ML20087P782
| ML20087P782 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 03/21/1984 |
| From: | Gournelos D, Grow R, Smolinske K CONTROL DATA CORP. |
| To: | |
| Shared Package | |
| ML20087P769 | List: |
| References | |
| UAI-84-17, UAI-84-17-R01, UAI-84-17-R1, NUDOCS 8404090364 | |
| Download: ML20087P782 (24) | |
Text
- - - - - - . - - . _ . _ _ _
g-I Attachment 2 ,
' i Summary Report Nuclear Criticality Re-analysis -
for 4.3 Weight Percent U-235 Fuel i
. In the New Fuel Storage Rack of - l the Joseph M. Farley Nuclear Plant i i
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[ , UAI 84-17
SUMMARY
REPORT NUCLEAR CRITICALITY RE-ANALYSIS FOR 4.3 w/o FUEL IN NEW FUEL STORAGE RACK OF JOSEPH M. FARLEY NUCLEAR PLANT -
0F ALABAMA POWER C0t1PANY Prepared by -
K. M. Smolinske Senior Project Engineer Reviewed by 1; =3 R. L.' Grow '
Manager, Special Projects Dept.
D.'T. Gournelos Senior Associate l
Approved by d Q. .' Fisher rechnical Director
~
l 1
Revision J- Itarch 21,1984 UTILITY ASSOCIATES INTERNATIONAL
@BcmT mTc h nm ,
0003 EXECUTIVE SOULEVAPO ROCKVILLE. MARYLANO 20052
Revision Number 1
Date of Revision March 21, 1984 Date Entered Revision Entered by UAIDOCUMENT REVISION NOTICE Title of Document Summary Report, Nuclear UAI Document Number Criticality Re-Analysis for 4.3 w/o Fuel in 84-17 New Fuel Storage Rack of Joseph M. Farley ' Copy Number ,
Muelmar plant nf Alahama Pnwar Cnmnany ,
Please replace entire document with the attached major revision.
As the superseded material is proprietary to UAI, you are requested to return it in the accompanying envelope.
Acknowledgement of receipt of this revision by return of the attached Document O Transmittal . sheet is requested.
Please. revise your copy of the UAI document indicated above in accordance with
- the changes listed below:
Pace Number Status Comment Title Page Revised Change title, revision number, and date i Revised Add Section 7.0
, 1-1 Revised Add FSAR reference, add second paragraph 1-2 Revised Editorial, add explanation that low densities are not credible 2-1 . Revised Add concrete thickness of KENO benchmark 2-2 Revised Add KENO and CASMO benchmarks .
2-3 Revis e'd Editorial, add criteria 8 3-1,3-2 Revised Editorial 4-1 Revised Entire section replaced
~
5-1, 5-2 Revised Rewrite introduction to section, new explanation of dropped assembly accident l
l
. - - -- .. -- .. . - - - - - _ ~
' Document Revision Nstice UAI 84-17 March 21,1984 Page 2 i
Page Number Status Comment 6-1 Revised Editorial, add detail on criteria 7-1 New Add references Figure 5' Revised Add concrete dimension ,
Please insert this Notice in proper sequence in document.
Prepared by ' Approved by:
K. M. Smolinske. (/J. R. ' Fisher Sr. Project Engineer Technical Director e
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UAI 84-17
.e LIST OF EFFECTIVE PAGES I
PAGE REVISION Cover -1 i 1 s 1-1 1 1-2 1 2-1 1 2-2 1 3-1 1 3-2 1 4-1 1 5-1 1 5-2 1 6-1 1 7-1 l' Table 1- 0 Figure 1 0 Figure 2 0 Figu're 3 0
- Figure 4 0 Figure 5- 1
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UAI 84-17 o
Revision .1 March 21,1984 ;
e l
TABLE OF CONTENTS
-Section_ Title Page
1.0 INTRODUCTION
1-1 l 2.0 MODEL DESCRIPTION 2-1 2.1 The Monte Carlo Transport Model 2-1 2.2 The Transport Theory Model 2-2 2.3 Calculation Assumptions 2-3 3.0 CALCULATIONAL RESULTS 3-1 3.1 Dry Case 3-1 3.2 Water Density Variation 3-1 3.3 Reference Case 3-2 4.0 UNCERTAINTIES AND TOLERANCE CONSIDERATIONS 4-1 5.0 ACCIDENT CONSIDERATIONS 5-1 5.1 Flooding 5-1 5.2 Dropped Assembly 5-1 6.0
SUMMARY
AND CONCLUSIONS 6-1
7.0 REFERENCES
7-1 g
0 S
i
UAf 84-17
~
Revision 1
- ~ March 21,1984 ;
1.0 INTRODUCTION
The design and licensing basis of the Farley new fuel storage racks are l described in FSAR Section 9.1.1. A nuclear criticality safety re-analysis was performed by Utility Associates International on the Joseph M. Farley new fuel storage racks. These racks had previously been approved for storageofnewfuelupto3.5w7oU-235. This report documents the re-analysis of these racks for-new fuel up to 4.3 w7o U-235. l The design basis for preventing criticality outside of the reactor is that, including uncertainties, there is a 95 percent probability at a 95 percent -'
confidence level that the effective multiplication factor (K,ff) of de j fuel assembly array will be less than 0.95 when fully flooded with unborated ;
water and less than 0.98 with fuel of the highest anticipated enrichment in :
place assuming optimum moderation, as recommended in ANSI N18.2-1973.I l Criticality of fuel assemblies in the new fuel storage rack is prevented by the design of the rack which limits fuel assembly interaction. This is accomplished by restricting the minimum separation between assemblies to take advantage of neutron absorption in water and stainless steel.
-The new fuel racks for the Farley plant consist of double rows of fuel 4
assemblies with a nominal' pitch of 21 inches in four separate storage pits. (
In one of the four pits, the fuel assemblies- are surrounded by a 0.075" thick stainless st. eel canister with an internal cavity clearance of 9.0".
The other three pits use 2.0" wide and 0.25" thick stainless steel angles in tile four corners around the assemblies. The internal clearance between these angles is also 9.0". These racks are illustrated in Figure 1.
Two of the four pits have nominal- distances from the center of the nearest assembly to the concrete wall of 18" on one side of the double row and 15" L - on the other.- The remaining two pits have nominal distances of 25.5" from j the center of the nearest assembly to the concrete wall on both sides of the double. row. These pits are illustrated in Figure 2.
1-1
UAI 84-17
- ' Revision 1 March 21,1984
-c
.The criticality analysis was based on fresh fuel with a U-235 enrichment of 4.3 w'/o. Although new fuel is nonna11y stored in the dry condition, the criticality requirements include the interjection of possible moderators.
Therefore, any water present was assumed to be fresh and non-borated.
- Additionally, no credit was taken for the burnable poison rods which may be present in the fuel assemblies. The parameter analysis used the transport 2
model, CASMO-2E ,3 to investigate various criticality safety-related aspects
~!
-of the rack design, including the flooded condition and low moderator (water) density. The full range of moderator densities, was considered for expediency and conservatism, even though the attainment of very low densities is not considered credible. These studies were performed using an infinite ' ,
4 '
lattice of fuel assemblies on a 21" pitch. The Monte Carlo Model, KENO-IV ,
was used to assure the reactivity of the rack design in the optimum moderator density condition is less than required by the criteria stated above. This report presents a description of the criticality analysis and the results for nonnal and adverse conditions.
Section 2.0 of this report describes the calculational models and the basic assumptions used in the analysis. Section 3.0 presents the calculational l results of the analysis. In Section 3.0, sub-section 3.1 presents the '
results of the dry case; sub-section 3.2 presents the results of the water density variation study; and sub-section 3.3 presents the results of i the reference case calculations. In this report, the reference cases are
- l not the nominal case but the near-optimum moderator density cases, which bound ,
the nominal case. -
Section 4.0 describes uncertainties and tolerance considerations. Section 5.0 addresses the criticality effects of two accident conditions: The case of an assembly dropped alongside of the assemblies in the rack; and the fully flooded' case. -Note that the optimum moderator condition is l
addressed in Section 3.3 by the reference cases.
Section 6.0 gives a summary of the results indicating that all applicable limits are met for 4.3 w7o U-235 enrichment.
1-2
~
a- UAI 84-17 Revision 1 March 21,1984 2.0 MODEL DESCRIPTION 2.1 The Monte Carlo Transport Model The finite rack reactivity calculation employs the KENO-IV model. The basic neutron cross-section data come from the master -
library of AMPX. AMPX is a 123-group GAM-THERMOS neutron library 5
prepared from ENDF/B version II data. The NITAWL module of the AMPX program is used to perform the Nordheim integral treatment
- of the U-238 resonances. The working library produced by the NITAWL/AMPX module retains the 123-group structure and is used directly by KENO-IV.
In the KENO-IV calculation, each fuel and water rod cell is represented discretely. The array option of KENO-IV is applied to arrange the fuel and water rod box types into a matrix representing the fuel assembly. Then a water region is added to the outside of this matrix, followed by stainless steel angles in each corner, an outer water gap region, and a concrete region on one side of the storage rack cell. To simulate the arrange-ment of two storage rack units wide and a large number of storage e rack units long, and for a non-leakage condition in the axial directions., reflection is applied to the five non-concrete sides of this storage rack cell. For the sixth side, the concrete side, a zero flux boundary is applied 12 inches into the concrete. (See Figure 5.)
The KENO-IV/AMPX code system has been benchmarked against several critical experiments.6 The two experiments cited below are used to benchmark ~ KENO-IV/AMPX-in this report because they contain no poison.
The first experiment is a critical configuration measured at ORNL which had 203 uranium metal rods of 4.95+ .05 w7o enrichment
-in water. The calculated result using KENO-IV/AMPX was K,ff= .9981 005.
2-1
- UAI 84-17
-- - Revision 1
, March 21, 1984 i
-This K,ff was -0.002 ak below the critical value of the experiment.
The second critical experiment was a measurement made on the Lacrosse reactor. This experiment was chosen, in part, because it approaches the fuel storage rack configuration in that it was an unborated, virtually unrodded configuration. The result of this benchmark was:
Keff Measured: 1.009 KENO-IV/AMPX adjusted to include grids: 1.008 Therefore, the KENO-IV/AMPX result was -0.001 ok below the measured value, which is consistent ~with the ORNL critical results.
i.
2.2 The Transport Theory Model The criticality analysis for the Farley new fuel racks employs the CASMO-2E'model to characterize the curve of k, vs. water density l for the infinite rack and to analyze the completely flooded and dry' conditions. CASMO-2E is a multi-group, two-dimensional, transport theory. code and uses a lib'rary containing data in 25
.' energy groups. The CASMO-2E model is an infinite lattice of assemblies separated by large water gaps. The storage rack. cell
~
in the CASMO-2E model has no leakage in any direction.
r The CASMO . transport code has been benchmarked against KENO-IV/AMPX for several configurations. ~These configurations-include cases with and without boron. The result from 16 configurations is:
-CASM0-PDQ l.002 + 0.003 ,
l KENO-IV/AMPX 0.997-+ 0.010 1
I
- 2-2 .
l
VAI 84-17 e
Rsvision 1 March 21, 1984 Therefore, the CASMO-PDQ results over-predicts the KENO-IV/AMPX results by -0.005 ok over-all.
2.3 Calculation Assumptions -
To ensure that the analysis follows a conservative approach and conforms to the general guidelines of criticality safe'ty analysis,
~.
the calculations were perfonned with the follow assumptions:
- 1. The fuel is fresh (most reactive point in life) at 4.3 w7o U-235.
- 2. The effect of U-234, U-236, and th,e spacer grids is not l included.
- 3. Other minor structural members in the assembly are replaced l
by water or void.
- 4. No soluble poison in water no. fixed poison in the fuel assembly is included.
- 5. Stainless steel angles in the corners were selected as the more comon configuration than the stainless steel canister. In the event that the K g7 was close to the applicable limit, the case would have been re-analyzed using the canister. However, as shown in the sumary in Section 6.0, this re-analysis was not necessary.
- 6. .The distance to the concrete wall does not have a large effect onathe reactivity of the cell, so the largest distance was used since for most densities water is more reflective than concrete.
- 7. The properties of Portland' concrete were used
- 8. No axial leakage was considered.
2-3
~
~
, UAI 84-17 .
Revision 1 !
- March 21,1984 3.0' CALCULATIONAL RESULTS The storage rack is comprised of two side-by-side rows of assemblies. The rack pitch is 21 inches. The size of this spacing is important in determining the optimum moderator density l for the criticality analysis.
The analys'is took no credit for the possible presence of any burnable poison rods in the assembly. Figure 3 shows the geometry layout of the l ,
storage cell and the dimensions and materials for various components.
The input parameters used in the analysis are listed in Table 1.
3.1 Dry Case l Using.the~ input data from Table 1 and the nominal dimensions from Figure 3, an upper limit of k= of the dry case at 68 F was-determined by using 0.1% density with no boron present.
(See Figure 4.) The CASMO-2E transport model yielded a k-=0.8883 in the infinite lattice configuration.
Since this case also does not include leakage, the true X,ff of the dry case would be considerably-lower than the above result.
' The reference cases,in Sub-section 3.3, illustrates the magnitude l of this difference.
3.2~ Water Density Variation
~
CASMO-2Ewasusedtodeterminek;vs.waterdensity. The l same input data from Table 1 and Figure 3 was used for these CASMO-2E cases as in the dry case, except that the water density
' was -varied.
3-1
UAI 84-17
-' Revision 0 e
March 9, 1984 Using the above described input data, the km values at 68 F were also calculated. The shape of k= vs. water density from the transport theory model is shown in Figure 4.
Since all the CASMO-2E cases, are infinite lattice, the K,ff for the actual rack geometry is much lower, and Figure 4 was used only to characterize the curve of k, vs. water density. ,
3.3 Reference Case
.The reference cases in this report are the KENO-IV calculations in the nominal rack geometry at 2% and 57, water density. The nominal rack ' geometry is the same as shown in Figure 3 except that the amount of water on one side is increased and bounded by concrete as shown in Figure 5. This water distance to the concrete corresponds to a 25.5" distance from the center of the assembly to the concrete wall.
These two cases were run to demonstrate that the finite rack configuration is substantially sub-critical for all water densities.
It also is noted that no credit has been taken for axial leakage, which for these low water densities is estimated to be about half
.of that in the XY direction.
Using ths inp0t described above, the K,ff values of the reference case at 680 F were calculated. The results from the Monte Carlo model is given below:
KENO-IV(nominalgeometry)
K,ff 95% Confidence Interval Keff, 2% density 0.7094 1 0106 0.6882 to 0.7306 K,ff, . 5% density 0.7480 i .0113 0.7254 to 0.7706 3-2
UAI 84-17
~ '
Revision 1 March 21, 1984 S
14.0 . UNCERTAINTIES AND TOLERANCE CONSIDERATIONS Consideration of uncertainties and tolerances include three items:
the 95% confidence interval for KENO-IV; the bias between KENO-IV and measure-ment; and the bias due to positional and dimensional tolerances. Since the reactivity of either reference case would need to increase by more .
.than 0.20 ak to approach a K,ff of 0.95, the total of the three listed considerations need to be close to 0.'20 ak before a detailed analysis
, is required. The magnitude of these considerations is discussed below.
In sub-section 3.1, the. largest 95% confidence interval is .0226 ak above the nominal. The KENO-IV bias is taken to be 0.001 ok to 0.002 ak
.as described in Sub-section 2.1.
To estimate the size of the positional and dimensional tolerance adjustment, one CASMO-2E was run at 5% moderator density with a pitch of 20.386 inches as a limiting case. This case yields an .0234 ok increase in reactivity which is on the same order of magnitude with UAI's past experience
(.010 to .018 ak) for positional and dimensional tolerance bias.
The sum' of the worst of these biases and uncertainties is:
.0226ak + .002ak + .0234ak = .048ak which is much less 'than 0.20 ak.
Adding-the total to the 0.05 g/cm3 water density case yields:
K,ff = 0.748 + 0.048 = 0.796 l
This K,ff is much less than the 0.98 limit allowed for optimum moderation, and since this. case is representative of the maximum reactivity expected, no case will exceed its limit of 0.95 or 0.98 as applicable.
4-1
i UAI 84-17 Revision 1
, March 21, 1984
- r ,
l 5.0 ACCIDENT CONSIDERATIONS Two accidents considered were flooding of the new fuel pit and'a dropped assembly between the. periphery of the new fuel rack and the fuel pit wall. Each of these accidents is presented below.
5.1 Flooding CASM0-2E was used to calculate the reactivity of the fully flooded case. The same input data from Table 1 and Figure 3 was used for this case as for the dry case, except that the rack was fully flooded.
Using the above described data, the K= value of the fully flooded case at 68'F was calculated. The result of the CASMO-2E transport model was K= = 0.8160 in the infinite lattice configuration. This result is plotted in Figure 4
'with an extrapolated curve to the 25% water density result.
Since this case did not include leakage, the K f the fully df flooded rack is considerably lower.
- 5.2 , Dropped Assembly The second a'ccident is a dropped assembly alongside of the assemblies in the rack. In this accident, a single assembly is dropped during fuel handling and lands on end in the pit between one of the rows of fuel assemblies and the fuel pit wall. The disign of the new
' fuel racks preclude the possibility of dropping an asserrbly between
' the two rows of assemblies; therefore, this case was not considered.
It was assumed that the accident occurs during normal conditions, that is, for a dry condition. The possibility of a low moderator density condition'is a result of an accident or malfunction and to postulate 5-1
VAI 84-17
.. . - Revision 1-e March 21, 1984 ;
a second accident, i.e., the dropped assembly, occurring simul-
~ taneously is not considered credible.
A conservative approach was taken to determine the upper limit of K,ff for this accident condition. An infinite array of assemblies-infinitely long was assumed. Since this eliminates neutron leakage, and there is no material between assemblies .
(i.e., the racks are dry) the K,ff of the configuration is independent of pitch and hence also represents an infinite array of close packed' assemblies. Therefore, the K,ff of this condition represents an upper limit of the actual rack configuration with an assembly dropped alongside, since the actual accident configuration
'is only three assemblies wide. Sub-section 3.1 showed tnat tne K,ff of an infinite array of assemblies infinitely long is less unan 0.8683, and hence the K,ff for the droppea assembly condition is much less than 0.8883.-
The K,ff of the accident condition of a fuel assembly on top of a fully loaded rack is also much less than the infinite configura-ti on . Since infinitely long fuel assemblies were assumed in the analyzed configuration and one assembly lying on top of the loaded storage rack is a much less reactive condition, K,ff <<
0.8883 for this configuration as well.
f e g 6
O
=
0 5-2
UAI 84-17 Revision 1
, March 21, 1984 6.0
SUMMARY
AND CONCLUSIONS The conclusion of this re-analysis of the Farley new fuel storage rack is.' that the current rack design is adequate for storage of
~4.3 w7o fuel under normal and abnormal conditions. The results of the analysis are:
<<0.8883 K,ff dry K,ff, flooded <<0.8160 3
K,ff, 0.02 g/cm water density 0.7094 1 0.106 (95% Confidence Interval 0.6802 to 0.7306) 3 Kgf, 0.05 g/cm water density ' O.7480 t .0113 l (95% Confidence Interval 0.7254 to 0.7706)
Kg f, dropped assembly <0.8883 All. values are from infinite lattice calculations using CASMO-2E, except the near-optimum K s which are nominal rack geometry calculations df using AMPX/KENU-IV. All values are less than the applicable acceptance criteria of 1 0.950 for dry or flooded conditions, and 1 0.980 for postu' lated accidents or optimum moderation.
4
=
6 z- - __ _ . . . _ _, . _ _ _ . . . . . - _ . . ~ . . . _ - - - -._. .
~ UAI 84-17 A Revision 1 March 21,1984
7.0 REFERENCES
-1. ANSI, N18.2 - 1973 " Nuclear Safety Criteria For The Design of Sationary Pressurized Water Reactor Plants."
- 2. M. Edenius, "CASMO A FueIl Assembly Burnup Program - User's Manual", Studsvik Energiteknik AB. ,
~3. " Appendix to CASM0-2' User's Manual" (Containing j CASMO-2E Option), Studsvik Energiteknik AB, 1982 and 1983.
. 4. L. M. Petrie and N. F. Cross, " KENO-IV - An Improved Monte Carlo Criticality Program", ORNL-4938, November 1975.
- 5. "NITAWL/XSDPNPM - User Inforamtion Manual", CYBERNET Services, Control Data Corporation.
- 6. ANSI N18.2-1973. " Validation of Calculational Methods for Nuclear Criticality Safety".
- 7. NAI 76-71, " Spent Fuel Rack Criticality Calculations for Carolina Power and Light Company", Nuclear Associates International.
- 8. M. Edenius, K. Ekberg (Studsvik), E. Pilat, D. VerPlanck (Yankee Atomic) ."CASMO Benchmarking for Fuel Rack Geometries",
Studsvik Energiteknik AB.
=
7-1
' ~
UAI 84-17 Revision 0 March 9,1984 TABLE 1 FUEL DATA 1.0 Fuel Assembly Type 17x17 Westinghouse 2.0 . Pellet 0.D. 0.3225" 3.0 Clad Data 3.1 0.D. 0.374" 3.2 Thickness 0.0225" 3.3 Material Zr-4 4.0 Fuel Rod Pitch 0.496" 5.0 U-235 Enrichment 4.3 w/o .
6.0 UO Density 95% theoretical 2
3 7.0 Stack Density 10.287 g/cm 8.0 -Water Hole Data 8.1 Thimble Material Zr-4 8.2 Thimble Dimensions Instrument Guids Tube Tube Upper Lower 0.D. 0.482" 0.482" 0.429" I.D. 0.450" 0.450" 0.397" Number 1 24 9.0 Active Fuel Length 143.7" 10.0 Plenum Length
- 6.3" 11.0 Number of Intermediate Grid Spacers 8 12.0 Dry Weight of Fuel Assen61y 1467 lbs.
I l
l
- UAI 84-17
. Rsvision 0 March 9, 1984
, FIGURE 1
. FARLEY NEW FUEL RACK DESIGNS
- 1) Two configurations of new fuel storage rack using angles:
,2.0" _2.0"'
~
O.25" 0.25" .
o y a ,,-. s '
a 2.0" l 2.0"
, N '
9.0" _ _
9.0" _
a r East Pit West Pit Units 1 and 2 Unit 1
- 2) Configuration of new fuel storage rack canisters:
0.075" o
9 A 9 I
9.0" _
& D West Pit Unit 2 NOTE: Diagram not to scale. All dimensions are in' inches.
UAI 84-17
<.- Revision 0 March 9, 1984
, FIGURE 2 FARLEY NEW FUEL STORAGE RACK PIT DESIGNS
~d a 21.0" ,
b '
= ,
o 21.0" 9
a Rack Storage Cell
/
a.
p . $ .4
'l .'
West Pits, Units 1 and 2: .
~-
a = 18.0", b= 15.0" East Pits, Units 1 and 2:
a = b = 25.5" NOTE: Diagram not to scale. All dimensions are in inches.
- - , , .- - ,, ,. - ---,m+r- w e 't+ --
-et- t "=-ee' -- -v-----eve- -< 'w w vi -
UAI 84-17
- Revision 0 March 9, 1984 FIGURE 3 FARLEY PWR NEW FUEL RACK CELL GE0 MERRY 21.0" z' a .
, j Fuel Rod Cell 2 9.0" '
Control Rod Cell X X X Xf
/'- _ , Instrument Cell X
X X X X -
X /
, s X X m
~
X X X X X X X X X _
X X X SS Angle 0.25"
[
L 8.432" -- !
l.- 2. 0 "--.l l i
l NOTES:
- 1) All dim msions given in inches. Diagram not to scale.
- 2) CASMO-2E cases run without SS angles.
- 3) KENO-IV case run with 25.5" from center of fuel assembly to concrete wall.
- * - VAI 84-17 A R; vision 0
' +
March 9, 1984 7
FIGURE 4 JOSEPH M. FARLEY l X- VS. WATER DENSITY FOR AN INFINITE LATTICE OF 4.3 wfo NEW FUEL l
(21" Pitch) 1.6 . . . . ,,,, ,,, n., , . ...
.... . uc.,. . .
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0.100 1.000 0.001 0.010 3
Water Density (g/cm ) .
-- . , . . - . _ - , - . , , , ~ . - , , _ - - - -
, . - . . VAI 84-17
' ' R2vicion 0 1 -
March 9, 1984 FIGURE 5 ?
FARLEY NEW FUEL STORAGE RACK CONFIGURATION FOR KENO Concrete 9.0" SS Angle .
~
'q
.,o
._- 8.432" -
o .-
'I e, e 17x17 W Ass'y i~
i i
i cIEE5 o
a.. . . ' i . . -
.. ~
2.0" 0.25" i
/ r ,.
-12" . 25.5" ,.
.l NOTE: Diagram not to scale. All dimsnsions are in inches.
i
/
w+