ML20195E046
| ML20195E046 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 06/09/1988 |
| From: | BALTIMORE GAS & ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20195E036 | List: |
| References | |
| NUDOCS 8806230281 | |
| Download: ML20195E046 (19) | |
Text
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DESIGN FEATURES VOLUME 5.4.2 The total water and steam volume of the reactor coolant system is 10,614 460 cubic feet at a nominal T avg of 532 F.
5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure 5.1-1.
5.6 FUEL STORAGE CRITICAllTY - SPENT FUEL 5.6.1 The spent fuel storage racks are designed and shall be maintained with a minimum 10 3/32" x 10 3/32" center-to-center distance between fuel assemblies placed in the storage racks to ensure a keff of 1 0.95 with the storage pool filled with unborated water. The kg gf of s 0.9S includes the conservative allowances for uncertainties described in Section 9.7.2 of the FSAR. The maximum fuel enrichment to be stored in the fuel pool will be 4rl'f b weight percent.
CRITICALITY - NEW FUEL 5.6.2 The new fuel storage racks are designed and shall be maintained i with a nominal 18 inch center-to-center distance between new fuel assemblies I such that keff will not exceed 0.98 when fuel having a maximum enrichment of i f,0bl4eight percent U-235 is in place and various densities of unborated water '
'are assumed including aqueous foam moderation. The keff of f 0.98 includes the conservative allowance for uncertainties described in Section 9.7.2 of the FSAR. I DRAINAGE 5.6.3 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 63 feet.
CAPACITY 5.6.4 The fuel storage pool is designed and shall be maintained with a combined storage capacity, for both Units 1 and 2, limited to no more than 1830 fuel assemblies.
5.7 COMPONENT CYCLIC Oi TRANSIENT t.IMITS 5.7.1 The components identified in Table 5.7-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.7-1.
8806230281 DR 880609 ADOCK 05000317 DCD CALVERT CLIFFS - UNIT 1 5-3 Amer.dment No. ////////JJ, pp,
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DESIGN FEATURES VOLUME 5.4.2 The total water and steam volume of the reactor coolant system is 10,614 460 cubic feet at a nominal T avg f 532 F.
5.5 HETEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure 5.1-1.
5.6 FUEL STORAGE CRITICALITY - SPENT FUEL 5.6.1 The spent fuel storage racks are designed and shall be maintained with a minimum 10 3/32" x 10 3/32" center-to-center distance between fuel l assemblies placed in the storage racks to ensure a gk gf of 5 0.95 with the storage pool filled with unborated water. The keff of s 0.95 includes the conservative allowances for uncertainties described in Section 9.7.2 of the FSAR. The maximum fuel enrichment to be stored in the fuel pool will be 5.0 l weight percent.
CRITICALITY - NEW FUEL 5.6.2 The new fuel storage racks are designu and shall be maintained with a nominal 18 inch center-to-center dist. c a between new fuel assemblies such that k3ff will not exceed 0.98 when fue having a maximum enrichment of 5.0 weight percent U-235 is in place and various densities of unborated water l are assumed including aqueous foam moderation. The kgff of 5 0.98 includes the conservative allowance for uncertainties described in Section 9.7.2 of the FSAP..
DRAINAGE 5.6.3 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 63 feet.
CAPACITY 5.6.4 The fuel storage pool is designed and shall be maintained with a combined storage capacity, for both Units 1 and 2, limited to no more than 1830 fuel assemblies.
5.7 COMPONENT C)CLIC OR TRANSIENT LIMITS i
5.7.1 The components identified in Table 5.7-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.7-1.
l CALVERT CLIFFS - UNIT 1 5-3 Amendment No. 27//47// S ,
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DESIGN FEATURES VOLUME 5.4.2 The total water and steam volume of the reactor coolant system is 10,614 460 cubic feet at a nominal T avg of 532*F. l 5.5 HETEOR0 LOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure 5.1-1.
5.6 FUEL STORAGE l CRITICALITY - SPENT FUEL 5.6.1 The sptnt fuel storage racks are designed and shall be maintained with a minimum 10 3/32" x 10 3/32" center-to-center distance between fuel !
assemblies placed in the storage racks to ensure a keff of f 0.95 with the i storage pool filled with unborated water. The kgff of 1 0.95 includes the I conservative allowances for uncertainties described in Section 9.7.2 of the FSAR. The maximum fuel enrichment to be stored in the fuel pool will be 3rlf 65 0 weight percent.
CRJTICALITY - NEW FUEL i 5.6.2 The new fuel storage racks are designed and shall be maintained with a nominal 18 ' 'h center-to-center distance between new fuel assemblies such that keff wili not exceed 0.98 when fuel having a maximum enrichment of I
('O'Arl,weightpercentU-235isinplaceandvariousdensitiesorunboratedwater are assumed including aqueous foam moderation. The k l
e f 1 0.98 includes the conservative allowance for uncertainties describe'ffd in Section 9.7.2 of the FSAR.
ORAINAGE 5.6.3 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 63 feet.
CAPACITY 5.6.4 The fuel storage pool is designed and shall be maintained with a combined storage capacity, for both Units 1 and 2, limited to no more than 1830 fuel assemblies.
5.7 COMPONENT CYCLIC OR TRANSIENT LIMITS 5.7.1 The components identified in Table 5.7-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.7-1.
CALVERT CLIFFS - UNIT 2 5-5 Amendment No. J////p//fE, E2,
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DESIGN FEATURES VOLUME 5.4.2 The total water and steam volume of the reactor coolant system is 10,614 460 cubic feet at a nominal T avg of 532*F.
5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure 5.1-1.
5.6 FUEL STORAGE l CRITICALITY - SPENT FUEL 5.6.1 The spent fuel storage racks are designed and shall be maintained I with a minimum 10 3/32" x 10 3/32" center-to-center distance between fuel i assemblies placed in the storage racks to ensure a keff of 10.95 with the storage pool filled with unborated water. The kg gf of s 0.95 includes the conservative allowances for uncertainties described in Section 9.7.2 of the FSAR. The maximum fuel enrichment to be stored in the fuel pool will be 5.0 l weight percent.
CRITICALITY - NEW FUEL 5.6.2 The new fuel storage racks are designed and shall be maintained with a nominal 18 inch center-to-center distance between new fuel assemblies such that keff will n t exceed 0.98 when fuel having a maximum enrichment of 5.0 weight percent U-235 is in place and various densities of unborated water l are assumed including aqueous foam moderation. The k e the conservative allowance for uncertainties describea.ff of s 0.98 in Section 9.7.2 of theincludes l FSAR.
l DRAINAGE 5.6.3 The spent fuel storage pool is designed and shall be maintained to ,
prevent inadvertent draining of the pool below elevation 63 feet. l CAPACITY l l 5.6.4 The fuel storage pool is designed and shall be maintained with a l l combined storage capacity, for both Units 1 and 2, limited to no more than 1830 fuel assemblies.
5.7 COMPONENT CYCLIC OR TRANSIENT LIMITS 5.7.1 The components identified in Table 5.7-1 are designed and shall be maintained within the cyclic or transient limits of Tatle 5.7-1.
CALVERT CLIFFS - UNIT 2 5-5 Amendment No. J///JE//ff, 57,
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ATIACHMENT 1 CRITICALITY ANALYSIS FOR STORING 5.0 wt. % U-235 i FUEL ASSEMBLIES IN NEW AND SPENT FUEL RACKS l
I.
SUMMARY
The effective multiplication factors have been calculated for the new and spent fuel storage racks for fuel with a maximum U-235 enrichment of 5.0 weight percent. The new fuel rack effective multiplication factor was calculated as a function of water density. The maximum multiplication factor of 0.89 occurred when full flood conditions exist. No uncertainty analysis was done as there is sufficient margin -to accommodate all possible uncertainties. The spent fuel storage rack effective multiplication factor including all uncertainties is 0.945 when the racks are fully loaded with 5.0 wt% L'-235 fuel.
II. INTRODUCTION This report is provided to support the changing of the Technical Specifications of Calvert Cliffs Nuclear Power Plant Unit No. 1 and Unit No. 2 to store fuel enriched up to 5.0 wt% U-235. It provides the criticality analysis for the new and spent fuel storage racks.
III. DESIGN BASIS III.1 New Fuel Storage Rack The new fuel storage rack design and dimensions are shown in Figures 1 and
- 2. The analysis was done as shown with a concrete wall surrounding the rack and assuming e 10 inch concrete floor. The top of the rack was reflected by the water density assumed within the new fuel storage rack.
The only rack structures modeled in the analysis were the corner angle irons.
III.2 Spent Fuel Storage Rack The cell dimensions were obtained from the BG&E letter to the NRC dated January 15, 1980, on spent fuel storage rack modifications. The inner v al of each storage cell is made of a 0.060 inch thick sheet of 304L stainless steel, formed into a square with an inner dimension of 8-9/16 inches. On the outside of each of the four sides of this inner wall, a poison sheet 6-1/2 inches wide is sandwiched between the inner wall and an external 0.060 inch thick stainless steel sheet. The poison sheet wag assumg to be 0.090 inches thick and was assumed to contain 0.020 gm/cm of B . The external sheet extends over two fuel storage cells. The average center to-center pitch between all storage boxes is 10.09375 inches.
IV. ANALYTICAL METHODS IV.1 General In order to more accurately predict the multiplication factor of the storage arcys , reliable calculation of the spatial flux distribution, especially in the neutron absorbing regions, is essential. For this reason, a two dimensional transport calculational model of the spent fuel storage rack is employed in which each component of the fuel storage array is explicitly represented. Thus, in the normal spent fuel storage cell 1
calculation, the fuel assembly, the water channel between poison box and wall are represented as separate regions. For the new fuel storage rack, the fuel assembly, the corner angle irons, the variable density water between the assemblies and the concrete surrounding the fuel rack are represented as separate regions. Due to'the importance of neutron leakage the calculations were done using a three dimensional Monte Carlo code.
The fuel assembly itself is represented as a 14x14 array of fuel pin cells containing moderator and either fuel pins, guide tubes or instrument tubes. Four neutron energy group cross-sections are generated f- aach fuel _ assembly cell and for each component of the storage cell with . -rial attention given to the effect of adjoining regions on the spatial thermal spectrum and hence broad group thermal cross-sections of each separate region of the storage cell.
IV.2 Cross-Section Generation The CEPAK lattice program (Version 2.3 Mod 4) is used to calculate four neutron energy group cross-sections. This program is the synthesis of a number of computer codes, many of which were developed at other laboratories, e.g., FORM (Reference 1), THERMOS (Reference 2), and CINDER (Reference 3). These programs are interlinked in a consistent way with an extensive library of differcntial neutron cross section data.
The data base for both fast and thermal neutron cross-sections for this version of the CEPAK program is derived from several sources, mainly ENDF/B-IV. This data base gives good agreement with measured data from critical experiments and operating reactors.
IV.3 Two-Dimensional Transport Calculations The two-dimensional, discrete ordinates transport code DOT-IV (Version 1.0 Mod 2) was used to determine the spatial flux solution and multiplication factor. An S-6 order of angular. quadrature is used with a 1.0005 convergence factor (the ratio of successive eigenvalues for each outer iteration). In the storage cell calculations, an assembly is represented with one mesh interval for each fuel pin cell; the surrounding water channel, steel, and water gap regions are calculated with 2 or more mesh intervals.
IV.4 Three-Dimensional Monte Carlo Calculations The three-dimensional, Monte Carlo Code KENO IV was used to determine the multiplication factor for the new fuel storage rack.
IV.5 Qualification of Analytical Methods Qualification of the calculational method and evaluation of calculational :
uncertainties and/or bias factors are based on the analysis of a variety l of reactor and laboratory experiments. A copy cf'the "Qualification of Analytical Methods Used in Spent Fuel Storage Rack Analyses" is included in Appendix A.
Included are UO2 critical experiments typical of reactor cores, BNL rod exponential experiments typical of isolated assemblies, and PNL critical 2
c-separation experiments with spacings and steel inserts typical of fuel storage racks. For the total. of 41 experiments the reactivity is overpredicted by 0.138 percent and has a 95/95 confidence level uncertainty of 0.714 percent.
V. RESULTS V.1 Spent Fuel Storage Rack Table V.1 summarizes the values of effective multiplication f9ctor obtained for the various pool temperatures and material and manufacturing tolerances.
The most adverse effects of temperature, eccentric placement of fuel assemblies and poison boxes, changes in steel thickness, change in center to center spac'ng, boron poison loss possible over the life of the material, the possible loss of Boraflex, and the calculational uncertainty 1 are included in the following manner. The squares of each delta K change are added and the square root of the sum is taken. This result is added to the bias and the nominal K,gg values. The e ffec tive multiplication factor including all uncertainties is 0.945. This is less than the design criteria of 0.95.
The reactivity addition from the possible loss of Boraflex neutron poison material due to shrinking was conservatively calculated. It was assumed that a 4 inch vertical loss of Boraflex had occurred at the active fuel centerline in every poison box in the spent fuel storage rack.
V.2 New Fuel Storage Rack I
The effective multiplication factor as a function of water density surrounding the fuel assemblies and new fuel storage rack is shown in 1 Figure 3. The maximum ef# etive multiplication factor occurs at a water density of 1.0 gram per . ic centimeter (full flood) . The K gg is.0.89. I Due to the large margin avoilable of 9% an uncertainty analy, sis was not done since typical uncertainty analysis results in uncertainties of less than 3.0%.
VI. CONCLUSIONS As previously presented in the results section the effective multiplication factors for the new fuel and spent fuel storage racks are less than 0.95 for 5.0 wt% U-235 fuel.
VII. REFERENCES
- 1. FORM - "A Fourier Transform Fast Spectrum Code for the IMB-7090,"
McGoff, D.J. , NAA-SRO Memo 5766 (September 1960) .
- 2. THERMOS - "A Thermalization Transport Theory Code for Reactor Lattice Calculations," Honeck, H., BNL-5816 (July 1961) .
- 3. CINDER -
"A One Point Depletion and Fission Product Program,"
England, T.R. , WPD-TM-334 (Revised June 1964) .
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TABLE V.1 Spent Fuel Rack 5 wt % U-235 K,77 aK Nominal 68'F 0.936783 Temp Effect 40*F 0.937346 160*F 0.929708 260*F 0.000563 0.915833 Minimum Poison 0.938266 0.001483 Minimum Center-to-Center 0.940335 0.003552 Minimum Wall Thickness 0.937568 0.000785 Most Reactive Position 0.936783 0 Calculation Uncertainty .
.00714 Boroflex Loss :
0.00436 Bias
+ .00138 1
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APPENDIX A 4
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A-1 QUALIFICATION OF ANALYTICAL METHODS USED IN SPENT FUEL STORAGE RACK ANALYSES I. Purpose The purpose of this memo is to provide qualification of the calculational model and evaluation of calculational uncertainties and/or biasyactors used in analyzing spent fuel storage racks, especially the HI-CAP racks employing steel boxes and super HI-CAPS containing boron carbide poison.
This is based on the analysis of a variety of reactor ano laboratory experiments. The methods of cross .aection generation are essentially those of C-E's physics design procedures modified appropriately for use in four group transport, discrete ordirate method criticality calculations, and Monte Carlo codes.
II. Calculational Uncertainty and Bias The results of the analysis of a series of UO2 critical experiments are summarized in Table I. These are calcrilated using the' methods described by Gavin (Reference 1) for CEPAK 2. 3, which is used in present storage rack calculations. Table I includes the mean and standard deviation for this CEPAK model.
Although the spatial solution for the flux distribution was obtained by use of a diffusion theory code such as PDQ-7, transport corrections for the reflector and heterogene0us lattice effects were employed. Thus, for example, in Reference 8 the 4 3 w/o infinite . lattice of close packed assemblies in room temperature nter had a K,gg of 1.4547 in PDQ and 1.4568 in DOT, the conservative bus in DOT of 0.0021 will be ignored, These calculations support use of the differential cross section data base and broad group cross section generation codes.
Since fuel storage arrays do involve the spacing of the fuel assemblies at larger separation distances than in typical PWR reactor lattices, the predictive capability of the calculational model was tested on the following experiments. In these analyses done for this memo, the spatial flux solution was obtained directly with the transport code, ANISN. To assess the accuracy of the calculational model in predicting the multiplication factor of fuel assemblies having a separation distance sufficiently large so as to be isolated, analyses were carried out for a group of subcritical exponential experiments on clusters of 3.0 w/o UO2 fuel pins clad with type 304 S. S. and moderated by H2 O (page 165 of Reference 9). The cluster sizes analyzed vary from 181 to 301 fuel rods
A-2 so as to encompass the range of sizes typical of current PWR fuel assemblies. The multiplication factors for.the lattices analyzed using l axial bucklings deduced from the reported relaxation lengths are tabulated -I below. l E
No. of Fuel Rods eff ~
181 0.9966 211 1.0011 i 235 0.9966 265 0.9988 301 0.9984 These results indicate that the calculational model predicts the multiplication factor for small clusters of fuel . rods in a water environment to a high degree of accuracy, i.e.' a bias of .0017. <
l To ascertain whether the calculational mode can predict the reactivity characteristics of thick stainless steel plates and boron poisoned plates .
an analysis (Reference 10) was made of PNW experimental (Reference 11) critical separations of 2.35 w/o U-235 UO2 suberitical clusters. The l results using the Monte Carlo code KENO IV are shown in Table II. !
1 Method of Calculation i
The calculation methods for these experimental comparisons which are also used to determine reactivity for fuel rack storage, fuel shipping containers plus other fuel configurations found in fuel manufacturing areas are based on CEPAK 2.3 (Reference 1) cross sections. Using an appropriate buckling value and taking proper account of resonance :
absorption, three fast groups are collapsed from 55 fine energy mesh )
groups in FORM and the one thermal group is collapsed from 29 thermal -
energy groups in THERMOS. In addition, each component such as water . gap, i or poison plate has its thermal cross section determined by a slab THERMOS calculation employing the proper fuel environment. FORM and THERMOS are sub-programs of CEPAK.
For one dimensional analyses such as the BNL exponential experiments the discrete ordinates code ANISN (Reference 12) is used. For two dimensional analyses DOT-2W (Reference 13) is used. For three dimensional analyses l (such as the critical separation experiments) KENO IV (Reference 14) is l used. l l
Results I The above analyses indicate a mean error between predicted and measured multiplication factors of +.00135 and a calculational uncertainty of
.00714 at the 95/95 confidence level for the complete series 'of UO 2 experiments.
b, A-3 Thus, using CEPAK 2.3 crcss sections we conclude the following -
Total Number of Results 41 l Mean Value ( ) 1.00138 Standard Deviation - 6 0.00337 Multiplier for 95/95 confidence 2.118 95/95 Confidence Level Uncertainty 0.00714 Bias ( - 1.0) +.00138
- Uncertainty Minus Bias .00575 It will be noted that the seven no boron steel cases have a bias of 0.00207 (i.e. the calculated value is .00207 greater than the critical K,gg value of unity) which is greater than the mean bias. The three boral cases have a bias of -0.00435 with unity particle self-shielding factor for the B 4C. Because of the size and distribution of the boron carbide particles the boron allows more transmission than an equivalent homogeneous boron carbide mixture. Neutron transmission experiments conducted by the University of Michigan for Brooks & Perkins, Inc.
(Reference 15) are consistant with using.a 0.9 self-shielding factor in the third of four CEPAK neutron group and a 0.75 self-shielding factor in the thermal group. These seif-shielding factors which are used in designing boron containing fuel racks make the bias for these boral cases
+0.00008.
References:
- 1. P. H. Gavin, "CEPAK 2.3 Mod 0," PHP-76 488, December 14, 1976.
- 2. T. C. Engelder, et al, "Spectral Shift Control Reactor, Basic Physics Program," B&W-1273, November 1963.
- 3. R. H. Clark, et al, "Physics Verification Program Final Report,"
B&W-3647 3, March 1967.
- 4. P. W. Davison, et al, "Yankee Critical Experiments," YAEC-94, April, 1959.
- 5. W. J. Eich and W. P. Rocacik, "Reactivity and Neutron Flux Studies in Multi-Region Loaded Cores," WCAP-1443, 1961.
- 6. F. J . Fayers , et al, "An Evaluation of Some Uncertainties in the Comparison Between Theory and Experimants for Regular Light Water Lattices, Brit. Nuc. En. Soc. J. 6, April 1967.
- 7. J. R. Brown, et al, "Kinetic and Buckling Measurements on Lattices of Slightly Enriched Uranium and UO2 Rods in Light Water," WAPD-176, 1958.
- 8. J. Handschuh, L. C. Noderer, R. C. for "Compact Spent Fuel Storage Criticality Analysis for Arkansas Power and Light, Unit 2 at 68 F, "6370 PH-RC010, April 8, 1975.
- 9. G. A. Price, "Uranium - Water Lattice Compilation Part 1. BNL Exponential nssemblies," BNL-50035 (T-449), December 1966.
I
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- 10. L. C. Noderer, "Analysis of Critical. Separation of Low Enriched Suberitical Clusters," PHD-79-38, May 11, 1979.
- 11. S. R. Bierman, E. D. Clayton and B..M. Durst, "Critical Separation Between Suberitical Clusters of 2.35 w/o U-235 Enriched UO in Water with Fixed Neutron Poisons," PNL-2438, October 1977.2 ' Rods '
- 12. Ward W. Engle, Jr. "A Users Manual for ANISN, A One Dimensional Discrete . Ordinates Transport Code With Anisotropic Scattering K-1693, March 30, 1967.
- 13. R. G. Sottesy, R. K. Disney, A Collier, "User's Manual for the DOT-IIW Discrete Ordinates Transport Computer Code," WANL TME-1982, December 1969,
- 14. L. M. Petrie and N. F. Cross, "KENO IV, An Improved Monte Carlo Criticality Program," ORNL-4938, November 1975.
j 15. James W. Bryson, John C. Lee and R. Robert Burn,. "Neutron Transmission Through Boral Shielding Material: Theoretical Model and Experimental Cc- Trison," University of = Michigan, Dept. of Nuclear Engineering, MAchigan Memorial-Phoenix Project, prepared for Brooks & Perkins, Inc. April 1978.
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A5 Table I Results of Analysis of Critical UO2 Systems l
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l No. Lattice B{ot Eeff*
1 B&W (2) 1 .88-2 1.00121 i 2 II .172-2 1.00534 )
3 X .79-2 .99838 1 4 XIII .701-2 1.00419 5 XX .202-2 1.00550 6 B&W (3) 1 .861-2 1.00269 7 2 .420-2 1.00443 ;
l 8 Yankee (4) 1 .408-2 1.00088 9 2 .531-2 1.00115 10 3 .633-2 1.00136 11 Yankee (5) 4 .688-2 1.00244 Winfrith (6) 12 R1-20 .660-2 1.00214 13 Rl-80 .626-2 .99942 14 R3 .510-2 1.00422 15 Bettis (7) 1 .326-2 1.00053 16 2 .355 2 1.00046 17 3 .342-2 1.00106 Average 1.00208
.00206 l
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- Using calculated radial bucklings and measured axial bucklings.
l l
A-6 TABLE II Calculated K eff Values For Separation Experiments Monte Carlo Eupt # Type Poison Plate K ogg (STD Deviation) 15 None 1.00227 .00534 04 None 0.99912 .00540 49 None 1.00221 .00473 18 None 1.00813 .00489 21 None 0.99589 .00461 28 304 S Steel 0.0 w/o Boron 1.00393 .00308 05 304 S Steel 0.0 w/o Boron 1.00329 .00303 29 304 S Steel 0.0 w/o Boron 1.00271 .00302 27 304 S Steel 0.0 w/o Boron 1.00418 .00273 26 304 S Steel 0.0 w/o Boron 0.99811 .00279
\ 34 304 S Steel 0.0 w/o Boron 0.99793 .00297 35 304 S Steel 0.0 w/o Boron 1.00436 .00290 32 304 S Steel 1.05 w/o Boron 0.99970 .00524 33 304 S Steel 1.05 w/o Boron 1.01173 .00491 38 304 S Steel 1.62 w/o Boron 1.00289 .00512 39 304 S Steel 1.62 w/o Boron 1.00208 .00506 20 Boral 0.99585 .00301 16 Boral 1.00020 .00288 17 Boral 0.99519 .00286 Mean K Value 1.00157 Std.d$hfation .0041?
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