ML20238F235
| ML20238F235 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 08/31/1987 |
| From: | Dzenis E, Frank F WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20238E926 | List: |
| References | |
| NUDOCS 8709160038 | |
| Download: ML20238F235 (21) | |
Text
.
f,,
y W
L.;,
4
-; i A
/
f 3
r I,.,
- RELOAD SAFETY EVALUATION
-l MILLSTONE Lc,. ;
UNIT'3 CYCLE 2-
- 1. ;
' August 1987 Edited by:
F. J.' Frank
'a -
Approved:
M. I.
A __
-E. A. Dzenis, Manager Core Operations Commercial Nuclear Fuel Division nurs-sta-8709160038 870909 PDR ADOCK 05000423 P
=_ :
[.'17 1
[
.; e-r 4
A 4
'-)
' i
' TABLE OF CONTENTS' i R11/E N',
Titie' Page_:
1.00
' INTRODUCTION AND
SUMMARY
-1 L1.1 Introduction 1
3 L
1.2' General Description 1"
1.3; Conclusions 2
,3 2.0
-REACTOR DESIGN 3
y 2.1' Mechanical Design 3
2.2 Nuclear Design 4
2.3 Thermal and Hydraulic Design:
-4 3.01 POWER CAPABILITY AND ACCIDENT EVALUATION-6 13.1 Power Capability 6~
3.2 Accident Evaluation 6
3.2.1
. Kinetic Parameters 7
3.2.2 Contro1~ Rod Worths-8 3.2.3 Core Peaking Factors 8
3.3~
Incidents Reanalyzed
'8
'4.0' TECHNICAL SPECIFICATION CHANGES 10-5.0-REFERENCES 11 li i
3978F 1-870813 t-
^
^
n.
/
n.
LIST OF TABLES' e
Table' Title Page o
i t t l'
Fuel: Assembly-Design Parameters-13 2'
Kinetic Characteristics.
14 -
3-Lend.of Cycle Shutdown Requirements and Margins 15 LIST OF FIGURES
~
Figu'e.
Title Page r
1
_ Core ~ Loading PatternLMillstone 3 Cycle 2'
.16
'2,
- K(z) - Normalized'F (z) as a Function of-Core Height 17' g
ifor'FourJLoop Operation K(z);- Normalized.F (z) as a Function of Core Height 18 3
t g
.for Three Loop' Operation-O_
e ii
.q an,-e-ene n o
')
n
^
1.0 ' INTRODUCTION AND
SUMMARY
1.1 INTRODUCTION
- e This report presents an evaluation for Millstone Unit 3, Cycle 2, which demonstrates'that the core reload will not, adversely affect the. safety of the plant for either N or N-l' loop operation.
This evaluation was accomplished utilizing the' methodology described in WCAP-9273-A, " Westinghouse Reload-Safety Evaluation" Methodology"(1)
, g
?!ncluded in this evaluation is consideration of Resistance Temperature-DetectorL (RTD) Bypass Elimination and operation with revised Technical Specifications.that allow a positive moderator temperature' coefficient (PMTC) while the reactor is critical. The primary documentation for these changes is fo;ind in References 10 and 11, respectively, which are referred to throughout th(s report.
Based upon the above referenced methodology, only those incidents analyzed and' reportedintheF5AR(2)whichcouldpotentiallybeaffectedbythisfuel reload have been reviewed for the Cycle 2 cesign described herein.
The justification-for the applicability of previous results is provided.-
'1'. 2 GENERAL DESCRIPTION The Millstone Unit 3 reactor core is comprised of 193 fuel assemblien arranged-1 in the core' loading pattern configuration'shown in Figure 1.
The Cycle 2 core loading configuration features a low leakage pattern. During Cycle 1/2-
)
refueling, 84 fresh region 4 assemblies will replace 65 region 1 fuel assemblies and 19 Region 2 fuel assemblies. A summary of the Cycle 2 fuel
.~
inventory is given in Table 1.
1 I
3978F 0-870613 l
l l
i, i
NominalIcore' design'parametersutilizedfor' Cycle 2 areas.follows:
Four-Loop (N)
Three Loop'(N-1)
'(LOCA)
(NON~LOCA)'
. Core Power:(MWt) 3411 2217 2560 System Pressure-(psia) 2250
.2250 J2250 Core! Inlet: Temperature'('F).
557.0 551.2 550.6 g
Thermal Design Flow (gpm) 378,400 298,800
.298,800
' Average. Linear'PowerLDensity(kw/ft) 5.434 3.532 4.078 1.3. CONCLUSIONS I
From the evaluation presented in this report, it is concluded that the Cycle 2-design does not cause the'previously acceptable safety' limits for any incident to be exceeded; This conclusion is based on the following:
1.
Cycle:1 burnup'of 18,000 h0 MWD /MTV.
'. 2.
Cycle 2 burnup-is-limited to the end-of-life full power capability *
(nominally 16,000 MWD /MTU) plus a 500 MWD /MTU power coastdown.
3.-
Thet. 3 adhei ace to plant operating limitations given in the plant technical' specifications and those referenced in Section 4.0.
4.
T'here'is adherence to the post-LOCA long term cooling requirement contained in Reference 12.
i
.l Definition:
Full-rated power and temperature (approximately 587'F Tavg),
control rods fully withdrawn and 0 to 10 ppm residual boron.
2 2078P 6-870813
,n h
2.0 REACTOR DESIGN L
2'.1 -MECHANICA'L DESIGN The mechanical design of ~ th' Region 4 fuel assemblies is the same'as the e
Region 1-3. assemblies, except for the use of reduced length (.387 in. versus
.530'in., FSAR Tables 4.1-1 and 4.3-1) chamfered pellets-(standardized pellets), 4g" pellet.holddown spring, 304L stainless steel grid sleeve material.
on the top and mid grids; 'a bullet-nose design bottom end plug, and~a new top nozzle spring' screw design.. Table.1 compares pertinent design parameters of the various fuel regions.
The Region 4 fuel pellets are the standardized chamfered' pellets which have an effective dish which maintains the same U0 1 ading in kgU/ as'sembly as the 2
previous non-chamfered pellets contained in the initial core fuel assemblies.
The pellet chamfer and 4g holddown spring will reduce pellet chipping during manufacturing and handling..The change in grid sleeve material from'304 stainless steel to 304L stainless steel further educes the already low potential for stress corrosion cracking of the grid sleeves. The fuel rod bottom end plug was changed from a chamfered.end to a radiused (bullet nose) end to improve rod loading and reduce the potential for grid damage during rod
' loading..The new top nozzle spring screw design includes a fillet between the head.and shank,.which reduces.the stresses at this point, and therefore the likelihood of screw failure.
The chamfer in the uppermost spring was increased to accommodate this new screw design. The Region 4 fuel has been designed according to the fuel performance model of Reference 3.
The fuel is designed and operated so that clad flattening will not occur, as predicted by the Westinghouse model, Reference 4.
The fuel rod internal pressure design basis, in Reference 5, is satisfied for all fuel regions.
I 3
1978F1-870813 i
L Westinghouse's experience with Zircaloy clad-fuel is desciibed in WCAP-8183, 1:
i"06erationalExperiencewithWestinghouseCores," Reference 6.
This report is a
updated' annually.
.2'.2 NUCLEAR DESIGN u
L f
The Cycle 2 core. loading is designed to. meet a F x P ECCS limit of 5-g 2.32 x K(Z)* for four loop operation and 5 2.60 x K(2)* for' three' loop j
I
. operation.. The flux difference.(AI) bandwidth during normal. operation-conditions is +3,.-12 percent AI for four loop' operation and +5,-5 percent' g
al for three loop operation..
Table '2 provides a summary of. changes' in the Cycle 2 kinetics. characteristics compared with the current limit. based on previously submitted accident' i
analyses,-Reference 2.
The Cycle 2 values fall within the current limits L th -
8 the exception of the most positive Moderator Temperature Coefficient, the i
least-negative Doppler-0nly Power Defect at BOC, the Delayed Neutron Fraction and the maximum Differential Rod Worth.
. Table 3 provides the' control rod worths and requirements at the most. limiting condition during the cycle.
The required. shutdown margin is based on previously submitted accident analyses, Reference 2.
The available shutdown margin exceeds the minimum required.
The loading shown in Figure 1 contair,s a' total of 448 fresh burnable absorber rods located in Region 4 fuel assemblies.
The locations:of the burnable j
absorber and secondary source rods are shown in Figure 1.
+
2.3. THERMAL AND HYDRAULIC DESIGN l
.No significant variations in thermal margins will result from the Cycle 2 reload.- Sufficient DNB margin exists for all events to meet the design criteria (References 2 and 13) for the Cycle 2 reload core.
1 K(Z) - See Figures 2 and 3.
4 3978F-6-870813
--_L L _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
Westinghouse has received questions' from the NRC concerning the applicability of the W-3 DNB correlation at low pressures (i.e., below 1000 psia).
Reference 13 was provided to the NRC in response to these questions and justifies a new limit DNBR for lower pressures. A correlation limit of 1.45 is applicable for pressures between 500 and 1000 psia. This represents a change from the original limit of 1.30 which is still applicable for pressures above 1000 psia.
i In Reference 13 Westinghouse stated that this letter would be referenced in response to future NRC questions on the subject (in other words, Westinghouse considers the response to be generic) and committed to apply the as yet unapproved limit DNBR in the future. Currently, the only Millstone 3 DNB transients that fall within the pressure range are the inadvertent opening of a steam generator relief or safety valve (FSAR Section 15.1.4) and the steam system piping failure (FSAR Section 15.1.5).
The safety analyses for these transients are based on a 1.45 DNBR limit.
The DNB core limits and safety analysis used for Cycle 2 are based on conditions given in Sections 1.0 and 3.0.
Fuel temperatures were calculated using the revised thermal safety model, described in Reference 7, and include the effects of chamfered pellets.
4 i
i l
5 l
3,,,,...,.,
+
I'i ;
th
,, y I
4 ig, j. ( ; :.
J 3.0' POWER CAPABILITY,AND ACCIDENT EVALUATION
.3.1 POWER CAPABILITY' y
?
E The plant power capability for three and-four loop operation has been
]s
]
u evaluated considering theJconsequences of those, incidents examined in the-j; FSAR,(2) using-the previously accepted ' design' basis.~ It is concluded that J
'the: core reload will not adversely" affect:the ability to' safely operate at 100:
. percent of rated thermal ~ power (RTP).for.four loop operation and.65 percent of-RTP.for three loop' operation during Cycle l2.
For'the evaluation performed to
.i address overpower concerns, the, fuel centerline temperature limit.of 4700*F can beiaccommodated with margin?in the Cycle 2 core using the methodology Ldescribed in Reference 1.! The~ time dependent densification model(8) was j
used for these fuel temperature evaluations.
The LOCA limit at rated power l
can be met by maintaining F at or below 2.32 for four ; loop and 2.60 for g
envel pes (Figures 2 three loop operation according.to their normalized F0 and 3).
3.2.. ACCIDENT EVALU' TION A
L ~
The effects of the reload on.the design basis and postuisted incidents analyzed.in the FSAR were examined.- fiso evaluations were performed-for features described.in Section 2.1.
As a result' of.the planned incorporation of a positive MTC Technical Specification for. Cycle 2, safety evaluations were performed (Reference 11) for each of the following transients:
a.
RCCA Misalignment' b.
Startup of an Inactive Reactor Coolant Loop c.
. Excessive Heat Removal Due to Feedwater System Malfunctions
' ~ '
d.
Excessive Increase in Secondary Steam Flow -
e.
Main Steamline Depressurization/ Steam System Piping Failure E
.f.
Feedwater System Pipe Break g.
Inadvertent Operation of the ECCS at Power h.
Loss of Coolant Accidents
)
i -.
Steam Generator' Tube Rupture (SGTR) 6~
3978F 6-870813
___..___.j L___1____1_._
_._._..____.z___.__...__
The Safety Analyses were evaluated to consider the potential impact of the chamfered pellet fuel design. Additionally, an evaluation was performed to consider an increase in the RWST boron concentration, which occurs as a result of the positive MTC Technical Specification.
In terms of the LOCA and non-LOCA safety analyses the RWST boron concentration increase along with the features described in Section 2.1 were found to be acceptable, and the conclusions of the FSAR remain valid.
The current analyses for the inadvertent opening of a steam generator relief or safety valve and the main steamline rupture transients (FSAR Sections 15.1.4 and 15.1.5)'were reviewed with regard to the revised DNBR correlation limit discussed in Section 2.3.
The existing DNBR margins for these events were adequate to accommodate the increase in the limit DNBR from 1.30 to 1.45, without reanalysis.
Therefore the current FSAR conclusions for these two events remain valid.
A safety requirement that the core remain subtritical on soluble boron alone in long term cooling following a Large Break LOCA is adhered to in the Cycle 2 design. This is accomplished by calculating the post-LOCA core cooling water boron concentration using conservative volumes and boron concentrations for sources of water that would mix together to provide the resultant core cooling mixture boron concentration.
This resultant boron concentration is then compared to the cold zero power critical boron requirement to demonstrate that the core remains subcritical.
A core reload can typically affect accident analysis input parameters in the following areas:
core kinetic characteristics, control rod worths, and core peaking factors. Cycle 2 parameters in each of these three areas were examined as discussed below to ascertain whether revisions to the accident analyses assumptions were required.
3.2.1 KINETICS PARAMETERS A comparison of the range of values encompassing the Cycle 2 kinetics parameters with the current limits is given in Table 2.
All the kinetics values fall within the bounds of the current safety analysis limits except for 7
l une s-ows 1
the most Positive MTC limit (N & N-1), least negative Doppler Only Power Defect at BOC (N & N-1), maximum Delayed Neutron Fraction (N), and rod withdrawal from suberitical maximum Differential Rod Worth (N).
These deviations were addressed by the reanalysis described in Section 3.3).
3.2.2 CONTROL R00 WORTHS Changes in control rod worths may affect differential rod worths, shutdown margin, ejected rod worths, and trip reactivity. Table 2 shows that the maximum' differential rod worth of two RCCA control banks moving together in
-their highest worth region for Cycle 2 does not meet the current limit.
This deviation is addressed by the reanalysis described in Section 3.3.
Table 3 shows that the Cycle 2 shutdown margin requirements are satisfied.
3.2.3 CORE PEAKING FACTORS Peaking factors for the dropped RCCA incidents were evaluated based on the NRC approved dropped rod methodology described in Reference 9.
Results for N &
N-1 show that the DNB design basis is met for all dropped rod events initiated from full power.
Peaking factors following control rod ejection are within the bounds of the current limits. The peaking factors for the steamline break and misaligned rod have been evaluated and are within the bounds of the previous safety analysis limits.
3.3 INCIDENTS REANALYZED 1
As a result of the RTD Bypass Elimination, PMTC Technical Specification implementation, and the kinetics parameter changes noted in 3.2.1, a number of j
accidents were reanalyzed to support the Cycle 2 Reload.
The non-LOCA accidents with reanalysis documented in Reference 10 are as follows:
a.
Uncontrolled Control Rod Bank Withdrawal at Power b.
Inadvertent Opening of a Pressurizer Safety or Relief Valve c.
Loss of External Electrical Load / Turbine Trip d.
Main Steamline Depressurization/ Steam System Piping Failure (at power only) l l
l g
3976F 6-870813 I
_________.___________________J
- 9r
-i t
U LThe accidents with reanalysis documented;in Reference 11 are as follows:
a.-
Losslaf Forced Reactor' Flow-b..
ReactorLCoolant Pump Shaft' Seizure-f c.
Uncontrolled RCCA Bank Withdrawal at Power l
d.
Uncontrolled RCCA Bank Withdrawal /from a Suberitical Condition e..
-Inadvertent Opening of a-Pressurizer Safety or Relief Valve
~
f.
. Loss'of External Elect'r.ical Load /TurbineLTrip
'g.-
Loss ~of Normal'Feedwater Flow / Loss'of Nonemergency AC Power h.
RCCA. Ejection 1
i I
L i
i 1
-l I
i 4
1 i
d l
L J
)
1 9
3978F 6-470s13 h....
-l
I
'4.0 TECHNICAL SPECIFICATION CHANGES The changes to the Millstone Unit 3 Technical Specifications required for Cycle 2 operation are contained.in References 10 and 11.
i e
i 4
)
t 10 mu s-men c__________
1' --
l l
b 9
'J 5.0 REFERENCE
- 1.. -Davidson, S. 1.., et. al.', " Westinghouse. Reload Safety Evaluation ~-
j Methodology," WCAP-9273-A, July 1985..
-2.
" Final: Safety: Analysis Report Millstone Generating Station, Unit 3,"
USNRC Docket No.-50-423, Amendment 18,-March 1986.
3.'
Miller; J. V.-(ed.)~, " Improved Analytical Model 'used in Westinghouse Fuel Rod Design Computations," WCAP-8785, 0ctober 1976.
4.
George,! R. A., et. al., " Revised Clad Flattening-Model," WCAP-8381,. July 1974.
5.
Risher, D. H., et. al., " Safety Analysis; for. the' Revised Fuel Rod.
Internal Pressure Design Basis,".WCAP-8964-A, August 1978.
'6.
Skaritka, J.; Iorii, J. A., " Operational Experience with Westinghouse Cores," (through December 31,1985),NS-NRC-86-3814, September 1986.
7.
Leech, W.
J., et. al.. " Revised PAD Code Thermal Safety Model,"
WCAP-8720, Addenda 2, October 1982.
3 8.
'Hellman, J. M. (ed.), " Fuel Densification Experimental Results and Models for Reactor Operation," WCAP-8219-A, March 1975.
w -
9.
Morita, T., et. al., " Dropped Rod Methodology for Negative Flux Rate Trip Plants," WCAP-10298-A, March 1983.
10'. - Rice, W. R.; Sterdis, R.
J., "RTD Bypass Elimination. Licensing Report for Millstone Unit 3", WCAP-11496 (Proprietary), WCAP-11497 (Non-Proprietary), June 1987.
11 3978F 6-670813 l
p n.
l 1
,4 a
q u.
4
]
.11.
Westinghouse Letter dated August 14, 1987, 87-NE*-G-0100,." Cycle 2 PMTC Report". Positive Moderator Temperature Coefficient Licensing Report for.
v..
Millstone Unit 3," August 1987.
4 l
12.'- Bordelon,'F. M., et.' al., " Westinghouse ECCS, Evaluation Model Summary,"-
WCAP-8339, June 1974.-
- 13. Westinghouse Letter dated March 25,'1986,..NS-NRC-86-3116, " Westinghouse.
I.
Response to-Additional Request.on'WCAP-9226-P/WCAP-9227-N-P, " Reactor Core Response to Excessive' Secondary Steam Release,"'.(Non-Proprietary)."
j 1
- l 4 s
[..
1 I
'5-l wes-men
W.m-.. '
r-t
.i
.g
-TABLE 1
. n, 1
)
FUEL ASSEMBLY DESIGN PARAMETERS
-]
MILLSTONE UNIT 3 --CYCLE 2
-]
q
- =
Region:
2.
3 4A-4B i
- Enrichment (w/o U-235*)..
2.899-
'3.395~
3.50' 3.80-Geometric Density.'
94.965 94.980 95-95
(%i heoretical)*
t Number of Assemblies 45 64 56 28
\\
Approx'imate Burnup at 19,600 14,200 0
0
' Beginning of Cycle'2
' (MWD /MTU)*
4 I
i.:
All values are as-built values except Region 4
+ Based on estimated E0C1 of 18,000 MWD /MTV.
13 39787 6-870806 lC..., '.
p d
. TABLE 2 KINETICS' CHARACTERISTICS J
MILLSTONE UNIT 3 -' CYCLE 2 4
N AND N-1 LOOP OPERATION-Cycle 2 Changes'.
i'~
' Current Limit to Current Limits Most Positive: Moderator Temperature.
O from 0:to
+5 < 70% RTP
. Coefficient (pcm/"F)*
100% RTP
+5 to 0 Ramp from 70% to 100% RTP Most Negative Unrodded Moderator
-40
' Temperature Coefficient-(pcm/*F)*
1
. Doppler Temperature Coefficient,
-2.9'to -1.4 l-(pcm/*F)*
Least' Negative: Doppler - Only Power
-10.18 to -6.68 Coefficient,'Zero to Full Power (pcm/% power)*
~
'Least Negative Doppler - Only Power
-1055 (4 Loop)
-970 (4 Loop)
- Defect at B0C (pcm)*
-900 (3 Loop)
-650 (3 Loop)
Most Negative Doppler - Only Power
-19.4 t'o -12.6 Coefficient, Zero to Full Power
- (pcm/%' power)*
Delayed Neutron Fraction B,ff, (%)
0.44 to 0.60 0.44 to 0.70 (4 Loop).
(4 Loop) 0.44 to 0.75 c
(3 Loop)
(3 Loop)
Maximum Differential Rod Worth to 86.7 77.3 Two Banks Moving Together at HZP (pcm/in)*
-5 pcm = 10 3,
w 14
' 3978F 6-870806
-_L_---__.
l l
TABLE 3 END-0F-CYCLE SHUTDOWN REQUIREMENTS AND MARGINS MILLSTONE UNIT 3 - CYCLE 2 4 Loop (N) 3 Loop (N-1)
Cycle 2 Cycle 2 Control Rod Worth (%Ap)
All Rods Inserted 7.82 7.82
. All Rods Inserted less Worst 6.78 6.78 Stuck Rod (1) Less 10%
6.10 6.10 Control Rod Requirements (%Ap)
Reactivity Defects (Combined Doppler, 3.64 2.60 Moderator Temperature, Void)
Rod Insertion Allowance 0.44 0.56 (2) Total Requirements 4.08 3.16 Shutdown Margin [(1)-(2)] (%Ap) 2.02 2.94 Requirement Shutdown Margin (%Ap) 1.60 1.60 15 3976F 6 870606
I.
l-MILLSTONE UNIT 3 CYCLE 2
{
CORE LOADING PATTERN l
448 BA'S IN 56 CLUSTERS 1'
R P
N M
L K
J H
G F
E D
C B
A~
180*
2 46 3
48 3
4B 2
~
2 3
4B 3
4B 3
4B 3
48 3
2 (8)
(8) 2 3
4A 3
4A 3
3 3
4A 0:
4A 3
2 (8)
SS (8) l I
3 4A 2
4A 2
4A 3
4A 2
4A 2-4A 3
(8)
(8)
(8)
(8) 2 AB 3
4A 3
4A 2
4A 2
4A 3
4A 3
48 2
(8)
(8)
(8)
(8)
(8) 4B 3
4A 2
4A 2
4A 2
4A 2
4A 2
4A 3
48 (8)
(8)
(8)
(8)
(8)
(8) 3 48 3
4A 2
4A 3
4A 3
4A 2
4A 3
4B 3
(8)
(8)
(8)
(8)
(8)
(8)-
(8) 4B 3
3 3
4A 2
4A 2
4A 2
4A 3
3 3
48 (8)'
(8)
(8)
(8) 3 48 3
4A 2
4A 3
4A 3
4A 2
4A 3
48 3
(8)
(8)
(8)
(8)
(8)
(8)
(8) 48 3
4A 2
4A 2
4A 2
4A 2
4A 2
4A 3
48 (8)
(8)
(8)
(8)
(8)
(8) 2 48 3
4A 3
4A 2
4A 2
4A 3
4A 3
48 2
(8)
(8)-
(8)
(8)
(8) 3 4A 2
4A 2
4A 3
4A 2
4A 2
4A 3
(8)
(8)
(8)
(8) 2 3
4A 3
4A 3
3 3
4A 3
4A 3
2 (8)
SS (8)
{
2 3
48 3
4B 3
4B 3
48 3
2 (8)
(8)
J 2
4B 3
48 3
48 2
0*
N YS bD FIGURE 1 i
16 3978F 1-470804
f, u,
l.
ll I ',
t
-s a.
t.
3 1
.e
_Ij f I
l-jj
_j 4
.1
'F l.
j H. R '^
- 1. 50 -
,8 t- - -
1 1
.t 1.~ 2 5 - --
/<
i
~
'i g-1.00
~
l L
w--
l o.
w
- 3 0.75 -
7 7^V 2.320 CORE HEIGHT K(2)
Z
.0.000 1.000 o
5 20 t 000 L
O. 50 - -
t:!
10.800 0.94 0'
',9 i 12.000 0.647
./
s t
I I' -
^O.25 -
4 a.q '
1 q
O.00 D
0
'2 4
16 8
10 J
12 x
,'s CORE HEIGHT (FT) 1' A!
a di
(
s.
i FIGURE 2 i
i K(2) - NORMALI2ED F (Z) AS A FUNCTION OF CORE HEIGHT
)
g FOR FOUR LOOP OPERATION r.
17 3974P 6-470806 I
. _ - l-
TJ::g7:; ',
a?,
(M ;;c m
. ~-
..e c.
y, s.
iM p
I 9.( ;
lv e-
)-
Tl ja'j.
/, -)e c9 f.
?,
1!!-
- 1. 50'
~
p-
!.jE t,
o-
.g 2
)
1.' 2 5 - -
ki e,7
.p.e-s p.
g u.
9
.1 :. 00 a-
-o w
..(
C w
o N
3 O.75 -
' TOTA' i
2.6o0 7-
. CORE HEIGHT K' ;)
1 E
i.
I O2 z o.ooo -
3,ooo i
C.50 -
6.000 1.ooo
,10.800 0.940 12.000 0.577 li\\Y ky,.,
- 1
' ' : 0. 2 5 : - -
J.
ai
,\\
-n.
O.00
, ~.
l',
,O 2
4 6
8 10 12 a
' CORE. HEIGHT (FT) l m
- Og FIGURE 3-1 K(Z)'- NORMALIZED F (2) AS A FUNCTION OF CORE HEIGHT
.g :
0 f0R THREE LOOP OPERATION q
/,.
. 3878F1-870806 t;
'i '
- N.l ~ ]
5
.i
]
>..