ML16127A483
Text
x A
B C
0 E
F G
w-H K
L M
N 0
P R
1 6
1 3
5 5
3 7
8 7
8 7
3 5
4 4
5 3
1 8
6 2
6 8
I 5
4 2
2 4
5 6
7 2
7 2
7 6
5 4
2 2
4 5
1 8
6 2
6 8
I 3
5 4
4 5
3 7
8 7
8 7
3 5
5 3
I 6
I I
2 3
4 5
6 7
8 9
10 I I 12 13 14 15
-y No.of I"'ods FuncTloo Z
WORTH-HZP
(/. ~ k/k)
Group Number Gr'OUP I
8 2
8 3
8 4
8 5
12 6
8 7 (TRANSIENT) 9 8
8 TOTAL # 69 Sofety Sofety Sofety Sofety ControL Contl"'oL ControL APSRs
.N......
TM! ~tt I Update-12 3/94 p.3.FIG-1 Control Rod Locatton and Group O** tgnatton. For TMI-I.Curr.nt Cycl*.
CAD FILE:SIA,SKM.OO,0343,OOO-,0001 Ftg.3.2-1 Rev. 21, 04/12 61
x A
B C
D E
F G
W-_H__
K L
M N
o p
R 111213141516H81911~111~1~14 15
-y RODS IN I.
GROUPS 5-7.
Or. WD 2.
GROUP 8 AT HFP NOMINAL POSITION.
HZP WORTH OF EJECTED ROD (" A k/k)
SEE TABLE 3.2-4
.N......
TMI Untt I Upaote-12 3/94 p.3.FIG-2 E,ected Rod Location BOC and ECC Co4 File SIA.SIOI.OO.DIR.GOO-.OOOI F1,.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
r
- 2.0 r
- 1.0 rfa 1200 I 1&
I I
10 loa l..rr. -liiO IlOIUS S I SLOPf (II P)
SLOPE (EIPEIIIENT)
Ie.)
(C~ 1-)
(I 10,2)
(110.2) 25
.7143 I 00 O. 11
~
01
.25 50
.3921 o 83 o 11
~
0&
I 0 19&1 o 71 0.&1
~
.05 2.0
.0910 o &1 0.13
~
05
__ liP
- - - HEllSTllNO IlO.IEI'.
HOINEI 02
.IM
.0&
.01
.10
.12
.14
.16
.11
.20
~--
0....
~--
0....
~Nucl..r TMI Unit 1 Update* 5 7/86
- p. 3.FIG-4 Fractions of Change in the Reasonance Integral as Function of Vf:v'BJ for UO 2 Rod (T in Degrees K)
Fig. 3.2-4
0.00
-0.02
-0.04
-0.0&
-0.08
~ -0.10 0
~u
~~
..0
~~
- JH
~i 20 C) tjI-0
\\
\\
\\
\\\\
\\\\.....-V-2.4 % 6k/k(wtth Stuck Rod}
\\\\\\ ~......_-----
5.40/. 6 kl1l.J
-~-- -~- --
I o
2 e
Ttme.sec 6
7 8
p.3.FIG-6 I£a]!J Nuclear TMI Unit 1 Percent Neutron Power Versus Time Following Trip, BOC Update* 9 7/90 Fig. 3.2-6
TMI-1 UFSAR Core Loading Diagram for TMI-1 Cycle 21 1
2 3
4 5
6 7
8 9
10 11 12 13 14 15 A
21C3 G11 21F2 D05 21E2 H02 21F2 D11 21C3 G05 B
22B O11 21D P12 23F F
23G F
22E O04 23G F
23F F
21D P04 22B O05 C
21F2 N04 23F F
23H F
23E F
22A2 O10 23B F
22A2 O06 23E F
23H F
23F F
21F2 D04 D
22B M13 23F F
22E O12 22D K02 22B F07 23B F
22E L14 23B F
22B F09 22D K14 22E N03 23F F
22B M03 E
21D N14 23H F
22D B09 22B K08 23A F
22C K04 22E B06 22C K12 23A F
22B H07 22D B07 23H F
21D N02 F
21C3 M07 23F F
23E F
22B G06 23A F
22C C08 23C F
22A2 E06 23C F
22C H13 23A F
22B G10 23E F
23F F
21C3 M09 G
21F2 E04 23G F
22A2 L13 23B F
22C D09 23C F
21C1 F11 19 23D F
21C1 M10 19 23C F
22C D07 23B F
22A2 L03 23G F
21F2 E12 H
21E2 P08 22E N13 23B F
22E B10 22E L02 22A2 L05 23D F
20A2 H15 19 23D F
22A2 F11 22E F14 22E P06 23B F
22E D03 21E2 B08 K
21F2 M04 23G F
22A2 F13 23B F
22C N09 23C F
21C1 E06 19 23D F
21C1 L05 19 23C F
22C N07 23B F
22A2 F03 23G F
21F2 M12 L
21C3 E07 23F F
23E F
22B K06 23A F
22C H03 23C F
22A2 M10 23C F
22C O08 23A F
22B K10 23E F
23F F
21C3 E09 M
21D D14 23H F
22D P09 22B H09 23A F
22C G04 22E P10 22C G12 23A F
22B G08 22D P07 23H F
21D D02 N
22B E13 23F F
22E D13 22D G02 22B L07 23B F
22E F02 23B F
22B L09 22D G14 22E C04 23F F
22B E03 O
21F2 N12 23F F
23H F
23E F
22A2 C10 23B F
22A2 C06 23E F
23H F
23F F
21F2 D12 P
22B C11 21D B12 23F F
23G F
22E C12 23G F
23F F
21D B04 22B C05 R
21C3 K11 21F2 N05 21E2 H14 21F2 N11 21C3 K05 Key xxx yyy zzz Batch ID Previous Cycle Location Previous Cycle Number (N/A for Feed, Cycle 20 if blank)
Note: F denotes Fresh Fuel
- p. 3.FIG-7 ECR TM 14-00484, Rev. 0 Page 21 of 30 TMI-1 UPDATED FINAL SAETY ANALYSIS REPORT Core Loading Diagram TMI-1 Current Cycle FIGURE 3.2-7 Rev 23, 04/16
TMI UFSAR Figure 3.2-8 Deleted
- 2. 7 4:1
~
2.8 2.5 2.4
~**:I 2.3 0.-..,
~
GI:
CD ZCI 2.2
- 2. 1 2.0 1.9 1.8 10 90 110 130 150 Distance fro. bottaM ot active lenlth.
in.
0:imNuclear 1MI Unit 1 Update* 6 7/87 p.3.FIG-9 Typical DNB Ratios (BAW-2) in the Hot Unit Cell (Cycle 5)
Fig. 3.2-9
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~
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~
~"-"-
"""'Ill",
~'
1,"
1..1 1.21
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u*....
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e: *
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u
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D.II
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4 I
D.Io 0.40 0.21 D.*
a 10 20 30 40 5.
10 70 10 10 1DO
'.rclnta***f fu.1 I'ds with Hilhlr
'.akinl factlrs Thin 'lint ValUls. I 0:iElNuclear TMI Unit 1 Update* 5 7/86
- p. 3.FIG-IO Distribution of Fuel Rod Peaking (Initial Cycle)
Fig. 3.2*10
Alial Peaks 2.0 1.7 1..8
- G 1.4
-;c I.2
~
1 I..0 J 0..8
- J
~
0..6 0..4 0..2 0.0 20 40 60 80 100 120 140 Active fuei lenath. Inches Bil!lNuclear 1MI Unit 1 Update* 5 7/86
- p. 3.FIG-l1 Maximum Allowable Axial Power Distributions for a Radial-Local Peaking Factor of 1.78 Fig. 3.2-11
A.I.' 'I.ILS "I
1.4
~~~--+-----TlI.J 2.'
1.1 e.*
1.4
~
1.2 1.1
-I ***
J I.'
- I
~
'.4
'.2
'.0 0
20 41 80 10 101 120 148
'ct*** fUll Llnat** IlehlS BimNuciear 1MI Unit 1 Update* 5 7/16
- p. 3.FIG-12 Equivalent Axial Power Distribution for a Radical Local Pelklng Factor 01 1.65 (Initial Cycle)
Fig. 3.2-12
2.0
,,-...... 1.8 0-nJ
~
~
uc CD 1.6 z
Q 99% Confidence Basi s 0-
'0 aJ U
'0
- 1. 4 aJ...
a.
aJ I
c c::
~
oJ:
u
- 1. 2
~
en aJ
~
~
0 c::-
- 1. 0 0
~
~ Design Overpower nJ ex CD (114")
z Q
0.8 100 110 120 130 140 150 REFERENCE DESIGI IIOWER (2568.1t ). %
- p. 3.FIG-13 BimNuclear 1MI Unit 1 DNB Ratios (W-3) in Hot Unit Cell Versus Reactor Power (Initial Cycle)
Update* 5 7/86 Fig. 3.2-13
20 II 16 14 12 10
~..... B
~
CI 8
4 2
0
-2
-4
/
V 2120 pS11 ~
LV I
V/
I
/
A21.5 pSle I V/
/ /
v /
I""
/
vV'
//
QUill ty V
I SuDcooled r-oOSII" Ove rpower I
15 17
'"c:..
90 93 c::L
- a CI Q
N
~6
.c:
U N
.c: -
99
.c:...
GI u
c
- a.-
103
... -t-.-
0
~-
~...
109 c
U C
.c:
U c:
116 o
-t
- o e:t 127 -
.c:...
GI C..
144
-t 100 110 120 130 140 150 160 REFEREIKE DESIGII flOWER (2561.1 t).,
Bi!JNuclear TMI Unit 1 Maximum Hot Channel Exit Quality Versus Reactor Power (Initial Cycle)
- p. 3.FIG-14 Update* 5 7/86 Fig. 3.2-14
an N
C N
g Cat
~
I ca. ca.
Ba
- sa CD A.~
CoD C
an ca N
~
~
A.
Z
\\:J M
C CIt
....a uz
~
lAo
~
a..
DNIR In Hot Channel BimNuclea, TMI Unit 1 Update* 5 7/86 p.3.FIG-15 Hot Channel DNB Ratio (W-3) Versus Power for Partial Pump Operation (Initial Cycle)
Fig. 3.2-15
~
&It CL E
~
~'
A-S CL aa
~
N I
CL 0
E eft
~
~
I A-N CD 1ft N.
W-i
~..
5
~
~
wa w
0 u*
w*w"-
w*
o.
2 o.
o.
2 o.'"
o o
Coolant Quality At Point of liniMU. DNIR In Hot Channel Bil!JNuclear 1MI Unit 1 Update* 5 7/86 p.3.FIG-16 Hot Channel Quality at Point 01 Minimum ONBR Versus Power lor Partial Pump Opel'ltion (Initial Cycle)
Fig. 3.2-16
..*"...---....---.-----...--....._,....----.....~
"OJ II tts I
I altl 'ISld On laD.
I£A'-4124
( I k de
- II -/C8 )
I I
~*a...
N
... 0
-=>>....
M c:
Q..U-..
=-'".-
I 2** t-----+-........---+-~---+-----+-----+;-...
u 3.**t---_-......---~t__---_+_---__I----_....._-_I
..-I w-I
~
~
~
Ta..*
50**
.00.
- l***
2.1
- 1.....
......._...1
~Nucl..r 1MI Unit 1 Update* 5 7/86 Thermal Conductivity of U02
- p. 3.FIG-17 Fig. 3.2-17
1.8 1.6 Gaussian Distribution 1.4 1.2 1.0 70
&0 lit SO
~c:-
0 Go-40 0
u.a 30 Ii
~z 20 10 0
0.6 0.8
~Nucl..r TMI Unit 1 Update* 5 7/86 Number of Data Points vs. <t>E/<t>C p.3.FIG-18 Fig. 3.2-18
1.6 1.5 1.4 1.3 Finite Sample 90~ Confidence 1.2 Finite Sample 99% Confidence 1.1 Infinite Sample 100S Confidence 50....
~
..L-
..L..
..J 1.0 60 90 100 DNS Ratio (SAI-2) 0]2]Nuclear TMI Unit 1 Update* 5 7/86
- p. 3.FIG-19 DNB Ratio (BAW-2) vs. Population for Various Confidence Levels Fig. 3.2-19
L 025 l.020 I.015 I.010 1.005 Q-U ftI w.
I. 000 60 70 80 90 100 GJ C
C l'Q 0.995
.s=
Co)-
Qz 0.990 0.985 FA (Interior Alllmbly Channel) 0.980 o 975 0.970 FA (Wall A_mbly Channln 0.965 0.960 Population Protected.,
- p. 3.FIG-20 BiI!lNuclear 1MI Unit 1 Hot Channel Factors vs. Percent Population Protected Update* 5 7/86 Fig. 3.2-20
~-
16 14 12 10 8
6
-f?
<J 4
CQ 2
~
c 0
- 2
- 4
- 6
- 8
- i th 5'Yo Flow Factory /
Dis t r i bu t i on
/
r., /
/ " V
/
v<-:.
No Flo.
~
uV V
DIs t r I bu t I on Factor
/
I V
V I
I
/
/
QuaJlt~
/
/~
~
SuDcooled
/,
/
/'
7'
~
DeSign O~erpo.er I
100 11 0 120 130 140 150 p.3.FIG-21 Rated Power (2568 MWt>. \\>
B:il!lNuclear TMI Unit 1 Hot Channel and Nominal Channel Exit Qualities Versus Reactor Power (Without Engineering Hot Channel Factors)
(Initial Cycle)
Update* 5 7/86 Fig: 3.2-21
- Bundle Burnout rest Conditions Where Stable Operations Were Observed
~ MaxiMuM Design Conditions, 11~1o Power
- MaxiMum Design Conditions, 13010 Power tt Most Pr.obable Conditions, 11~1o Power Most Probable Co~itions, 130~ Power
\\
3.0 1o-----4-----+-----+------ir---+---+--------I o
5 10 15 20 Qual i t1 (I It
".,or/t~t.1 lit),
~
25 30 Ia:ENuclear 1MI Unit 1 Update* 5 7/86 Flow Regime Map for the Hot Unit Cell p.3.FIG-22 Fig. 3.2-22
3.0 2.5 w
I 0
)(
N.,
2.0
~
I
~
.I:-
4 u
0..
1.5
>--..s 1.0
.5
+
Bundle Burnout Test Conditions 'here Stable Operations lere Observed
- MUlmum Deslin Conditions. 114' Power Maximum Deslin Conditions, 130' Power \\
- Most Probable Conditions, 114' Power Most Probable Conditions. 130' Power f'+
+
+
to
.+ +~...,..
- +
ft*
\\
+..
... ~
+
r+
t t
't'+
~
.t
.ubbl. To
~.....
~
~ +
~
Annular
(**k.r)
++
+
... !'t+....
+
++
++
.+ *....4A +,
+
.... +
~~
to
)
'u~~I. To r~
SI ug (.'k.r)
'9' L/
~
-5 o
5 10 15 20 Quality (Ib vapor/total Itt).
~
Ia:il!lNuclear 1MI Unit 1 25 Update* 5 7/86 p.3.FIG-23 Flow Regime Map for the Hot Control Rod Cell Fig. 3.2-23
3.0 2.5 I9 III N..
2.0 I
~
.I:
~
u0..
1.5
.~
s 1.0
.5 Bundle Burnout Test Conditions Ihere Stable Operations lere Observed.
MUIIIlum Deslin Conditions.
114\\ Power MUlmum Deslin Conditions.
130\\ Power
\\
4t Most Probable Conditions.
114" Power Most Probable Conditions.
13Q~ Paler
\\
~
Ie
.u....,. To Annu'ar
(.ak.r)
~
.-..~
.u....,. To j
Slug (.ak.r)
/
~
-5 o
5 10 15 20 Qua'it, ('" vapor/total Ib),
~
Bi!JNuclear TMI Unit 1 25 30 Update* 5 7/86 p.3.FIG-24 Flow Regime Map lor the Hot Wall Cell Fig. 3.2-24
3.0 2.5
.0*2
)(
2.0 N...
'P*
~
.I:.-*
u 1.5 0
U
~
........z 1.0
.5
~undle Burnout Test ConditIons Where Stabl! Operations Were Observed
- Maximum Des1in Conditions.
114\\ Power
- Maximum Deslin Conditions, 130'\\ Power \\
Most PrObable Conditions, 114\\ Power Most Probable Conditions, 130~~ Power
\\
- .~. * -.
- 4 Bubble To Annular
( Baker)
~
- . ~.
--~
'ubbl. To
\\
J Slul (.....r)
~l/
-5 o
5 10 15 20 Qual i ty (I b vapor/ total I b).
IJ.
Bil!INuclear 1MI Unit 1 25 30 Update* 5 7/86 p.3.FIG-25 Flow Regime Map for the Hot Corner Cell Fig. 3.2-25
150
/
DNBR (1-3).1.30 I
I 90 140 Desia" Flow Ratl
~
J.. (131.32 I
1D6 Ib/hr)
I
~
130
-- -I--
-~ -,t.
~..
I/~
120 Ii I~
Dlsia" Overpower
~
110
~ (1141 I 2568 **t)
~
100
/
I I
I 2400 2600 2800 3000 3200 3400 3600 Reactor Core Po.er,.It BENue.ear 1MI Unit 1 Update* 5 7/86 p.3.FIG-26 Reactor Coolant System Flow Versus Power (Initial Cycle)
Fig. 3.2-26
2.4 2.2 LINE flOI MIXING COEff.
1 1101
.02 cJ 2
1001
.02 u
2.0 3
901
.02 II 100S
.0&
Il 5
1001
.01 Il C"")
I*
1.8 Q
~
ftI ac CD
- z Q
1.6 cu c
C ftI
.c
~
~
1.4 0:c 1.30 (1-3)-
1.2 1.0 o I 100 110 120 130 140 150 REFERENCE DESIGN IIOWER (2511 Btl. I BENuclear 1MI Unit 1 Update* 5 7/86 p.3.FIG-27 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
~.-
~a.
E lU.-'-
3600
~.-c
~
u
~
~
u..
3200 2800 2400 O.0095"Clearance uo
~oeSIRn r
Overpower (114',
~ 100' Power
'-- Maximum Design Clearance Nominal Clearance 6
8 10 12 14 16 18 20 22 24 26 28 30 linear Heat Rate, kw. ft BiI!INuclear 1MI Unit 1 Update* 5 7/86 p.3.FIG-28 Fuel Center Temperature for Beginning-of-Cycle Conditions (Initial Cycle)
Fig. 3.2-28
5200 3600 cu....
c:
cu c....)
3200 2800
.0095" Clearance
~ Design Overpower (1141) 100~ Power
"- Maximum Design Clearance Nominal Clearance 6
8 10 12 16 18 20 22 28 30 linear Heat Rate, kw/ft
~Nuclear 1MI Unit 1 Update* 5 7/86 p.3.FIG-29 Fuel Center Temperature lor End-ai-Cycle Conditions (Initial Cycle)
Fig. 3.2-29
5200 1ft--
a-N 4800 4400 4000*
BOC (100 IIO/ITU) 3&00
~-....
3200 Q,*..
to-
~...
2800
~
2400 2000 1608 1200 IZ!imNuclear TMI Unit 1 Update* 6 7/87 p.3.FIG-30 Typical Post-Initial Cycle - Center Line Fuel Temperature vs. Linear Heat Rate Hot Pin (Cycle 5)
Fig. 3.2-30
5000
""" 4800 u...
~-
lW...u i
4600
...u...
c::
u U
u
~
""" 4400 4200 o 8
16 11.63 kwlt tHat Spot (1001 Power) 24 32 40 5000 u...
~-
lW...
4000 I....
U-c::u u
3000 Burnup (l.llrU I 10 -3)
EOl 110 40.900 iTii 2000 o
12 16 20 24
- p. 3.FIG-31 llnllr Hilt Rltl, kw/tt
~Nuclear TMI Unit 1 Burnup Effect on Fuel Center Temperature (Initial Cycle)
Update* 5 7/86 Fig. 3.2-31
3600 3200 2800 lL 2400
&)
~
- )-
~
&)
a.
E 2000
&)
~
&)
- I 1600 1200 800
\\
\\
\\
~....
~'"
~
o 20 40 60 80 100 YOIU.' Fraction of Totll Fuel.,
(at or IDove Fuel T.perature)
BENuclear 1MI Unit 1 Update* 5 7/86
- p. 3.FIG-32 Fuel Temperature Versus Total Fuel Volume Fraction for Equilibrium Cycle at End-ol-Cycle Fig. 3.2-32
BimNuclea, TMI Unit 1 l+--"-
0.119 NU.'lr erells Iurn.d Update* 5 7/86 p.3.FIG-33 Typical Reactor Fuel Assembly Power Distribution at End-ol-Cycle Equilibrium Cycle Conditions lor 1/8 Core Fig. 3.2-33
10kwi t t FueI 1200....--------+----------+--~----+---
2400 2800 1----------~-------_+_--_;__-~~-_____4 3200
.... 2000 0
lU...
~
6 kwIt t to...
lU
/---
~
E 1600 lU to-0.24 0.20 0.16 0.12 0.08 0.04 S80 0
F_TIVIl COOl:::!
800 400 "-
...A..-
~___""___6__""'___.....
0.0 fuel Radius, in BiI!lNuclear TMI Unit 1 Update* 5 7/86 Fuel Rod Temperature Profiles at 6 and 10 KW/Ft p.3.FIG-34 Fig. 3.2-34
§...
I...
I...
I
~
I u-
- I.....-
§..a.
I!
N
~
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t'\\
--... j'"---.......
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~
4 hp 8
8 8
8 8
o o-o V'o o
fi ** ten c** 1111.'ld, I
~Nuclear TMI Unit 1 Update* 5 7/86 p.3.FIG-35 Percent Fission Gas Released as a Function of the Average Temperature of the U02 Fuel Fig. 3.2-35
d-of-Life lUll;
- 3. DI,I-IIVIU 1.10 (partial Rod Insertion)
I I
---~ FUll lidplln.
Cor. I.ttll I
PIP - 1. 50 (Iodifi.d COlin.)
- 1. 4'---~---+--~~-r-.....-~.........
1.1--....-
....-~..............
I
....-._--t--I--/-...'~,£.. PIp*
1.1
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/
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1-I
- o. 0....~_~_..........,j~....._..A..-_........._....._....-._~
0.' ~..,...-......---...+---t---+--+-----t~.....-
......~-+...............
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- a. 4....--.....----....t~--+_-.....--I---+----+~-~....__t O. 2....--......--~~--+_-.....--t---.....---+--~...--tI
'L Iii
~
~
D.' ~~~-+-....._~..---+--...........--.......+---......~+-......t-t.......
20 40
&0 10 110 120 14a Dlltlnc. trGl lott.. of Acti,. FUll. in.
Iml!]Nuclear 1MI Unit 1 Update* 5 7/86 Axial Local to Average Burnup and Instantaneous Power Comparisons
- p. 3.FIG-36 Fig. 3.2-36
Maximum DeSii" Diametral Gap "0
800
~
u 0-600 lU
- )
400 200 0
0 8
12 16 20 24 linear Heat Rate,.
kw/ f t 2400 2200 2000
~
0 1800 I
~-
I... J600
- l-CD Noml na I lU 1400 DI a.e t raI Gap u="'
T" 1200 u
- )
l:'c:
0 u
1000 c.
ftI e.,:,
0:imNuclear 1MI Unit 1 Update* 6 7/87 Fuel to Clad Gap Conductance 'or End-O'-Cycle Conditions (Initial Cycle)
- p. 3.FIG-37 Fig. 3.2-37
2D I
I I
~
I I
lalimull DesiRR Clearance 1.1 Alial Power and EDl Burnup Snape with Closed Pores 1.7 Ilill Power Ind EDl Bu rnup Shape
- ith Op.en ~
'ores 1.5 Ilial POler Ind EDL Burnup Snape with Closed PQres 5
15
~
10 1:1 U
~..u o
o 2
4 6
I Initial Cold Diametral Clearance. in I
103 10 BimNuclear TMI Unit 1 Update* 6 7/87
- p. 3.FIG-38 Fission Gas Release for 1.5 and 1.7 MaxlAvg. Axial Power Shapes (Initial Cycle)
Fig. 3.2-38
3500 3000 Desi an l illi t 114" Overpower 100~ Power 2500
&Ita.
c Closed u
m 2000
=.-
1.7 AI I aI Po.er an uc EDt Bu rnup Snape
&Itc 1.5 Axial Power ana u
EDl Burnup Snape
~
- )
&It Pores
&It 1500 u
~
1.7 Ali al Power and Q,.
&It EDl Burnup Snape CoD 1000 MIII.U. Desie" Cleara"Ce~
500 o
2 4
6 8
10 Initial Cold Diaetral Clearance.
in I 103 BiE]Nuclear TMI Unit 1 Update* 6 7/87 p.3.FIG-39 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
, I Ik In. Fie tor CD HO' UN I' Cfll (ntllllp, IIISI Flctor CD "OJ UU CELL CD HO' CO'NlI CELL
HOT COhUOl 100 ULL Bil!INuclear TMI Unit 1 Update* 6 7/87
- p. 3.FIG-40 Nominal Fuel Rod Power Peaks and Cell Exit Enthalphy Rise Ratios (Initial Cycle)
Fig. 3.2-40
~--
t I
hclllr "1 flCtor Q)
NOT UNIT CEll (nUIII" ** 11 'acur
~
MOT tAll CEll OJ MOT CUNEa CEll
~
MOT CONTaOl 100 ell l BENuelear TMI Unit 1 Update - 6 7/87
- p. 3.FJG-41 Maximum Fuel Rod Power Peaks and Cell Exit Enthalphy Rise Ratios (Initial Cycle)
Fig; 3.2-41
1.1 1.5 1.4 1.3 1.2 1.1 CD 1.0 I
CI-..
N 0.9 I...
- a 0.1 0.7 IC
- a-
~...
0.&
z-..
0.5 u
0
~
0.4 0.3 0.2 011
\\
G* 2.21 I 101 IIi/tH-f t2
\\.'\\
I\\:
1-3 DNa Milt Flu.
!\\.
(DISII" Lillit)
'\\
~'\\
r\\.
lin IIIUII ON** 1.55
'\\
\\
\\
~
\\
~
/
V
~~
\\
~ I
/
1\\
V V
Calculatld Surflc.
\\
Milt Flu.
~
J'
\\I o
540 5&0 510 600 620 140 110
&10 100 120 LOCII Enthalpy, Itu/III BimNuclear 1MI Unit 1 Update* 5 7/86 p.3.FIG-42 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
1.1 1.5 1.4 1.3 1.2 1.1 UI 1..0 I
0-
)(
0.9 N...-I 0.8 s::"-
~...
CD O. 7
- I-..-
0.6 O.~
lQ U
Q
~
0.4 0.3 0.2 O. 1
~\\
\\
,.2.5' I 10' Ib/tH-f t2
~
I I
I I
~
1 r
I 1
~
1-3 ONI H**t Flu.
(D.Slln l'.' t)
~
\\
1\\
\\
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\\
\\
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~
V V
~
I
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Calculated Surface
~
J J
....t Flul
\\,
o
~40 560 510 600 620
&40 6&0 680 700 720 local Enthalpy, Itu/lb 0ENuclear TMI Unit 1 Update* 5 7/86 p.3.FIG-43 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 ROO ASSEMBLY PL ENUM ASSEMBLY OUTLET NOZ ZLE
~;URvE ILLANCE 5PECI~I1EN HOLDE R TUBE
~OD GUIDE TUBE CORE SUPPORT SHIELD INLET NOZZLE FuEL ASSEMBLY REACTOR VESSEL THERMAL SHIELD GUIDE LUGS INCORE INSTRUMENT GUIDE TUBES
________ IN CO REI NS TRUM EN T N 0 ZZ L E S Bil!1NucIe.r TMI Unit 1 Update*5 7186 p.3.FIG-44 Reactor VI_lind Intlmlll
- Gen. Arnngement Fig. 3.2-44 TMI Unit 1 Updated Final Safety Analysis Report Reactor Vessel and Internals - Gen. Arrangement Figure 3.2-44 Rev. 20, 04/10
CONTROL ROD ASSEMBLY LDCATION INCORE INSTRUMENT LOCATION REACTOR VESSEL THERMAL SHIELD CORE BARREL SURVEILLANCE SPECIMEN HOLDER TUIE BimNuclea, TMI Unit 1 Update* 5 7/86
- p. 3.FIG-45 Reactor Vessel and Internals - Cross Section Fig. 3.2-45
0imNuclear 1MI Unit 1 Core Flooding Arrangement p.3.FIG-46 COlE fL ODD IN&
NOZZLE Update* 5 7/86 Fig. 3.2-46
1-,
--+----+++---tr---- t ---~--+;---+--
rn I
eMf I....'
MilL.
SICT** I-I Bi!lNucl.r 1MI Unit 1 UI'C..I lie
.u".
'URl
.IU Update* 5 7/86
- p. 3.FIG-47 Internals Vent Valves Fig. 3.2-47
BimNuclear 1MI Unit 1 SEE SECTION BELOW Update* 5 7/86 p.3.FIG-48 Internals Vent Valve Clearance Gaps Fig. 3.2-48
TOP VIEW Update* 5 7186
~NucIe8r TMI Unit 1 FUll AlUmbly COHTROl. ROO GUIDE TUH
,'NSTRUMENTATION I
TUM I
r i
i J
, ',',r.
I i
i
' \\,
9-~
FUEL ROO ASSEMILY i
I r::
1:7' Ir1I n
'n I
r I I
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1 1 CROSS SECTION I
1
, i I
1
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i ] 1 *
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- r-'---I'ir '~
I INSTRUMNTATION TUH.-l CONNECTION SNeER GRID U"O ENO trlTTING
)
)
Fig. 3.2-49 p.3.FIG-49
SPIDER TOP VIIW COUPLINt NEUTRON ~IING MATERIAL CONTROl. ROD p.3.FIG-50 ra::imNuclea, TMI Unit 1 Control Rod Assembly Update* 5 7/86 Fig. 3.2-50
SPIDER TOP VI!W NEUTRON ABSORItNG MATERIAL~
~Nucl..r 1MI Unit 1 Axial Power Shaping Rod Assembly Update* 5 7/86 p.3.FIG-51 Fig. 3.2-51
SPIOfR ----
IURNAILI POISON ROD TOP VIIW
.URNA.LI POIION MATIRIAL
- p. 3.FIG-52 0iI!JNuclear 1MI Unit 1 Burnable Poison Rod Assembly Update* 5 7/86 Fig. 3.2-52
TOP VIEW SPIDER-----....
ORIFICE ROO
- p. 3.FIG-'53 Bil!lNuclear TMI Unit 1 Orifice Rod Assembly Update* 5 7/86 Fig. 3.2-53
tOJSING ASSY LOAD ARM ASSDfSl.Y p.3.FIG-54 0:W!lNuclear TMI Unit 1 Side View of BPRA Retainer Update* 5 7/86 Fig. 3.2-54
UPPER CORE PLAT! ASSY PAD TYP p.3.FIG-55
~Nucl..r TMI Unit 1 Top View of BPRA Retainer During Operation Update* 5 7/86 Fig. 3.2-55
POsmoN INDICATQIIt AlIDIILY STATOR.IIIMLV
- p. 3.FIG-56 I
I CCUIL.NI AlIDIILV i i BimNuclear TMI Unit 1 Control Rod Drive - Gen. Arrangement Update* 5 7/86 Fig. 3.2-56
\\)
SECII** Z*Z SECII** I-I SIC"'I,.,
/E
//
.c:...- WlI' VALl( ASSEI.U
/IIT.I "Il lUI SUI' II' USEI.U L. l,l, Slell.. 1*1 SECII** 1*1 lUI SCll' **.*l **S*.I.
-r~j L
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r~
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l
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L~."-
L S
SEC"" 1111 SEC 110.,.,
SEC""
S-S Bil!INucI**r TMI Unit 1 UpdItI. 5 7111 p.3.FIG-57 Fig. 3.2*57 TMI-UNIT 1 REV.
19, APRI L 2008 p.3.FIG-3 Stuck Rod Location BOC and EOC FIGURE 3.2-3
x A
B C
0 E
F G
w-H K
L M
N 0
P R
1 6
1 3
5 5
3 7
8 7
8 7
3 5
4 4
5 3
1 8
6 2
6 8
I 5
4 2
2 4
5 6
7 2
7 2
7 6
5 4
2 2
4 5
1 8
6 2
6 8
I 3
5 4
4 5
3 7
8 7
8 7
3 5
5 3
I 6
I I
2 3
4 5
6 7
8 9
10 I I 12 13 14 15
-y No.of I"'ods FuncTloo Z
WORTH-HZP
(/. ~ k/k)
Group Number Gr'OUP I
8 2
8 3
8 4
8 5
12 6
8 7 (TRANSIENT) 9 8
8 TOTAL # 69 Sofety Sofety Sofety Sofety ControL Contl"'oL ControL APSRs
.N......
TM! ~tt I Update-12 3/94 p.3.FIG-1 Control Rod Locatton and Group O** tgnatton. For TMI-I.Curr.nt Cycl*.
CAD FILE:SIA,SKM.OO,0343,OOO-,0001 Ftg.3.2-1 Rev. 21, 04/12 61
x A
B C
D E
F G
W-_H__
K L
M N
o p
R 111213141516H81911~111~1~14 15
-y RODS IN I.
GROUPS 5-7.
Or. WD 2.
GROUP 8 AT HFP NOMINAL POSITION.
HZP WORTH OF EJECTED ROD (" A k/k)
SEE TABLE 3.2-4
.N......
TMI Untt I Upaote-12 3/94 p.3.FIG-2 E,ected Rod Location BOC and ECC Co4 File SIA.SIOI.OO.DIR.GOO-.OOOI F1,.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
r
- 2.0 r
- 1.0 rfa 1200 I 1&
I I
10 loa l..rr. -liiO IlOIUS S I SLOPf (II P)
SLOPE (EIPEIIIENT)
Ie.)
(C~ 1-)
(I 10,2)
(110.2) 25
.7143 I 00 O. 11
~
01
.25 50
.3921 o 83 o 11
~
0&
I 0 19&1 o 71 0.&1
~
.05 2.0
.0910 o &1 0.13
~
05
__ liP
- - - HEllSTllNO IlO.IEI'.
HOINEI 02
.IM
.0&
.01
.10
.12
.14
.16
.11
.20
~--
0....
~--
0....
~Nucl..r TMI Unit 1 Update* 5 7/86
- p. 3.FIG-4 Fractions of Change in the Reasonance Integral as Function of Vf:v'BJ for UO 2 Rod (T in Degrees K)
Fig. 3.2-4
0.00
-0.02
-0.04
-0.0&
-0.08
~ -0.10 0
~u
~~
..0
~~
- JH
~i 20 C) tjI-0
\\
\\
\\
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\\\\\\ ~......_-----
5.40/. 6 kl1l.J
-~-- -~- --
I o
2 e
Ttme.sec 6
7 8
p.3.FIG-6 I£a]!J Nuclear TMI Unit 1 Percent Neutron Power Versus Time Following Trip, BOC Update* 9 7/90 Fig. 3.2-6
TMI-1 UFSAR Core Loading Diagram for TMI-1 Cycle 21 1
2 3
4 5
6 7
8 9
10 11 12 13 14 15 A
21C3 G11 21F2 D05 21E2 H02 21F2 D11 21C3 G05 B
22B O11 21D P12 23F F
23G F
22E O04 23G F
23F F
21D P04 22B O05 C
21F2 N04 23F F
23H F
23E F
22A2 O10 23B F
22A2 O06 23E F
23H F
23F F
21F2 D04 D
22B M13 23F F
22E O12 22D K02 22B F07 23B F
22E L14 23B F
22B F09 22D K14 22E N03 23F F
22B M03 E
21D N14 23H F
22D B09 22B K08 23A F
22C K04 22E B06 22C K12 23A F
22B H07 22D B07 23H F
21D N02 F
21C3 M07 23F F
23E F
22B G06 23A F
22C C08 23C F
22A2 E06 23C F
22C H13 23A F
22B G10 23E F
23F F
21C3 M09 G
21F2 E04 23G F
22A2 L13 23B F
22C D09 23C F
21C1 F11 19 23D F
21C1 M10 19 23C F
22C D07 23B F
22A2 L03 23G F
21F2 E12 H
21E2 P08 22E N13 23B F
22E B10 22E L02 22A2 L05 23D F
20A2 H15 19 23D F
22A2 F11 22E F14 22E P06 23B F
22E D03 21E2 B08 K
21F2 M04 23G F
22A2 F13 23B F
22C N09 23C F
21C1 E06 19 23D F
21C1 L05 19 23C F
22C N07 23B F
22A2 F03 23G F
21F2 M12 L
21C3 E07 23F F
23E F
22B K06 23A F
22C H03 23C F
22A2 M10 23C F
22C O08 23A F
22B K10 23E F
23F F
21C3 E09 M
21D D14 23H F
22D P09 22B H09 23A F
22C G04 22E P10 22C G12 23A F
22B G08 22D P07 23H F
21D D02 N
22B E13 23F F
22E D13 22D G02 22B L07 23B F
22E F02 23B F
22B L09 22D G14 22E C04 23F F
22B E03 O
21F2 N12 23F F
23H F
23E F
22A2 C10 23B F
22A2 C06 23E F
23H F
23F F
21F2 D12 P
22B C11 21D B12 23F F
23G F
22E C12 23G F
23F F
21D B04 22B C05 R
21C3 K11 21F2 N05 21E2 H14 21F2 N11 21C3 K05 Key xxx yyy zzz Batch ID Previous Cycle Location Previous Cycle Number (N/A for Feed, Cycle 20 if blank)
Note: F denotes Fresh Fuel
- p. 3.FIG-7 ECR TM 14-00484, Rev. 0 Page 21 of 30 TMI-1 UPDATED FINAL SAETY ANALYSIS REPORT Core Loading Diagram TMI-1 Current Cycle FIGURE 3.2-7 Rev 23, 04/16
TMI UFSAR Figure 3.2-8 Deleted
- 2. 7 4:1
~
2.8 2.5 2.4
~**:I 2.3 0.-..,
~
GI:
CD ZCI 2.2
- 2. 1 2.0 1.9 1.8 10 90 110 130 150 Distance fro. bottaM ot active lenlth.
in.
0:imNuclear 1MI Unit 1 Update* 6 7/87 p.3.FIG-9 Typical DNB Ratios (BAW-2) in the Hot Unit Cell (Cycle 5)
Fig. 3.2-9
~~
""'lIIIIIiI~ "-."
~
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~
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4 I
D.Io 0.40 0.21 D.*
a 10 20 30 40 5.
10 70 10 10 1DO
'.rclnta***f fu.1 I'ds with Hilhlr
'.akinl factlrs Thin 'lint ValUls. I 0:iElNuclear TMI Unit 1 Update* 5 7/86
- p. 3.FIG-IO Distribution of Fuel Rod Peaking (Initial Cycle)
Fig. 3.2*10
Alial Peaks 2.0 1.7 1..8
- G 1.4
-;c I.2
~
1 I..0 J 0..8
- J
~
0..6 0..4 0..2 0.0 20 40 60 80 100 120 140 Active fuei lenath. Inches Bil!lNuclear 1MI Unit 1 Update* 5 7/86
- p. 3.FIG-l1 Maximum Allowable Axial Power Distributions for a Radial-Local Peaking Factor of 1.78 Fig. 3.2-11
A.I.' 'I.ILS "I
1.4
~~~--+-----TlI.J 2.'
1.1 e.*
1.4
~
1.2 1.1
-I ***
J I.'
- I
~
'.4
'.2
'.0 0
20 41 80 10 101 120 148
'ct*** fUll Llnat** IlehlS BimNuciear 1MI Unit 1 Update* 5 7/16
- p. 3.FIG-12 Equivalent Axial Power Distribution for a Radical Local Pelklng Factor 01 1.65 (Initial Cycle)
Fig. 3.2-12
2.0
,,-...... 1.8 0-nJ
~
~
uc CD 1.6 z
Q 99% Confidence Basi s 0-
'0 aJ U
'0
- 1. 4 aJ...
a.
aJ I
c c::
~
oJ:
u
- 1. 2
~
en aJ
~
~
0 c::-
- 1. 0 0
~
~ Design Overpower nJ ex CD (114")
z Q
0.8 100 110 120 130 140 150 REFERENCE DESIGI IIOWER (2568.1t ). %
- p. 3.FIG-13 BimNuclear 1MI Unit 1 DNB Ratios (W-3) in Hot Unit Cell Versus Reactor Power (Initial Cycle)
Update* 5 7/86 Fig. 3.2-13
20 II 16 14 12 10
~..... B
~
CI 8
4 2
0
-2
-4
/
V 2120 pS11 ~
LV I
V/
I
/
A21.5 pSle I V/
/ /
v /
I""
/
vV'
//
QUill ty V
I SuDcooled r-oOSII" Ove rpower I
15 17
'"c:..
90 93 c::L
- a CI Q
N
~6
.c:
U N
.c: -
99
.c:...
GI u
c
- a.-
103
... -t-.-
0
~-
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109 c
U C
.c:
U c:
116 o
-t
- o e:t 127 -
.c:...
GI C..
144
-t 100 110 120 130 140 150 160 REFEREIKE DESIGII flOWER (2561.1 t).,
Bi!JNuclear TMI Unit 1 Maximum Hot Channel Exit Quality Versus Reactor Power (Initial Cycle)
- p. 3.FIG-14 Update* 5 7/86 Fig. 3.2-14
an N
C N
g Cat
~
I ca. ca.
Ba
- sa CD A.~
CoD C
an ca N
~
~
A.
Z
\\:J M
C CIt
....a uz
~
lAo
~
a..
DNIR In Hot Channel BimNuclea, TMI Unit 1 Update* 5 7/86 p.3.FIG-15 Hot Channel DNB Ratio (W-3) Versus Power for Partial Pump Operation (Initial Cycle)
Fig. 3.2-15
~
&It CL E
~
~'
A-S CL aa
~
N I
CL 0
E eft
~
~
I A-N CD 1ft N.
W-i
~..
5
~
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wa w
0 u*
w*w"-
w*
o.
2 o.
o.
2 o.'"
o o
Coolant Quality At Point of liniMU. DNIR In Hot Channel Bil!JNuclear 1MI Unit 1 Update* 5 7/86 p.3.FIG-16 Hot Channel Quality at Point 01 Minimum ONBR Versus Power lor Partial Pump Opel'ltion (Initial Cycle)
Fig. 3.2-16
..*"...---....---.-----...--....._,....----.....~
"OJ II tts I
I altl 'ISld On laD.
I£A'-4124
( I k de
- II -/C8 )
I I
~*a...
N
... 0
-=>>....
M c:
Q..U-..
=-'".-
I 2** t-----+-........---+-~---+-----+-----+;-...
u 3.**t---_-......---~t__---_+_---__I----_....._-_I
..-I w-I
~
~
~
Ta..*
50**
.00.
- l***
2.1
- 1.....
......._...1
~Nucl..r 1MI Unit 1 Update* 5 7/86 Thermal Conductivity of U02
- p. 3.FIG-17 Fig. 3.2-17
1.8 1.6 Gaussian Distribution 1.4 1.2 1.0 70
&0 lit SO
~c:-
0 Go-40 0
u.a 30 Ii
~z 20 10 0
0.6 0.8
~Nucl..r TMI Unit 1 Update* 5 7/86 Number of Data Points vs. <t>E/<t>C p.3.FIG-18 Fig. 3.2-18
1.6 1.5 1.4 1.3 Finite Sample 90~ Confidence 1.2 Finite Sample 99% Confidence 1.1 Infinite Sample 100S Confidence 50....
~
..L-
..L..
..J 1.0 60 90 100 DNS Ratio (SAI-2) 0]2]Nuclear TMI Unit 1 Update* 5 7/86
- p. 3.FIG-19 DNB Ratio (BAW-2) vs. Population for Various Confidence Levels Fig. 3.2-19
L 025 l.020 I.015 I.010 1.005 Q-U ftI w.
I. 000 60 70 80 90 100 GJ C
C l'Q 0.995
.s=
Co)-
Qz 0.990 0.985 FA (Interior Alllmbly Channel) 0.980 o 975 0.970 FA (Wall A_mbly Channln 0.965 0.960 Population Protected.,
- p. 3.FIG-20 BiI!lNuclear 1MI Unit 1 Hot Channel Factors vs. Percent Population Protected Update* 5 7/86 Fig. 3.2-20
~-
16 14 12 10 8
6
-f?
<J 4
CQ 2
~
c 0
- 2
- 4
- 6
- 8
- i th 5'Yo Flow Factory /
Dis t r i bu t i on
/
r., /
/ " V
/
v<-:.
No Flo.
~
uV V
DIs t r I bu t I on Factor
/
I V
V I
I
/
/
QuaJlt~
/
/~
~
SuDcooled
/,
/
/'
7'
~
DeSign O~erpo.er I
100 11 0 120 130 140 150 p.3.FIG-21 Rated Power (2568 MWt>. \\>
B:il!lNuclear TMI Unit 1 Hot Channel and Nominal Channel Exit Qualities Versus Reactor Power (Without Engineering Hot Channel Factors)
(Initial Cycle)
Update* 5 7/86 Fig: 3.2-21
- Bundle Burnout rest Conditions Where Stable Operations Were Observed
~ MaxiMuM Design Conditions, 11~1o Power
- MaxiMum Design Conditions, 13010 Power tt Most Pr.obable Conditions, 11~1o Power Most Probable Co~itions, 130~ Power
\\
3.0 1o-----4-----+-----+------ir---+---+--------I o
5 10 15 20 Qual i t1 (I It
".,or/t~t.1 lit),
~
25 30 Ia:ENuclear 1MI Unit 1 Update* 5 7/86 Flow Regime Map for the Hot Unit Cell p.3.FIG-22 Fig. 3.2-22
3.0 2.5 w
I 0
)(
N.,
2.0
~
I
~
.I:-
4 u
0..
1.5
>--..s 1.0
.5
+
Bundle Burnout Test Conditions 'here Stable Operations lere Observed
- MUlmum Deslin Conditions. 114' Power Maximum Deslin Conditions, 130' Power \\
- Most Probable Conditions, 114' Power Most Probable Conditions. 130' Power f'+
+
+
to
.+ +~...,..
- +
ft*
\\
+..
... ~
+
r+
t t
't'+
~
.t
.ubbl. To
~.....
~
~ +
~
Annular
(**k.r)
++
+
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+
++
++
.+ *....4A +,
+
.... +
~~
to
)
'u~~I. To r~
SI ug (.'k.r)
'9' L/
~
-5 o
5 10 15 20 Quality (Ib vapor/total Itt).
~
Ia:il!lNuclear 1MI Unit 1 25 Update* 5 7/86 p.3.FIG-23 Flow Regime Map for the Hot Control Rod Cell Fig. 3.2-23
3.0 2.5 I9 III N..
2.0 I
~
.I:
~
u0..
1.5
.~
s 1.0
.5 Bundle Burnout Test Conditions Ihere Stable Operations lere Observed.
MUIIIlum Deslin Conditions.
114\\ Power MUlmum Deslin Conditions.
130\\ Power
\\
4t Most Probable Conditions.
114" Power Most Probable Conditions.
13Q~ Paler
\\
~
Ie
.u....,. To Annu'ar
(.ak.r)
~
.-..~
.u....,. To j
Slug (.ak.r)
/
~
-5 o
5 10 15 20 Qua'it, ('" vapor/total Ib),
~
Bi!JNuclear TMI Unit 1 25 30 Update* 5 7/86 p.3.FIG-24 Flow Regime Map lor the Hot Wall Cell Fig. 3.2-24
3.0 2.5
.0*2
)(
2.0 N...
'P*
~
.I:.-*
u 1.5 0
U
~
........z 1.0
.5
~undle Burnout Test ConditIons Where Stabl! Operations Were Observed
- Maximum Des1in Conditions.
114\\ Power
- Maximum Deslin Conditions, 130'\\ Power \\
Most PrObable Conditions, 114\\ Power Most Probable Conditions, 130~~ Power
\\
- .~. * -.
- 4 Bubble To Annular
( Baker)
~
- . ~.
--~
'ubbl. To
\\
J Slul (.....r)
~l/
-5 o
5 10 15 20 Qual i ty (I b vapor/ total I b).
IJ.
Bil!INuclear 1MI Unit 1 25 30 Update* 5 7/86 p.3.FIG-25 Flow Regime Map for the Hot Corner Cell Fig. 3.2-25
150
/
DNBR (1-3).1.30 I
I 90 140 Desia" Flow Ratl
~
J.. (131.32 I
1D6 Ib/hr)
I
~
130
-- -I--
-~ -,t.
~..
I/~
120 Ii I~
Dlsia" Overpower
~
110
~ (1141 I 2568 **t)
~
100
/
I I
I 2400 2600 2800 3000 3200 3400 3600 Reactor Core Po.er,.It BENue.ear 1MI Unit 1 Update* 5 7/86 p.3.FIG-26 Reactor Coolant System Flow Versus Power (Initial Cycle)
Fig. 3.2-26
2.4 2.2 LINE flOI MIXING COEff.
1 1101
.02 cJ 2
1001
.02 u
2.0 3
901
.02 II 100S
.0&
Il 5
1001
.01 Il C"")
I*
1.8 Q
~
ftI ac CD
- z Q
1.6 cu c
C ftI
.c
~
~
1.4 0:c 1.30 (1-3)-
1.2 1.0 o I 100 110 120 130 140 150 REFERENCE DESIGN IIOWER (2511 Btl. I BENuclear 1MI Unit 1 Update* 5 7/86 p.3.FIG-27 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
~.-
~a.
E lU.-'-
3600
~.-c
~
u
~
~
u..
3200 2800 2400 O.0095"Clearance uo
~oeSIRn r
Overpower (114',
~ 100' Power
'-- Maximum Design Clearance Nominal Clearance 6
8 10 12 14 16 18 20 22 24 26 28 30 linear Heat Rate, kw. ft BiI!INuclear 1MI Unit 1 Update* 5 7/86 p.3.FIG-28 Fuel Center Temperature for Beginning-of-Cycle Conditions (Initial Cycle)
Fig. 3.2-28
5200 3600 cu....
c:
cu c....)
3200 2800
.0095" Clearance
~ Design Overpower (1141) 100~ Power
"- Maximum Design Clearance Nominal Clearance 6
8 10 12 16 18 20 22 28 30 linear Heat Rate, kw/ft
~Nuclear 1MI Unit 1 Update* 5 7/86 p.3.FIG-29 Fuel Center Temperature lor End-ai-Cycle Conditions (Initial Cycle)
Fig. 3.2-29
5200 1ft--
a-N 4800 4400 4000*
BOC (100 IIO/ITU) 3&00
~-....
3200 Q,*..
to-
~...
2800
~
2400 2000 1608 1200 IZ!imNuclear TMI Unit 1 Update* 6 7/87 p.3.FIG-30 Typical Post-Initial Cycle - Center Line Fuel Temperature vs. Linear Heat Rate Hot Pin (Cycle 5)
Fig. 3.2-30
5000
""" 4800 u...
~-
lW...u i
4600
...u...
c::
u U
u
~
""" 4400 4200 o 8
16 11.63 kwlt tHat Spot (1001 Power) 24 32 40 5000 u...
~-
lW...
4000 I....
U-c::u u
3000 Burnup (l.llrU I 10 -3)
EOl 110 40.900 iTii 2000 o
12 16 20 24
- p. 3.FIG-31 llnllr Hilt Rltl, kw/tt
~Nuclear TMI Unit 1 Burnup Effect on Fuel Center Temperature (Initial Cycle)
Update* 5 7/86 Fig. 3.2-31
3600 3200 2800 lL 2400
&)
~
- )-
~
&)
a.
E 2000
&)
~
&)
- I 1600 1200 800
\\
\\
\\
~....
~'"
~
o 20 40 60 80 100 YOIU.' Fraction of Totll Fuel.,
(at or IDove Fuel T.perature)
BENuclear 1MI Unit 1 Update* 5 7/86
- p. 3.FIG-32 Fuel Temperature Versus Total Fuel Volume Fraction for Equilibrium Cycle at End-ol-Cycle Fig. 3.2-32
BimNuclea, TMI Unit 1 l+--"-
0.119 NU.'lr erells Iurn.d Update* 5 7/86 p.3.FIG-33 Typical Reactor Fuel Assembly Power Distribution at End-ol-Cycle Equilibrium Cycle Conditions lor 1/8 Core Fig. 3.2-33
10kwi t t FueI 1200....--------+----------+--~----+---
2400 2800 1----------~-------_+_--_;__-~~-_____4 3200
.... 2000 0
lU...
~
6 kwIt t to...
lU
/---
~
E 1600 lU to-0.24 0.20 0.16 0.12 0.08 0.04 S80 0
F_TIVIl COOl:::!
800 400 "-
...A..-
~___""___6__""'___.....
0.0 fuel Radius, in BiI!lNuclear TMI Unit 1 Update* 5 7/86 Fuel Rod Temperature Profiles at 6 and 10 KW/Ft p.3.FIG-34 Fig. 3.2-34
§...
I...
I...
I
~
I u-
- I.....-
§..a.
I!
N
~
§......-
N U:.
C
§ N..
I 0:-
I-
§ I
§
- ~
l-
-- ~
p l-I.. ~
0
!iII i*'".-
I-0-
I-N-
rb
... '- l-I-
4
.. ~ -' i
+
~
U.
ut... 4U 0-+4-I-l
~~
1 4
~ -
~
1II
~
.J.. *
~~ \\ *
\\.(1)
II
,,~
if'.
r"\\
~~
~
~
~
~
I~~
"""~*""'"
~
t'\\
--... j'"---.......
~"'"
~
4 hp 8
8 8
8 8
o o-o V'o o
fi ** ten c** 1111.'ld, I
~Nuclear TMI Unit 1 Update* 5 7/86 p.3.FIG-35 Percent Fission Gas Released as a Function of the Average Temperature of the U02 Fuel Fig. 3.2-35
d-of-Life lUll;
- 3. DI,I-IIVIU 1.10 (partial Rod Insertion)
I I
---~ FUll lidplln.
Cor. I.ttll I
PIP - 1. 50 (Iodifi.d COlin.)
- 1. 4'---~---+--~~-r-.....-~.........
1.1--....-
....-~..............
I
....-._--t--I--/-...'~,£.. PIp*
1.1
\\1
/
\\
1-I
- o. 0....~_~_..........,j~....._..A..-_........._....._....-._~
0.' ~..,...-......---...+---t---+--+-----t~.....-
......~-+...............
I I
- a. 4....--.....----....t~--+_-.....--I---+----+~-~....__t O. 2....--......--~~--+_-.....--t---.....---+--~...--tI
'L Iii
~
~
D.' ~~~-+-....._~..---+--...........--.......+---......~+-......t-t.......
20 40
&0 10 110 120 14a Dlltlnc. trGl lott.. of Acti,. FUll. in.
Iml!]Nuclear 1MI Unit 1 Update* 5 7/86 Axial Local to Average Burnup and Instantaneous Power Comparisons
- p. 3.FIG-36 Fig. 3.2-36
Maximum DeSii" Diametral Gap "0
800
~
u 0-600 lU
- )
400 200 0
0 8
12 16 20 24 linear Heat Rate,.
kw/ f t 2400 2200 2000
~
0 1800 I
~-
I... J600
- l-CD Noml na I lU 1400 DI a.e t raI Gap u="'
T" 1200 u
- )
l:'c:
0 u
1000 c.
ftI e.,:,
0:imNuclear 1MI Unit 1 Update* 6 7/87 Fuel to Clad Gap Conductance 'or End-O'-Cycle Conditions (Initial Cycle)
- p. 3.FIG-37 Fig. 3.2-37
2D I
I I
~
I I
lalimull DesiRR Clearance 1.1 Alial Power and EDl Burnup Snape with Closed Pores 1.7 Ilill Power Ind EDl Bu rnup Shape
- ith Op.en ~
'ores 1.5 Ilial POler Ind EDL Burnup Snape with Closed PQres 5
15
~
10 1:1 U
~..u o
o 2
4 6
I Initial Cold Diametral Clearance. in I
103 10 BimNuclear TMI Unit 1 Update* 6 7/87
- p. 3.FIG-38 Fission Gas Release for 1.5 and 1.7 MaxlAvg. Axial Power Shapes (Initial Cycle)
Fig. 3.2-38
3500 3000 Desi an l illi t 114" Overpower 100~ Power 2500
&Ita.
c Closed u
m 2000
=.-
1.7 AI I aI Po.er an uc EDt Bu rnup Snape
&Itc 1.5 Axial Power ana u
EDl Burnup Snape
~
- )
&It Pores
&It 1500 u
~
1.7 Ali al Power and Q,.
&It EDl Burnup Snape CoD 1000 MIII.U. Desie" Cleara"Ce~
500 o
2 4
6 8
10 Initial Cold Diaetral Clearance.
in I 103 BiE]Nuclear TMI Unit 1 Update* 6 7/87 p.3.FIG-39 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
, I Ik In. Fie tor CD HO' UN I' Cfll (ntllllp, IIISI Flctor CD "OJ UU CELL CD HO' CO'NlI CELL
HOT COhUOl 100 ULL Bil!INuclear TMI Unit 1 Update* 6 7/87
- p. 3.FIG-40 Nominal Fuel Rod Power Peaks and Cell Exit Enthalphy Rise Ratios (Initial Cycle)
Fig. 3.2-40
~--
t I
hclllr "1 flCtor Q)
NOT UNIT CEll (nUIII" ** 11 'acur
~
MOT tAll CEll OJ MOT CUNEa CEll
~
MOT CONTaOl 100 ell l BENuelear TMI Unit 1 Update - 6 7/87
- p. 3.FJG-41 Maximum Fuel Rod Power Peaks and Cell Exit Enthalphy Rise Ratios (Initial Cycle)
Fig; 3.2-41
1.1 1.5 1.4 1.3 1.2 1.1 CD 1.0 I
CI-..
N 0.9 I...
- a 0.1 0.7 IC
- a-
~...
0.&
z-..
0.5 u
0
~
0.4 0.3 0.2 011
\\
G* 2.21 I 101 IIi/tH-f t2
\\.'\\
I\\:
1-3 DNa Milt Flu.
!\\.
(DISII" Lillit)
'\\
~'\\
r\\.
lin IIIUII ON** 1.55
'\\
\\
\\
~
\\
~
/
V
~~
\\
~ I
/
1\\
V V
Calculatld Surflc.
\\
Milt Flu.
~
J'
\\I o
540 5&0 510 600 620 140 110
&10 100 120 LOCII Enthalpy, Itu/III BimNuclear 1MI Unit 1 Update* 5 7/86 p.3.FIG-42 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
1.1 1.5 1.4 1.3 1.2 1.1 UI 1..0 I
0-
)(
0.9 N...-I 0.8 s::"-
~...
CD O. 7
- I-..-
0.6 O.~
lQ U
Q
~
0.4 0.3 0.2 O. 1
~\\
\\
,.2.5' I 10' Ib/tH-f t2
~
I I
I I
~
1 r
I 1
~
1-3 ONI H**t Flu.
(D.Slln l'.' t)
~
\\
1\\
\\
~
.inl~u. ON'** 1.12
\\
\\
~ I
~
V V
~
I
\\
V "1
Calculated Surface
~
J J
....t Flul
\\,
o
~40 560 510 600 620
&40 6&0 680 700 720 local Enthalpy, Itu/lb 0ENuclear TMI Unit 1 Update* 5 7/86 p.3.FIG-43 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 ROO ASSEMBLY PL ENUM ASSEMBLY OUTLET NOZ ZLE
~;URvE ILLANCE 5PECI~I1EN HOLDE R TUBE
~OD GUIDE TUBE CORE SUPPORT SHIELD INLET NOZZLE FuEL ASSEMBLY REACTOR VESSEL THERMAL SHIELD GUIDE LUGS INCORE INSTRUMENT GUIDE TUBES
________ IN CO REI NS TRUM EN T N 0 ZZ L E S Bil!1NucIe.r TMI Unit 1 Update*5 7186 p.3.FIG-44 Reactor VI_lind Intlmlll
- Gen. Arnngement Fig. 3.2-44 TMI Unit 1 Updated Final Safety Analysis Report Reactor Vessel and Internals - Gen. Arrangement Figure 3.2-44 Rev. 20, 04/10
CONTROL ROD ASSEMBLY LDCATION INCORE INSTRUMENT LOCATION REACTOR VESSEL THERMAL SHIELD CORE BARREL SURVEILLANCE SPECIMEN HOLDER TUIE BimNuclea, TMI Unit 1 Update* 5 7/86
- p. 3.FIG-45 Reactor Vessel and Internals - Cross Section Fig. 3.2-45
0imNuclear 1MI Unit 1 Core Flooding Arrangement p.3.FIG-46 COlE fL ODD IN&
NOZZLE Update* 5 7/86 Fig. 3.2-46
1-,
--+----+++---tr---- t ---~--+;---+--
rn I
eMf I....'
MilL.
SICT** I-I Bi!lNucl.r 1MI Unit 1 UI'C..I lie
.u".
'URl
.IU Update* 5 7/86
- p. 3.FIG-47 Internals Vent Valves Fig. 3.2-47
BimNuclear 1MI Unit 1 SEE SECTION BELOW Update* 5 7/86 p.3.FIG-48 Internals Vent Valve Clearance Gaps Fig. 3.2-48
TOP VIEW Update* 5 7186
~NucIe8r TMI Unit 1 FUll AlUmbly COHTROl. ROO GUIDE TUH
,'NSTRUMENTATION I
TUM I
r i
i J
, ',',r.
I i
i
' \\,
9-~
FUEL ROO ASSEMILY i
I r::
1:7' Ir1I n
'n I
r I I
I T
1 1 CROSS SECTION I
1
, i I
1
\\
I [
i ] 1 *
[
J l!
I r LOWER END FITTING
~
Ii I
-'-~--'
- r-'---I'ir '~
I INSTRUMNTATION TUH.-l CONNECTION SNeER GRID U"O ENO trlTTING
)
)
Fig. 3.2-49 p.3.FIG-49
SPIDER TOP VIIW COUPLINt NEUTRON ~IING MATERIAL CONTROl. ROD p.3.FIG-50 ra::imNuclea, TMI Unit 1 Control Rod Assembly Update* 5 7/86 Fig. 3.2-50
SPIDER TOP VI!W NEUTRON ABSORItNG MATERIAL~
~Nucl..r 1MI Unit 1 Axial Power Shaping Rod Assembly Update* 5 7/86 p.3.FIG-51 Fig. 3.2-51
SPIOfR ----
IURNAILI POISON ROD TOP VIIW
.URNA.LI POIION MATIRIAL
- p. 3.FIG-52 0iI!JNuclear 1MI Unit 1 Burnable Poison Rod Assembly Update* 5 7/86 Fig. 3.2-52
TOP VIEW SPIDER-----....
ORIFICE ROO
- p. 3.FIG-'53 Bil!lNuclear TMI Unit 1 Orifice Rod Assembly Update* 5 7/86 Fig. 3.2-53
tOJSING ASSY LOAD ARM ASSDfSl.Y p.3.FIG-54 0:W!lNuclear TMI Unit 1 Side View of BPRA Retainer Update* 5 7/86 Fig. 3.2-54
UPPER CORE PLAT! ASSY PAD TYP p.3.FIG-55
~Nucl..r TMI Unit 1 Top View of BPRA Retainer During Operation Update* 5 7/86 Fig. 3.2-55
POsmoN INDICATQIIt AlIDIILY STATOR.IIIMLV
- p. 3.FIG-56 I
I CCUIL.NI AlIDIILV i i BimNuclear TMI Unit 1 Control Rod Drive - Gen. Arrangement Update* 5 7/86 Fig. 3.2-56
\\)
SECII** Z*Z SECII** I-I SIC"'I,.,
/E
//
.c:...- WlI' VALl( ASSEI.U
/IIT.I "Il lUI SUI' II' USEI.U L. l,l, Slell.. 1*1 SECII** 1*1 lUI SCll' **.*l **S*.I.
-r~j L
lUI SCIII II"'"
'U. sml......"
r~
~:=
~~~:ZZ~~~~'~. /
c-~ -~
- tE:.-~
w~I.tiSii5iSi~~~~~~~
l
~ _~-=:-O~
~'IICTI' Cllllll.... ""1lI
"" CII"
L~."-
L S
SEC"" 1111 SEC 110.,.,
SEC""
S-S Bil!INucI**r TMI Unit 1 UpdItI. 5 7111 p.3.FIG-57 Fig. 3.2*57 TMI-UNIT 1 REV.
19, APRI L 2008 p.3.FIG-3 Stuck Rod Location BOC and EOC FIGURE 3.2-3