ML18117A361
Text
--
A B
c D
E F
I G
w-H 6
K L
I M
N 0
p R
I 2
x I
6 3
5 7
8 7
3 5
4 8
6 2
5 4
2 7
2 7
5 4
2 8
6 2
3 5
4 7
8 7
3 5
I 6
3 4
5 6
7 8
z WORTH-HZP ( /. ~ k/k)
BOC EOC SEE TABLE 3.2-7 D Group Number 5
4 2
2 4
5 9
I 3
8 7
5 3
6 8
I 4
5 2
7 6
-Y 4
5 6
8 I
5 3
8 7
3 I
10 I I 12 13 14 15 Group No.of rods Function I
8 2
8 3
8 4
8 5
12 6
8 7 (TRANSIENT) 9 8
8 TOTAL # 69 Safety Safety Safety Safety Control Control Control APSRs BilUNuole...
Update-12 3/94 p.3.FIG-1 TMI ~tt I Control Rod Locatton and Group Oeatgnatton* For TMl-1,Current Cycl*.
CAD FILE:SIA,SKM.00,0343,000-,0001 Ftg.3.2-1 Rev. 21, 04/12 61
x A
B c
D E
F G
W-H
-Y K
L M
N 0
p R
I 1l21314151+1sH+11~1J1415 RODS IN I. GROUPS 5-7. or. WO
- 2. GROUP 8 AT HFP NOMINAL POSITION.
BOC AND EOC EJECTED ROD p.J.FIG-2 z
HZP WORTH OF EJECTED ROD ( r. A ldk)
SEE TABLE 3.2-4 BiBINucle...
TMI Untt I Upaote-12 3/94 Elected Rod Locatlon BOC and EOC Co4 File SIA.me.oo.cam.000-.0001 Flg.3.2-2
TMI-1 UFSAR RODS IN HZP WORTH OF STUCK ROD (%k / k)
Groups 1-7, 0% WD See Table 3.2-4 BOC Maximum Worth Stuck Rod EOC Maximum Worth Stuck Rod
- p. 3.FIG-3 A
B C
D E
F G
H K
L M
N O
P R
1 2
3 4
5 6
7 8
9 10 11 12 13 14 15 TMI-1 UPDATED FINAL SAFETY ANALYSIS REPORT Stuck Rod Location BOC and EOC FIGURE 3.2-3 Rev. 23, 04/16
.28
.18
.16
.14
. 12
~ -
.10 0....
~ -
0 a....
.OI
.O&
- °'
02 Ct 0
UDIUS S I SLOPE (11 P>
SLOP£ (EIPERllENT>
1t*>
(c-2 a*>
(I 10-2)
(I 10*2) 25
.7143 I 00
- 0. 11 ~
DI
.25
!10
. 3921 0 83 0 71 ~
DI I 0 19&1 0 71 0.61 ~.05 2.0
.0910 OH 0.&3 ~ 05
--RIP
- - - HELLSTUND. ILOHU;. HORNER r a 1. 0 2
I I I 10 IDG K..rr. JJiO Id ijl I Nuclear TMI Unit 1 I&
r
- 2. 0 I II 1200 "
Update
- 5 7/86
- p. 3.FIG-4 Fractions of Change in the Reasonance Integral as Function of VT-v'BJ for UO 2 Rod (T in Degrees K)
Fig. 3.2*4
0.00
- 0.02
- 0.04
~
- 0.06
- 0.08
~ *O. I 0 0
.,..... 0. 12
- O. 14
. 0. 1 &
- O. 18
- 0.20 10
- p. 3.FIG-5
~
~ri II 00 PPM 20
"'~ " '\\.
JO
'Void.
40 ld5J11Nuclear TMI Unit 1 50 60 Update - 5 7/86 Uniform Void Coefficient for 177 Assembly Core Fig. 3.2-5
110 100 w
~
u
_J 80 H
~
H M~. 60
- w ffi d
=- >
~u
~~ 6'0
~~
- > H
~i 20
(!)
tjl -
0 0
\\
\\
\\
\\\\
\\\\,
.. V-2.4°/o6k/k(wtth Stuck Rod}
1
\\
\\'-
~...... __ *---. ~
5.4o/o Ak/k y~
I 2
e Ttme.sec p.3.FIG-6 6
7 1 ca:im Nuclear TMI Unit 1 Percent Neutron Power Versus Time Following Trip, BOC 8
Update
- 9 7/90 Fig. 3.2-6
TMI-1 UFSAR
Core Loading Diagram for TMI-1 Cycle 22
TMI UFSAR Figure 3.2-8 Deleted
~
4 N
~
N t * =
0 *-..,
~
GI:
CD z
Cl
- 2. 7 2.8 2.5 2.4 2.3 2.2
- 2. 1 2.0
- 1. 9
- 1. 8 50 70 90 110 130 150 Distance fro* notta* at active len1tn. in.
- p. 3.FIG-9 Id ijl I Nuclear TMI Unit 1 Update - 6 7181 Typical DNB Ratios (BAW-2) in the Hot Unit Cell (Cycle 5)
Fig. 3.2-9
I.II 1.11 '---"-..
1.21
~
u
- c: *
~ -
u
- :a
~-
D.11
~ z::
4 I
D.10 D.40 0.21 D.OG D
""111111 ~
~ '
i\\..
\\
' *~
~'\\
20 30 40 51 10 70 10 P1rc1nt111 1f fuel lods with Mi1h1r P11kin1 F1ct1rs Than Point Values. I ld 5Jl INuclear TMI Unit 1
- p. 3.FIG-10 Distribution of Fuel Rod Peaking (Initial Cycle)
~
IO 1DG Update - 5 7/86 Fig. 3.2*10
2.0 1..8 LG 1.4
-;c
~
1.2 l 1..0 J 0.. 8
- I
~ 0.. 6 0.. 4 0.. 2 0.0 20 40 60 Active fuei
- p. 3.FIG-11 80 100 Len&th, Inches
(?i 5Jl I Nuclear TMI Unit 1 120 Alia I Peaks
- 1. 7 140 Update - 5 7/86 Maximum Allowable Axial Power Distributions for a Radial-Local Peaking Factor of 1.78 Fig. 3.2*11
2.1 I. I I. I I. 4
~
1.2
.. I. I
-I I.I J I.I
- I
~
1.4 1.2 l.D 0
2D A111 I P111L1 41 H
ID IOI 120 141
&ctu1 fuel Len&U*. l*chu
- p. 3.FIG-12 1?1 ijl J Nuclear TMI Unit 1 Equivalent Axial Power Distribution for a Radical Local Pe1klng Factor of 1.65 (Initial Cycle)
I. I 1.4 Update - 5 7/16 Fig. 3.2*12
0-nJ u
c m
z Q
0-
'O a.>
u
'O a.>...
- a.
~
a.>
c c nJ
.c LI Cl) a.>
0 c -
0 nJ a:
m z
Q 2.0
- 1. 8 1. 6 1. 4 1. 2 1. 0 0.8 100 99% Confidence Basis I
(114%)
~Design Overpower 110 120 130 140 REFERENCE DESIGI llOWER ( 256 8 II t), %
Id iJJJNuclear TMI Unit 1
- p. 3.FIG-13 DNB Ratios (W-3) in Hot Unit Cell Versus Reactor Power (Initial Cycle) 150 Update - 5 7/86 Fig. 3.2-13
20 t8 16 I 4 I 2 I 0
~ -- B
~
Cl 8
4 2
0
- 2
- 4 100 2120 ps11 2115 psi e Qu1l 1 ty SuDcooled II 0 120 130 140 150 REFHHCE DESIGll P'OWIR ( 2561 Ml t), \\
- p. 3.FIG-14 Id ijl I Nuclear TMI Unit 1 Maximum Hot Channel Exit Quality Versus Reactor Power (Initial Cycle) 15 17 c:..
90 93 - "'
.. Cl.
- a Cl Q N
H
.c: -
UN
.c: -
99
.c: -
GI u
c
- a..
I 03... _. -..
0
~ -
109 c
u
"""c
.....c:
u
- c:
116 0
0
... ct 127
.c: -
De c..
144 _.
160 Update - 5 7/86 Fig. 3.2-14
N a
N I
ca. ca.
Ba
- s a A.~
g a
GI a m a.....
a..
~
- CD co N
~
~
A.
z
\\:J M
Cit....
a u z
~
~
DNIR In Hot Channel
- p. 3.FIG-15 l?j iJ1 I Nuclear TMI Unit 1 Update - 5 7/86 Hot Channel DNB Ratio (W-3) Versus Power for Partial Pump Operation (Initial Cycle)
Fig. 3.2-15
0.
2 CL E
~
A.
CL a a
~
N I
CL E
~
A.
N 0
0 0
0.
0 S?
8 0
Cft i
0.....
0 co
~
I -
ca.,,
N.
Ck
~
5
~
a w
u
- w
- w w
- Coolant Quality At Point of lini*u* DNIR In Hot Channel
- p. 3.FIG-16 lljijJJNuclear TMI Unit 1 Update - 5 7/86 Hot Channel Quality at Point of Minimum DNBR Versus Power for Partial Pump Operation (Initial Cycle)
Fig. 3.2-16
- .*r-------...... -------..------...... --------..... ------..... ---
I I
- I N
3.011--..-.-e--_.,. ______ ___,..._ ______ -+-______ _.,. ________...._ __ _.
.. -I..
I
~
~
~
T a.. -
N
... a
-=>>
M c: -.
Q..
u --
a.,, *-
D1t1 l111d On lml.
CEAP*412'
( / k dt
- 11 */m)
I I
- 2. *.,_ ___...,_....., __ _,.._,._ _ __. ____,._. ___..._.. _ _.
u I
~
- a...
- 1..... ________________..... ______...... ______ __. ________...... __
~
I 2HI
- p. 3.FIG-17 1111 1?1 iJl I Nuclear TMI Unit 1 Thermal Conductivity of U02 5080 Update - 5 7186 Fig. 3.2-17
70
&O so
~
c: -
0 A. -
40 0
u a
30 Ii
~
z 20 10 0
0.6 0.8
- 1. 0
- p. 3.FIG-18 Gaussian Distr1but1on
- 1. 2 1.4 Id ijJ I Nuclear TMI Unit 1
- 1. 6 Number of Data Points vs. <t>El<t>C
- 1. 8 Update - 5 7/86 Fig. 3.2-18
'C G.-
u G
~
ca...
A.
c ca
~
CQ
=
Q.
ca A.
100 90 BO 70 60 Inf tnite Sample lOOS Confidence F i n i te S amp I e 90% Confidence Fini te Samp I e 99% Conf tdence 50.........._ ____________ _,_ ______ _,_ ______ _._ ______ -..L ______ __J
- 1. 0 1. 1
- 1. 2
- p. 3.FIG-19 1. 3 1. 4 DNB Ratio (BAl-2)
Id ijl I Nuclear TMI Unit 1 1. 5 DNB Ratio (BAW-2) vs. Population for Various Confidence Levels
- 1. 6 Update - 5 7186 Fig. 3.2-19
L 025 i.020 I. 015
- 1. 010
- 1. 005 Q -
u ftl w..
I. 000 G1 c c l'Q 0.995
.c
(.) -
Q z 0.990 0.985 0.980 0 975 0.970 0.965 0.960 60 70 80 90 FA (Wall Ammbly Channel)
Population Protected. '
- p. 3.FIG-20 Id 5Jl I Nuclear TMI Unit 1 Hot Channel Factors vs. Percent Population Protected 100 FA (Interior A111mbly Channel)
Update - 5 7/86 Fig. 3.2-20
~-
16 14 12 10 8
6
-.......'?
<)
4 CQ 2
~
c 0
. 2,
- 4
-6
-8 100 Ii th 5 'Yo FI ow Factory /
D i s t r i but i on
/
~.v
/".v
!/
.,. K No Flo*
~
.. V v
0 I s tr I Du t I on Factor
/
I I
v v
I I
/
/
QuaJ1t~
/ //
~
SuDcooled
/
/ /
11 0
- p. 3.FIG-21
~
Des I gn o~er power I
120 130 140 Rated Power (2568 MWt). ~
18 iJl I Nuclear TMI Unit 1 Hot Channel and Nominal Channel Exit Qualities Versus Reactor Power (Without Engineering Hot Channel Factors)
(Initial Cycle) 150 Update
- 5 7/86 Fig: 3.2-21
- Bundle Burnout Test Conditions Where Stable Operations Were Obaerved
~ Maxi*u* Design Conditions, 11~3 Power
- MaxiMum Deaign Conditions, 1303 Power
~Most Pr.obable Conditions, 11~1o Power
- Most Probable Co~itions. 130~ Po\\118r
.. ~
i.s...... ~------'-~--~--'---------+-----~~+-----~--...... ---------t s
I. 0
.s 0
5
- p. 3.FIG-22 lubble To Annular Bubble To Slue Baker (laker) 10 15 20 Qual i t1 (lb vapor/t~tal h), ~
Id ijl I Nuclear TMI Unit 1 25 Flow Regime Map for the Hot Unit Cell Update - 5 7/86 Fig. 3.2-22
3.0 2.5 w
I 0
JI(
N.,
2.0
~
I I: -
4 u
~
- I. 5 s
1.0
. 5
-5 0
+
Bundle Burnout Test Conditions lhere Stable Operations lere Observed.
Maximum Oes1an Cond1t1ons. 110 Power Maximum Oes1an Cond1t1ons, 130' Power \\
Most Probable Cond1t1ons, 110 Power Most Probable Cond1t1ons. 130% Power
+
+
r+
+
+
+
.... * +
++ +~ +..,..
+
+ +
f+.
+
\\
+ +
't ~ +
I+
+ + +
t
't'...
\\
lubbl* To
....i..... +, ~ +
~
+
Annular
+
(laker)
.. + rt++++
+
++
++
... t ** +4A + '
+
+
+
+
+ + +
+
++*
+
N ~
+
lubbl* To SI ug (laker)
.. ~
~
5 10 15 20 25 Quality (lb vapor/total lb),~
Id ijl J Nuclear TMI Unit 1 Update - 5 7186
- p. 3.FIG-23 Flow Regime Map for the Hot Control Rod Cell Fig. 3.2-23
3.0 2.5 I 9 llC N..
2.0 I
~
.I:.
~
u 0..
I. 5 s
1.0
.5
-5
- Bundle Burnout Test Cond1t1ons lhere Stable Operations Were Observed.
- Mu1111um Des1an Cond1t1ons. 110 Power
- Maximum Oes1an Cond1t1ons. 130\\ Power
\\
~Most Probable Cond1t1ons. 110 Power
- Most Probable Cond1t1ons.
13Q~ Poter
\\
I
\\
lunle To Annular (laker)
- "' ~
lubbl* To J
Slug (laker)
_/
0 5
- p. 3.FIG-24 10 15 20 Oualit1 (lb vapor/total lb},~
l?j ijl I Nuclear TMI Unit 1 25 Flow Regime Map tor the Hot Wall Cell 30 Update
- 5 7/86 Fig. 3.2-24
3.0 2.5
'° I
2 2.0 N....
op I...
.I:. -*
u I. 5 0 u
>........ z 1.0
.5
-5 0
~undle Burnout Test Cood1t1ons Where Stabl!
Were Observed Operations Maximum Oes1an Cond1t1ons. 110 Power Maximum Oes1an Cond1t1ons. 130'\\ Power \\
Most Probable Cond1t1ons. 110 Power Most Probable Cond1t1ons.
130~~ Power
\\
4 Bubble To Annular (Baker)
~-
t Bubble To l J
Slug ( laer)
~
~
5 10 15 20 25 30 Quality (lb vapor/ total lb), '4 11 ijJ I Nuclear TMI Unit 1 Update - 5 7/86
- p. 3.FIG-25 Flow Regime Map for the Hot Corner Cell Fig. 3.2-25
150 140
~
0 130
~
s::
.a.
- 120 0
~
0 -
u....
GI:
110 0....
100 90 2400 Desian Flow Rate
_J_ (
131.32 l 106 lb~
2600
- p. 3.FIG-26 I
Desian Overpower
~
(1141 l 2561 llt)
I DNBR (1*3)a 1.30 2800 3000 3200 3400 Reactor Core Power. llt l?j illl 1 Nuclear TMI Unit 1 Reactor Coolant System Flow Versus Power (Initial Cycle) 3600 Update
- 5 7/86 Fig. 3.2-26
2.4 2.2 2.0
("")
I
- I. 8 a
~
ftl ac CD
- z Q
- 1. 6 cu c c ftl
.c
~
~
- 1. 4 a :c
- 1. 2 1. 0 0 I 100 LINE FLOI MIXING COEFF.
1 1101
.02 cJ 2
1001
.02 u
3 901
.02 ll 4
IOOS
. 06 "
5 1001
. 0 I "
1.30 (1*3) -
110 120 130 140 150 REFERENCE DESIGN llOWER ( 2511 Rt) I
- p. 3.FIG-27 i?j ijl I Nuclear TMI Unit 1 Update - 5 7/86 Hot Channel DNB Ratio (W-3) Versus Power with Reactor System Flow and Energy Mixing as Parameters (Initial Cycle)
Fig. 3.2-27
5200 4800 4400
.; 4000
~
u Cl.
E IU t-3600 u..-c u
(.)
u
~
u...
3200 2800 2400 6
uo
~Desi&n r Overpower (114SJ
- 0. 0095" CI ear a nee
~
100i Power 8
10
'--Maximum Design Clearance Nominal Clearance 12 14 16 18 20 22 linear Heat Rate. kw. ft Id 5JJ I Nuclear TMI Unit 1 24
- p. 3.FIG-28 Fuel Center Temperature for Beginning-of-Cycle Conditions (Initial Cycle) 26 28 30 Update - 5 7/86 Fig. 3.2-28
5200 uoo 4400
~ 3600 c:
cu
~
3200 2800 2400 6
~Design Overpower (114%)
.0095" Clearance 8
10 100% Power
"--Maximum Design Clearance Nominal Clearance 12 14 16 18 20 22 Linear Heat Rate. kw/ft
- p. 3.FIG-29 Id CIP I Nuclear TMI Unit 1 Fuel Center Temperature for End-of-Cycle Conditions (Initial Cycle) 24 26 28 30 Update - 5 7/86 Fig. 3.2-29
.a -
=i -....
- Q. *..
=i
-411' 5200 4800 4400 4000*
3600 3200 2800 2401 2000 160D 1200 BOC (100 MID/ITU)
- p. 3.FIG-30 Id ijl I Nuclear TMI Unit 1 Update - 6 7/87 Typical Post-Initial Cycle
- Center Line Fuel Temperature vs. Linear Heat Rate Hot Pin (Cycle 5)
Fig. 3.2-30
u... = -
IW...
u i
5000 4800 4600 u...
c::
u u
u =
""" 4400 u...
- a -
IW...
4200 5000 4000 I
u -
c::
u u
3000 2000 0
0 8
4
- p. 3.FIG-31 16 17.63 kwlt t Hot Spot ( 1001 Power) 24 32 40 Burnup (ll01ITU a 10*3, I
EDL 110
- 40. 900 Fu 16 20 L1n11r H11t R1t1. kw/tt C!Iim Nuclear TMI Unit 1 Burnup Effect on Fuel Center Temperature (Initial Cycle) 24 Update - 5 7/86 Fig. 3.2-31
lL
&)
=> -
&)
- a.
E
&)
&)
- I 2400 2000 1600 0
20
- p. 3.FIG-32 40 60 Yolu*e Fraction of Total Fuel. I
<at or aoove Fuel Tnperature)
BE Nuclear TMI Unit 1 80 100 Update - 5 7/86 Fuel Temperature Versus Total Fuel Volume Fraction for Equilibrium Cycle at End-of-Cycle Fig. 3.2-32
- 1. 012 0.854
- p. 3.FIG-33 0.92&
- 1. 326 18 ijl I Nuclear TMI Unit 1
- 1. 282
- 0. 919 Nu*D*r Crcl1s lurned Anl*Dlf P/P Update
- 5 7/86 Typical Reactor Fuel Assembly Power Distribution at End-of-Cycle Equilibrium Cycle Conditions for 1/8 Core Fig. 3.2-33
0 IU
~ -
"° Cl)
~
E IU to-3200 2400 2000 1600 800 Fu e I 10 kw/ t t 6 k w/t t S80°F-Tav11 Cool8.::i i Cladr-1 I
I 400 --------------~----------------------------------
- 0. 0 0.04 0.08
- p. 3.FIG-34
- 0. 12
- 0. 16 fuel Radius, in 18 ijl I Nuclear TMI Unit 1 0.20 0.24 Update - 5 1186 Fuel Rod Temperature Profiles at 6 and 10 KW/Ft Fig. 3.2-34
8 8
~
~b j
I 4
0 1... '*
~
It 0
+
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- 4
~
4
\\ *
~
(~
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10:
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4 8
8 8
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I I
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._ ~ _,9--
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I
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ii n
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- I
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~ § *
- a.
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N u....
~ i u.... -
N u..
c
§ N..
I 0
=--
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§ I
.. § f i11llft C11 l1l1111d, I
- p. 3.FIG-35 ld 51P J Nuclear TMI Unit 1 Update
- 5 7/86 Percent Fission Gas Released as a Function of the Average Temperature of the U02 Fuel Fig. 3.2-35
1.1 I
..,.__ _ _..,_..,._./_,...'__....£... P/P
- u
\\I I
\\
J_
I
- 1. 70 (partial Rod Insertion)
- 1. 50 (lodif i1d Cosine) 1.4L-----~-----+----~.... --..-...... --~~......
I 3M D1y1*IVIU a-of-Life IU/ii I
~ ~
0.11--~~....... -----oj~--....... -........... --..... +---.... -+-.......... --t Core l*tt*
I 20 40
&O I
.--~ Fu* I lidp I 1n1 I I Core Top 144* ---~
1 --~--__.,....
10 110 120 140 D11t1nc1 tr111 Iott.. of Acti** Full, in.
- p. 3.FIG-36 Id ijl )Nuclear TMI Unit 1 Update - 5 7/86 Axial Local to Average Burnup and Instantaneous Power Comparisons Fig. 3.2-36
2400 2200 2000
~
0 1800 I
c-....
I... 1600
=
- l -
CD cu 1400 u =
T" 1200 u
- )
l:2 c:
0 u
1000
- c.
ftl
~
"O 800
~
u 0 -
600 cu
- )
400 200 0
0 8
Nomi na I D1a*etra1 Gap 12 16 lax imum Desi i" Di ametral Gap 20 24 l 1 n e a r H e a t A a t e,. k */ f t
- p. 3.FIG-37 18 CIP I Nuclear TMI Unit 1 Update - 6 7181 Fuel to Clad Gap Conductance for End-Of-Cycle Conditions (lnltlal Cycle)
Fig. 3.2-37
1::1 u
20 15
~ 10 5
0 D
1.1 A1ial Power and EOL Burnup Snape with Closed Pores 1.7 A1ial Power ana EOL Bu rnup Snape
- i th Op_ en ~
Pores 1.5 A1i1I Power ana EOL Burnup Snape with C I o sea Po re s 2
4 luimu11 Desi1n Clearance 6
I I
I
~
I I
I Initial Cold Diametral Clearance, in 1 103 18 ijl I Nuclear TMI Unit 1 10 Update - 6 7/87
- p. 3.FIG-38 Fission Gas Release for 1.5 and 1.7 Max/Avg. Axial Power Shapes (Initial Cycle)
Fig. 3.2-38
3500 3000 2500
&It
- a.
'C u
m 2000
=. -
u
'C -"'
c u -
~
1500 u -
A.
ftl c.a 1000 500 0
Desi an L i11i t 114% Overpower
10H Power C loseo 1. 7 Ax 1 a I Po we r an EOL Bu rnup Snape 1.5 Axial Power and EOL Burnup Snape Pores 1. 7 lxi al Power ana EOL Burnup Snape M111*u* Desi&n Clearance_!_,JI 2
4 6
8 10 Initial Cola Dia.etral Clearance. in x 103
- p. 3.FIG-39 18 ijl I Nuclear TMI Unit 1 Update
- 6 7/87 Maximum Gas Release to Pressure Inside the Fuel Clad for Various Axial Burnup and Power Shapes (Initial Cycle)
Fig. 3.2-39
t-* -
.t I
I
© HOJ UN I J C fl l CD "°'
Ull CELL HO J COINU CELL
© HOT COhUOl IOO
- p. 3.FIG-40
, 1 lk 1 n a F 1 ct o r lnttulpf 11111 Factor CHL llj ijllNuclear TMI Unit 1 Update - 6 7/87 Nominal Fuel Rod Power Peaks and Cell Exit Enthalphy Rise Ratios (lnltlal Cycle)
Fig. 3.2-40
~--
t I
©
~
v
~
MOT UNIT CELL MOT llll CELL MOT CUNEa CHL MOT CONU OL IUD
- p. 3.FJG-41 CH l hclur '""I f1ttor lnt1111u *111 facur Id iJJ I Nuclear TMI Unit 1 Update - 6 7/87 Maximum Fuel Rod Power Peaks and Cell Exit Enthalphy Rise Ratios (Initial Cycle)
Fig; 3.2-41
I. I I. 5 1. '
- 1. 3
- 1. 2 1. 1 co
- 1. 0 I Cl N
0.9 I...
- a 0.1
- 0. 7 IC =
~
O.&
z -..
0.5 u
0
~
0.4 0.3 0.2 0 I 1 0
\\
G. 2.21 I 10* lll/ttr-ft2
\\. \\
I\\:
1-3 DNI M11t Flu1
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(D1111n Li*it)
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lin 111u* DN*
- 1. 55
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~
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/
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v I
I Calcul1t1d Surt1c1
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Miit Flu1 l
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540 510 510 600 620 140 110 610 700 720
- p. 3.FIG-42 Loc11 Enthalpy, ltu/lb Id ijl J Nuclear TMI Unit 1 Update - 5 7/86 Calculated and Design Limit Local Heat Flux Versus Enthalpy in the Hot Unit Cell at the Most Probable Condition (Initial Cycle)
Fig. 3.2-42
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- 5 7/86 Calculated and Design Limit Local Heat Flux Versus Enthalpy in the Hot Unit Cell at the Maximum Design Condition (Initial Cycle)
Fig. 3.2-43
CONTF,OL ROD ASSEMBLY PLENUM ASSEMBLY OUTLET NOZZLE COPE BAPREL
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- p. 3.FlG-45 CONTROL ROD ASSEMBLY LDCATION INCORE INSTRUMENT LOCATION REACTOR VESSEL THERMAL SHIELD CORE BARREL SURVEILLANCE SPECIMEN HOLDER TUIE Id ijJ I Nuclear TMI Unit 1 Reactor Vessel and Internals - Cross Section Update - 5 1186 Fig. 3.2-45
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COUPLING SPIDER TOP VIEW NEUTRON ABSORltNG MATERIAL~.
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SPIDER----
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- p. 3.FIG-52 1?1 ijlJNuclear TMI Unit 1 Burnable Poison Rod Assembly Update
- 5 1186 Fig. 3.2-52
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- p. 3.FIG-54 tOJSJNG ASSY Id C):JJNuclear TMI Unit 1 Side View of BPRA Retainer Update
- 5 7/86 Fig. 3.2*54
- p. 3.FIG-55 UPPER CORE PLAT! ASSY PAO TYP Id iJl J Nuclear TMI Unit 1 Top View of BPRA Retainer During Operation Update - 5 7/86 Fig. 3.2*55
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19, APRI L 2008 p.3.FIG-3 Stuck Rod Location BOC and EOC FIGURE 3.2-3