _ . _ = _ . . - - - . - . . . - . . - . .- . . .. - . -

. . . . ~ . . ~. ... .. ... . . - _ . . . . .- -. . - . - _ . ~, . - - . . - - . - - . _ - . .

ATTACilMENT 2 ANO-2 REACTOR VESSEL BELTLINE MATERIALS PROPERTIES Il 1

l-l

[

. iW.,Y :4 '51 8: 1'4 FROM PRODUCT LEVELOPMENT PAGE.002 ABB ASEA BROWN BOVEMI May 8, 1991 MCC-91-229 Mr. Jay Miller Entergy Operations, Inc.

Rte. 3 Box 137G Generation Support Building Russellville, AR 72801

Subject:

ANO-2 Reactor Vessel Beltline Materials Properties

References:

1) Report No. A-MPS ER-002, " Final Report on Reactor Vessel Appendix G Pressure-Temperature Limits for Arkansas Nuclear One - Unit 2 for 21 EFPY," May 1991, combustion Engineering, Inc.
2) P. F. Lawlor, " Reply to NRC fluestionnaire for Pressure Vessel Fracture Toughness Properties of ANO-2," March 9, 1978, Combustion Engineering. Inc. Report A MCM-106.

(Basis for D. H. Williams (ANO) to J. F. Stolz (NRC), AP&L Letter 2CAN067810, dated June 13, 1978, " Pressure vessel Fracture Toughness Properties).

3) J. T. Enos (ANO) to G. W. Knighton (NRC), " Arkansas Nuclear One - Unit 2, Docket No. 50-368, License No. NPF-6.

Pressurized Thermal Shock Submittal Required by 10CFR50.61.

" January 22, 1986, Arkansas Power & Light Company.

Dear Mr. Miller:

In response to your request of May 2, 1991, a comparison was made of the ANO-2 vessel beltline data presented in Re rerence 1 and 2. Table 1 of Reference 1 provides chemical content and initial RTNDT for each of the beltline plates and welds; these data were used to generate adjusted reference temperatures for operation to 21 EFPY. Reference 2 provided comparable beltline data for use in responding to NRC questions in 1978 regarding fracture toughness properties. Reference 2 was used in part by Arkansas Power & Light to prepare Reference 3 in response to 10CFR50.61.

Differences between Table 1 of Reference 1 and the data presented in Reference 2 are as follows:

ABB Combustion Engineering Nuclear Power en E we n. ggym 3.,genen

  • tax W9297 COMBE *4 W50m Wmet. C enec'ca 060%C$00

, , %' 14 '91 5:15 F R Of1 PRODUCT D E V E L O P t1E tit F%GE.000 MCC-91-229 Page 2 a) Initial RT NOT f Plates - for both sets, the longitudinal Charpy data were converted to trans"crse data, and then to RI NDT following NRC guidelines (MTE8 5-2). The newer values (Reference

1) resulted from a more rigorous application of MTEB 5-2 to the data and the use of baseline surveillance data for the surveillanca program plate (TR MCD-002, March 1976).

b) Initial RT NDT f Welds F r w id seams 2-203 A, B. & C and 3-203 A, B, & C, RT NOT was estimated based on MTEB 5-2 guidelines for the 1978 report. C-E subsequently developed a generic value of initial RT for certain automatic submerged arc weldments (CEN-NDT 189, December 1981); the generic value was used in the 1991 report (as well as in Reference 3). For weld scam 9 203, three applicablu measurements of initial RTNDT are available. Reference 2 used the higher of two welding material qualification test results; Reference 1 used the value from the surveillance program weld baseline test results (TR-MCD-002, March 1976). Both sets of measurements are valid, but the Reference 1 approach is more conservative. (Note that whichever value is used, neither PTS nor PT-LTOP is affected adversely.)

c) Chemical Content of Welds - Weld seams 2-203 A, B, & C, and 3-203 A, B, & C, were fabricated using the same heat of wire and type of flux (but different flux lots), for which there were four chemical analyses reported on the deposited weld wire ranging from 0.04 to 0.05% c.opper. Reference 2 used two of those analyses, whereas Reference 1 based copper content on all four analyses. Nickel content was reported for an analysis representing weld seams 3-203 A, B & C, and this value was used for both of the longitudinal (2-203 and 3-203) welds in Reference 1 because the same weld wire heat was employed. Weld seam 9-203 was fabricated using the same consumables as the surveillance program weld; therefore, the

_.-.~-__m._

, tieY 14 '91 8:16 FROM P R O D'J C T DEVELOPMENT PAGE 004 MCC-91-229 Page 3 copper and nickel content of the surveillance weld was used for weld 9-203 in Reference 1. A weld deposit analysis reporting copper, but not nickel, was used in Reference 2.

The initial RTNDT values and chemical content data are summarized in Tables 1 and 2, respectively. The values from Reference I were independently reviewed and verified in accordance with ABB/CE quality assurance procedures (QAM-100 and QAM-101). Furthermore, R.tference 1 data were derived using a more rigorous approach and more complete data than in Reference 2, and, therefore, are recommended for use in ANO-2 vessel materials evaluations.

Sincerely, ABB COMBUSTION ENGINEERING NUCLEAR POWER

  • .7. b L S. T. Byrne STB /b cc: P. J. Hijeck N. S. Zavacky W. R. Gahwiller S. M. Schloss ht/

R. P. O'Neill R. C. Sykes (RSSM)

,h 'fi la '91 3: 16 FROM PRODUCT DEVELOPMENT FAGE,005 MCC-91 229 Page 4 Table 1 INITIAL RT NDT OF ANO-2 BELTLINE MATERIALS Reported Initial RTNUT Material Reference 1 Reference 2 Reference 1

("F) ("F) ("F)

C-8009-1 -26 5 5 C-8009-2 0 10 10 C-8009-3 0 35 0 C 8010-1 12 10 10 C-8010-2 -28 -20 -20 C-8010-3 -30 -20 -20 2-203 A, B, C -56 10 -56 3-203 A, B, C -56 10 -56 9-203 -10 -40 -56

M Q/ l -1 '91 5: 16 FROM PPODUCT DEVELOPMENT PAGE.006 MCC 91-229 Page 5 Table 2 CHEMICAL CONTENT OF ANO 2 BELTLINE MATERIALS Reported Chemical Content Matnrial Refnrenen 1 Reference ? Reference 3 C_y(l) M C.y M [it M C-8009-1 .12 .63 C-8009-2 .08 .59 NO NO C-8009-3 .08 .60 CHANGE CHANGE C-8010-1 .08 .59 C-8010-2 .07 .66 C-8010-3 .07 .65 2-203 A, B, C .05 .18 .05 N/R(2) .05 .18 3-203 A, B, C .05 .18 .04 N/R .04 .18 9-203 .05 .08 .05 N/R .05 .18 (1) Cu (copper) and Ni (nickel) content in weight percent.

(2) N/R - Not Reported

++ TOTAL PAGE.OO6~++

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