ML20106A670
| ML20106A670 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 09/19/1984 |
| From: | Varga S Office of Nuclear Reactor Regulation |
| To: | Otoole J CONSOLIDATED EDISON CO. OF NEW YORK, INC. |
| Shared Package | |
| ML20106A673 | List: |
| References | |
| NUDOCS 8410110554 | |
| Download: ML20106A670 (15) | |
Text
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fW UNITED STATES IT
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NUCLEAR REGULATORY COMMISSION
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$ASHINGTON, D. C. 20555 g s..Jk h.f f
%,',,,, 7 September 19,.1984 Docket No. 50-247 Mr. John D. O'Toole Vice President Nuclear Engineering and Quality Assurance Consolidated Edison Company of New York, Inc.
4 Irving Place New York, New York 10003
Dear Mr. O'Toole:
SUBJECT:
REACTOR VESSEL FLIW AT THE INDIAN POINT NUCLEAR GENERATING PLANT,, UNIT NO. 2 (IP-2)
By letter dated Sepember 7, 1984 you submitted the fracture mechanics evaluation regarding the above subject. Our evaluation is based upon the review of the Westirghouse Report WCAP-10651, " Fracture P?chanics Evaluaticn of Inservice Inspection Indication, Indian Point Unit 2 Reactor Vessel".
In order to determine the safety margin between the ASME Code allowable flaw and the potential flaw in the IP-? beltline, we request that you respond to the questions and concerns which r contained in Attachment 1.
In addition, attachment 2, Draft Regulatory o ide 1.99, Rev. 2, dated July 23, 1984, is the staff's most "up-to-date" method of estimating the amount of irradiation damage to base metal and weld metal.
Although the Draf t Regulatory Guide has not been formally approved, its effect upon the safey margins for the potential flaw in the IP-2 reactor vessel should be evaluated.
You earliest response is requested.
The reporting and/or recordkeeping requirements of this letter affect fewer than ten respondents; therefore, OMB clearance is not required under P.L.96-511.
Sjncerely,-
Q hk k ven A. Varga, Branch Chief Operating Reactors ranch #1 Division of Licensing
Enclosure:
As stated cc w/ enclosure:
/0 // Ofd Ye4 See next page e-w w cwn1
_---a
4 l'r. John D. O'Toole In'dian Point Station, Unit 1 Consolidated Edison Company Indian Point Nuclear Generating Unit 2 of New York, Inc.
cc: Mayor, Village of Buchanan Regional Radiation Representative 236 Tate Avenue EPA Region II Buchanan, New York 10511 26 Federal Plaza New York, New York 10007 Michael Blatt Director Regulatory Affairs Director, Technical Development Consolidated' Edison Company Programs of New York, Inc.
State of New York Energy Office Braodway and Bleakley Aveunues Agency Building 2 Buchanan, New York, 10511 Empire State Plaza Albany, New York 12223 Robert L. Spring Dr. Lawrence R. Quarles l
Nuclear Licensing Engineer Apartment 51 Consolidated Edison Company Kendal at Longwood of New York, Inc.
Kennett Square, PA 19346 1
4 Irving Place New York, New York 10003 Mr. Charles W. Jackson U.S. Nuclear Regulatory Commission Vice President, Nuclear Power Post Office Box 38 Consolidated Edison Company Buchanan, NY 10511 of New York, Inc.
j Broadway and Bleakley Avenues Brent L. Brandenburg Buchanan, New York 10511 Assistant General Counsel Consolidated Edison Company Mr. Frank Matra of New York, Inc.
Resident Construction Manager 4 Irving Place - 1822 Consolidated Edison Company New York, NY 10003 of New York, Inc.
Bro dway and Bleakley Avenues Regional Administrator - Region i Buchanan, New York 10511 U.S. Nuclear Regulatory Commission 631 Park Avenue Ezra I. Bialik King of Prussia, PA 19406 Assistant Attorney General.
Environmental Protection Bureau Carl R. D' Alvia, Esquire New York State Department of Law Attorney for the Village of 2 World Trade Center Buchanan, New York New York, New York 10047 Ms. Ellyn Weiss Sheldon, Harmon and Weiss 1725 1 Street, N.W., Suite 506 Washington, DC 20006 Senior Resident Inspector U.S. Nuclear Regulatory Commission Post Office Box 38 Buchanan, NY 10511
Corcolidated Edisc-Company of New York Indian Point Urrit Nc. 2 (IP-2)
Docket No. 50-247 To demonstrate the safety margins against brittle fracture for the potential flaw indica +1on in the IP-2 reactor vessel beltline, the licensee has pro-vided to the staff a fracture mechanics analysis which is contained in Westinghouse Report WCAP 10651 (Proprietary Class 2), " Fracture Mechanics Evaluation of Inservice Inspection Indication Indian Point Unit 2 Reactor Vessel." The Westinghouse report was submitted for staff review in a letter from J. D. O'Toole to S. A. Varga dated September 7, 1984. The following questions and comments relate to the analysis documented in the report.
1.
The events analyzed in determining the ASME Code allowable flaw indication should include the Turkey Point Unit 4 LTOP event which occurred on November 28 and 29, 1981.
Based upon the frequency of this type of event in all operating PWRs, the licensee should deter-mine whether the event is considered upse,c or emergency and faulted.
In analyzing this event for the IP-2 vessel, the pressures and temperatures to be considered should be those which would occur if the event were terminated by lifting of the IP,2 Pressurizer Safety Valve.
If the Turkey Point set of events had occurred at 1P-2, without operator action to terminate the transient, how much time would it take for the pressure to reach the Pressurizer Safety Valve set point?
2.
If the flaw indication were located in the adjacent HAZ or base metal (Plate B 2003-1), what would be the ASME Code allowable flaw indication during normal, upset, test, emerg'ency and faulted conditions?
3.
Compare the end-of-life RT and ASME Code allowable flaw indication NDT using the amount of increase in RTNDT predicted by the "Guthrie" fonnula in Comission Report SECY 82-465 and the model in Draft Regulatory Guide 1.99 Rev. 2 (Attachment 2).
- 4.
Indicate the references and heat numbers, and lot numbers for the weld wire and flux for each weld chemistry in Table 3-1.
5.
Indicate the heat number and lot number for the weld wire and flux
.~or the weld in Table 3-2.
6.
Figure 3-2 indicates that the current fast neutron exposure at the 18 2
inside surface - 345* Azimuthal Angle is 1.5 x 10 n/cm. Consolidated
. Edison has reported to the staff in a telecon that after completing the sixth fuel cycle using a low leakage core, the current fast neutron 18 2
exposureattheinside' surface-345'AzimuthalAngleis1.77x10 n/cm,
ExplainthedifferencI'intheseestimatesandusethemoreaccurate number in the analysis.
e S
5 e
July 23, 1954.
Office of Nuclear Regulatory Research t
DRAFT REGULATORY GUIDE 1.99, REVISION 2 I
RADIATION DAMAGE TO REACTOR VESSEL MATERIALS A.
INTRODUCTION General Design Criterion 31, " Fracture Prevention of Reactor Coolant Pressure Boundary," of Appendix A, " General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50, "L,1 censing of Production and Utilization Facili-ties," requires,inpart,(Ntthereactorcoolantpressureboundarybe designed with sufficient margin to ensure that, when stressed under operating, maintenance, testing, and postulated accident conditions:
(1) the boundary behaves in a,nonbrittle manner, and (2) the probability of rapidly propagating fracture is minimized.
Appendix G, " Fracture Toughness Requirements," and Appendix H, " Reactor Vessel Material Surveillance Program Requirements," which implement, in part, Criterion 31, necessitate the calculation of changes in l
fracture toughness of reactor vessel material's caused by neutron radiation throughout the service life.
This guide /-scribes general procedures acceptable to the NRC staff for calculatlog the effetts of neutron radiation damage to the low-alloy steels currently used for light water-cooled reactor vessels.
The Advisory Committee on Reactor Safeguards will be consulted concerning this guide.
B.
DISCUSSION The principal examples of NRC requirements that necessitate calculation of radiation damage are:
1.
Paragraph V.A. of Appendix G requires:
"The effects of neutron radiation...are to be predicted from the results of pertinent radiation ef fect studies...."
This guide provides such results in the form of calculative procedures that are acceptable to the NRC.
07/23/84 i
RG 1.99 REV 2 l
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2.
Paragraph V.B. of Appendix G desc,ribes.the basis for setting the
~
upper limit for pressure as a function of temperature during heatup and cooldown for a given service period in terms of the predicted value of the adjusted reference temperature at the end of the service period.
l 3.
The definition of reactor vessel beltline given in Paragraph II.F.
of Appendix G requires identification of:
... regions of the reactor vessel that are predicted to experience sufficient neutron radiation damage to be
)
considered in the selection of the most limiting material...." Paragraphs III.A.
and IV. A.1. specify the additional test requirements for beltline materials that supplement the requirements for reactor vessel materials generally.
4.
Paragraph II.B. of Appendix H incorporates ASTM E185 by reference.
Paragrhph 5.1 of ASTM E185-82 requires that the materials to be placed in sur-veillance be those that may limit operation of the reactor during its lifetime, i.e., those expected to have the highest adjusted reference temperature or the lowest Charpy upper-shelf energy at end of life.
Both measures of radiation damage must be considered.
In Paragraph 7.6 of ASTM E185-82 the requirements for number of capsules and withdrawal schedule are based on the calculated amount o_f_ radiation damage at end of. life.
The two measures of radiation damage used in this guide are obtained from the results of th'e Charpy V-notch impact test.
Appendix G to 10 CFR Part 50 requires that a full curve of absorbed energy versus temperature be obtained through the ductile-to-brittle transition temperature region.
The adjustment of the reference temperature, ARTNDT, is defined in Appendix G as the tempera-ture shift in th'e Charpy curve for the irradiated material relative to that
~
for the unirradiated material, :.easured at the 30-foot pound energy level. The second measure.of radiation damage is the decrease in'iIle Charpy upper-shelf energy level,'which is defined in ASTM E185-82.
Revision 2 of this guide updates the calculative procedures for the adjustment of reference tempera-ture; however, calculative procedures for the decrease in upper-shelf energy are unchanged, because the preparatory work had not been completed in time to include them in Revision 2.
07/23/84 2
RG 1.99 REV 2
~
The basis for equation (2) for ART surface, given in Position C.I.a.2.
NDT of this Guide, is cdntained in publications by G. L. Guthrie1 and G. R. Odette.2 Both authors used as their data base surveillance data from commercial power reactors, but their analysis techniques were different.
Both authors recommended the following:
(1) separate correlation functions for weld and base metal, (2) the function should be the product of a chemistry factor and a fluence factor, (3) the parameters in t..e chemistry fac, tor should be the elements, copper and nickel, and (4) the fluence factor should provide a trend curye slope of atout 0.25 to 0.30 on log-log paper at 1029 n/cm3 (E>l MeV), steeper at low fluences and flatter at high fluences.
Position:C.1.a. is a blend of the correlation functions presented by the two authors.
Some test. reactor data were used as a guide in establishing a cutoff for the chemistry facto'r for low-copper materials. The da~ta base for Position C.1.b. is t' hat given*by Spencer H. Bush.3 The measure of fluence used herein is the number of neutrons per square centimeter having energies greater than 1 million electron volts (E>l MeV).
The differences in energy spectra at the surveillance capsule and the vessel inner surface locations do not appear to be great enough to warrant the use of
(
a damage function such as displacements per atom (dpa)4 in the analysis of the surveillance data base.5 2G. L. Guthrie, "Charpy Trend Curves B; sed on 177 PWR Data Points," from LWR Pressure Vessel Surveillance Dosimetry Improvement Program, Quarterly Progress Report April 1983 - June 1983, HanfoM Engineering Development Laboratory.
NUREG/CR-3391, Vol. 2, HEDL-TME 83-22.
2G. R. Odette and P. M. Lombrozo, " Physically Based Regression Correlations h
of Embrittlement Data From Reactor Pressure Vessel Surveillance Programs,"
EPRI NP-3319, Final Report, January 1984, Prepared for Electric Power Research Institute.
3 Spencer H. Bush, " Structural Materials for Nuclear Power Plants," 1974 ASTM Gillett Memorial Lecture, published in ASTM Journal of Testing and Evaluation, November 1974, and its addendum, "Radfation Damage in Pressure Vessel Steels for Commercial Light-Water Reactors."
4 ASTM E 693-79, " Standard Practice for Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements Per Atom (dpa)."
5W.N. McElroy, Editor. " LWR Power Reactor Surveillance Physics - Dosimetry Data Base Compendium," NUREG/CR 3319, HEDL TME 84-2 March 1984.
e 07/23/84 3
RG 1.99 REV 2
However, the neutron energy' spectrum d,oes change significantly with location in the vessel wall; hence for calculation of attenuation of radiation damage through the vessel wall, a damage function should be used to de rmine ART versus radial distance into the wall.
The most widely accepted damage NDT function at this time is dpa and the attenuation formula (3) given in Position C.1.a.(2), is based on the attenuation of apa through the vessel wall.
Sensitivity to neutron radiation damage may be affected by elements otiier
Revisions 0 and 1 of this guide had a phosphorus term in the chemistry factor, but the studies upon which this revision was based found other elements such as phosphorus to be of secondary importance, i.e.,
including t, hem in the analysis did not produce a significantly better fit of the data.
Scatter in the data base used for this guide is relatively significant, as evidenced by the fact tha't' the standard deviations for Guthrie's derived formulas are 28 F for welds and 17*F for base metal, despite extensive statis-tical analysis.
Thus, the use of surveillance data from a given reactor (in place of the calculative procedures given in this guide) requires considerable engineering judgment to evaluate the credibility of the data and assign suitable margins.
When surveillance data from the reactor in question become available, the weight given to them relative to the information in this guide should depend on the credibility of the surveillance data as judged by the following criteria:
1.
Materials in the capsules should be those judged most likely to be controlling with regard to radiation damage according to the provisions of this guide.
2.
Scatter in the plots of Charpy energy versus Temperature for the irradiated and unirradiated conditions should be small enough to permit the determination of the 30 ft-lb temperature and the upper shelf energy unambig-uously.
3.
When there are two or more surveillance data from one reactor, the scatter of ART values about a best fit line drawn as described in NDT Position C.2.a. normally should be less than 28 F for welds and 17 F for base metal.
Even if the fluence ra.nge is large (two or more orders of magnitude) the scatter should not exceed twice those values.
Even if the data fail this criterion for yse in shift calculations, they may be credible for determining 07/23/84 4
RG 1.99 REV 2
~
decrease in upper shelf energy if the upper, shelf can be clearly determined, following the definition given in ASTM E 185-82.
4.
The irradiation temperature of the Charpy specimens in the capsule should match vessel wall temperature at the cladding-base metal interface within 125*F.
5.
The surveillance data for the correlation monitor material in the capsule should fall within the scatter band of the data base for that material.
In using plant surveillance data to develop a plant-specific relationship of ARTNDT.t fluence, it was deemed advisable (because of scatter) to determine the slope, i.e., t'he fluence factor, from other than the plant data.
- Instead, Equation 2, paragraph C.1.a.(2), is to be fitted to the plant surveillance data. Of several possible ways to fit such data, the method that minimizes the sums of the squares of the'erNors was cnosen somewhat arbitrarily.
Its use is justified in part by the fact' that "least squares" is a common method for curve fitting.
Also, when.there are only two data points, the least squares method gives greater weight to the point with the higher ARTNDT; which seems reasonable.for fitting surveillance data, because generally that datum will be, the more recent on1 and therefore will represent more modern procedures.
C.
REGULATORY POSITION 1.
SURVEILLANCE DATA NOT AVAILABLE When credible surveillance data from the reactor in question are not available, calculation of neutron radiation damage to the beltline of reactor vessels of light water reactors should be based on the'iollowing p'rocedures, within the limitations in Paragraph C.1.c.:
a.
The adjusted reference temperature (ART) for each material in the beltline is given by the following expression:
ART = Initial RTNDT + ARTNDT + Margin (1)
(1)
" Initial RT
" is the reference temperature for the NDT unirradiated material as defined in Paragraph NB-2331 of Section III of the ASME Boiler and Pressure Vessel Code.
In cases where measured values of Initial RT f r the material in question are not available, generic values NDT 07/23/84 5
RG1.99REh2
for that class 6 of material may b'e used if there are sufficient test results to establish a mean and standard deviation for the class.
Additional guidance for the estimation of initial RT is given in the Standard Review Plan, NDT NUREG-0800, Section 5.3.2.
(2)
" ART
" is the mean value of the adjustment in reference NDT temperature caused by irradiation and should be calculated as follows:
ART surface = [CF]f(0.28-0.10 log f)
(2)
NDT The chemistry factor, "CF,"
F, a function of copper and nickel content, is given in Table I for welds and Table II for base metal (plates and forgings).
Linear interpolation is permitted.
In Tables I and II, " Percent Copper" and " Percent Nickel" are the best-estimate values for the material, which will normally be the mean of the measured values for a plate or forging or for weld samples made with the weld wire heat number that matches the critical vessel weld.
If such values are not available, the upper limiting values given in the material specifications to which the vessel was built may be used.
If not available, conservative estimates (mean plus one standard deviation) based on generic data may be used if justification is provided.
If there is no information available, 0.35%
copper and 1.0% nickel should be assumed.
The fluence, "f," is the calculated value of the neutron fluence at the inner surface of the vessel at the location of the postulated defect, n/cm>
(E>l MeV) divided by 1019 The fluence factor, f.28 - 0.10 log f,isdetermifiedbycalculationor 0
from Figure 1.
To calculate ART at any depth, (e.g., at 1/4T or 3/4T), the following NDT attenuation formula should be used:
NDT "L NDT 5For the welds with which this guide is concerned, for estimating Initial RTNDT, class is generally determined by the welding flux; for base me'.al, by the ASTM Standard Specification.
07/23/84 6
RG 1.99 REV 2
E e
where "h' (in inches) is_ the depth.into the. vessel wall. measured from the
~
f.
vessel.. inner. surf ace.
p.
(3) " Margin" is the quantity, 'F, that is.to be added to L
I~ obtain' conservative, upper-bound values of adjusted reference temperature for
(
the calculations required by Appendix G, 10 CFR Par l. 50.
1 Margin = 2'4 oz+03-(4)
I a
If a measured value of-Initial RT f r the material in question is NDT used, o may be taken as zero.
If a generic value of Initial RT is used, y
NDT o should be obtained from the same set of data (see paragraph C.I.a.(1)).
.I g
The standard deviations for ART
""A, are 28'F for welds and 17'F for base NDT' metal, except o need not exceed 0.50 times the mean value of ART surface.
g NDT b.
Charpy upper'-s'helf energy should be assumed to decrease as a function of fluence and copier content as indicated in Figure 2.
Linear
-. interpolation is permitted.-
c.
Application of the foregoing procedures should be subject to the following limitations:-
(1) The procedures. apply-to those grades of SA-302, 336, 533, and.508 steels having minimum specified yield strengths of 50,000 psi and under and to their welds and heat-affected zones.
(2) The procedures are valid for a nominal irradiation tempera-ture of 550'F.
Irradiation below 525*F should be considered to produce greater damaga, and irradiation above 590*F may be considered to produce less damage.
The correction factor used should be justified by reference to actual data.
(3) Application of these procedures to fluence levels or to copper or nickel content beyond the ranges given in Fi'ure I and Tables I g
and II or to materials having chemical compositions beyond the range found in the data bases used for this guide, should be justified by submittal of data'.
2.
SURVEILLANCE DATA AVAILABLE i
When two or more credible surveillance data as defined in the Discussion, Section B, become available from the reactor in question,.they may be used to determina the adjusted reference temperature and the Charpy upper-shelf energy of.the beltline materials as described in-the following Paragraphs a. and b.,
respectively.
07/23/84 7_
RG 1.99 REV 2
a.
The adjusted reference tempe,rature may be obtained by first fitting the surveillance data using Equation 2, paragraph C.1.a.(2), to obtain the relationship of ART surface to fluence.
To do so, calculate the chem-NDT istry factor, "CF," for the best fit as follows.
Multiply each measured ART by its corresponding fluence factor, sum the products and divide by the NDT sum of the squares of the fluence factors.
The resulting value of CF when entered in Equation 2 will give the relationship of ART surface to fluence NDT that fits the plant surveillance data in such a way as to minimize the sums of the squares of the errors.
To calculate the Margin in this case, use the procedure given in paragraph C.1.a.(3),,except the values given there for o may be cut in half.
g If this procedure gives a higher va.lue of adjusted reference temperature than that given by using th'e 'p'rocedures of paragraph G.1.a, the former should be used if the surve' illa)ce ' data meet the criteria for credibility.
b.
The decrease in upper-shelf energy may be obtained as follows.
Plot the reduced plant surveillance data on Figure 2 of this Guide.
Fit the data with a line drawn parallel to the existing lines as the upper bound of all the data. This line should be used in preference to the existing graph.
3.
REQUIREMENT FOR NEW PLANTS For beltline materials in the reactor vessel foy.a new plant, the content of residual elements such as copper, phosphorus, sulfur, and vanadium should be controlled to low levels.
The copper content should be such that the calculated adjusted reference temperature at the 1/4T position in the vessel wall at end of life is less than 200*F.
D.
IMPLEMENTATION The purpose of this section is to provide information to applicants and licensees regarding the NRC staff's plans for utilizing this regulatory guide.
Except in those cases in which the applicant proposes an acceptable alternative method for complying with specified portions of the Commission's regulations, the positions described in this guide will be used by the NRC staf f as follows:
07/23/84 8
RG 1.99 REV 2
- 1. ' The method described in regulatory positions C.1 and C.2 of this guide will be used in evaluating all predictions of radiation damage called
'for in Appendices G and H to 10 CFR Part 50 submitted on or after (60 days j
after publication); however, if an applicant wishes to use the recommendations i
f regulatory position C.1 and C.2 in developing submittals before (60 days o
after publication), the pertinent portions of the submittal will be evaluated
=
on the basis of this guide.
2.
Following publication of this guide in final form, the owners of all operating reactors and all applicants for an operating license should promptly review the basis for the pressure-temperature limits in their Technical Speci-fications for consistency with Positions C.1 or C.2 as appropriate.
Those for i
whom the allowable operating period has been reduced or has already expired,
).
when judged by the criteria OI Revision 2, should promptly revise their operat-I ing procedures, as a'pproprist'e, to conform with the criteria of Revision 2 of l
this guide and submit the appropriate revision to their Technical Specifications
{
within six months of the date of publication of. Revision 2 of this guide in final form. '
i Those for whom the allowable operating period has been extended, when j
judgedbythecriteriaofRevision2,shouldskbmittheappropriaterevision to their TSs no later than 90 days prior to the expirat'on of their current operating period.
3.
The recommendations of regulatory position C.3 are unchanged from l
those used to evaluate construction permit applications docketed on or after June 1, 1977.
,de 1
I 1
l f
07/23/84 9
RG 1.99 REV 2
TAB 1E'I.
CHEMISTRY FACTOR FOR WELDS, *F
Wt. %
0 0.20 0.40 0.60 0.80 1.00 1.20 p
0 20 20 20 20 20 20 20.
- 0.01 20 20 20 20 20 20 20 0.02 21 26 27 27 27 27 27 0.03 22 35 41 41 41 41 41 0.04 24 43 54 54 54 54 54 O.05 26 49 67 68 68 68 68 0.06 29 52 77 82 82 82 82 0.07 32 55 85 95 95 95 95 0.08 36 58 90 106 108 108 108 0.09 40 61 94 115 122 122 122 0.10 44 65' 97 122 133 135 135
. 0.11 49 -
- 68..
101 130 144 148 148 0.12 52 72 103 135 153 161 161 0.13 58 76 106 139 162 172 176 0.14 61 79 109 142 168 182 188 0.15 66 84 112 146 175 191 200 0.16 70 88 115 149 178 199 0211 0.17 75 92 119 151 184 207 221 0.18 79 95 122 154 187 214 230 0.19 83 100 126 157 191 220 238 0.20 88 104 129 160 194 223 245 0.21 92 108 133 164 197 229 252 0.22 97 112 137 167 200 232 257 0.23 101 117 140 169 203 236 263 0.24 105 121 144 173 206 239 268 0.25
.110 126 148 176 209 243 272 0.26 113 130 151 180 212.~
246 276 0.27 119 134 155 184 216 249 280 0.28 122 138 160 187 218 251 284 0.29 128 142 164 191 272 254 287 0.30 131 146 167 194 225 257 290 O.31 136 151 172
198 228 260 293 0.32 140 155 175 202 231 263 296 0.33 144 160 180 205 234 266 299 0.34 149 164 184 209 238 269 302 0.35 153 168.
187 212 241 272 305 0.36 158 172 191 216 245 275 308 0.37 162 177 196 220 248 278 311 0.38 166
~182 200 223 250 281 314 0.39 171 185 203 227 254 285 317 0.40 175 189 207 231 257 288 320 07/23/84 10 RG 1.99 REV 2 l
g
F TABLE II CHEMISTRY FACTOR FOR 5ASE METAL, 'F
Wt. %
0 0.20 0.40 0.60 0.80 1.00, 1.20
.0 20 20 20 20 20 20 20 0.101 20 20 20 20 20 20 20 0.02 20 20 20 20 20 20 20 0.03 20 20 20 20 20 20 20 0.04 22 26 26 26 26 26 26 0.05 25 31.
31 31 31 31 31
- 0.06 28 37 37 37 37 37 37
- 0.07 31 43 44 44 44 44 44 0.08 34 48 51 51 51 51 51 0.09 37 53 58 58 58 58 58 0.10 41' 58 '
65 65 67 67 67 0.11 45 -
62,.
72 74 77 77 77 0.12 49 67 79 83 86 86 86 0.13 53 71 85 91 96 96 96 0.14 57 75 91 100 105 106 106 0.15 61 80 99 110 115 117 117 0.16 65 84 104 118 123 125 125 0.17 69 88 110 127 132 135 135 0.18 73 92 115 134 141 144 144 0.19 78 97 120 142 150 154 154 0.20 82 102 125 149 159 164 165 0.21 86 107 129
'155 167 172 174 0.22 91 112 134 161 176 181 184 0.23 95 117 138 167 184 190 194 0.24 100 121 143 172 191 199 204 0.25 104 126 148 176
.199 208 214 O.26 109 130 151 180 205.~
216
'221 0.27' 114 134 155 184 211 225 230 0.28 119 138 160 187 216 233 239 0.29 124 142 164 191 221 241 248 0.30 129 146 167 194 225 249 257 0.31 134 151 172 198 228 255 266 0.32 139 155 175 202 231 260 274 0.33 144 160 180 205 234 264 282 0.34 149 164 184 209 238 268 290 0.35 153 163 187 212 241 272 298 1
0.36 158 173 191 216 245 275 303 0.37 162 177 196 220 248 278 308 0.38 166 182 200 223 250 281 313 0.39 171 185 203 227 254 285
-317 0.40 175 189 207 231 257 288 320 l
07/23/84' 11 RG 1.99 REV 2
___ _ _ _ _ _ _. _ _