ML18117A361

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Revision 24 to Updated Final Safety Analysis Report, Chapter 3, Figures
ML18117A361
Person / Time
Site: Three Mile Island Constellation icon.png
Issue date: 04/06/2018
From:
Exelon Generation Co
To:
Office of Nuclear Reactor Regulation
Shared Package
ML18117A343 List: ... further results
References
TMI-18-047
Download: ML18117A361 (57)
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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 I Control Rod Locatton and Group Oeatgnatton*

For TMl-1,Current Cycl*. CAD FILE:SIA,SKM.00,0343,000-

,0001 Ftg.3.2-1 x A B c D E F G W-H -Y --K L M N

  • 0 p R I 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 RODS IN HZP WORTH OF STUCK ROD (%k / k) Groups 1-7, 0% WD See Table 3.2-4

  • BOC Maximum Worth St uck Rod** EOC Maximum Worth Stuck Rod
p. 3.FIG-3*

.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 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 . 60 *w ffi d =-> 6'0 ::> H 20 .... (!) tjl -0 0 -\ \ \ \\ \\, .. V-2.4°/o6k/k(wtth Stuck Rod} 1 \ \'-...... __ *---. 5.4o/o Ak/k 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 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%)

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 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 U0 2 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 .,, 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 / /

/ // SuDcooled

/ / / 11 0 p. 3.FIG-21 Des I gn 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, Power
  • MaxiMum Deaign Conditions, 1303 Power Pr.obable Conditions, Power *Most Probable 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 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 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 \

Probable Cond1t1ons.

110 Power *Most Probable Cond1t1ons.

Poter * * ** * .. * * * * * * -* * * ** * * * * * * * * * .. \ * .... * * .. . * *

  • I \ *
  • lunle To Annular (laker) * * ** * * * * * , . * * * * * ' * * * * * ** '. * * * * * .. * * *"' lubbl* To J --Slug (laker) _/ _._..-----

0 5 p. 3.FIG-24 10 15 20 Oualit1 (lb vapor/total 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 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 .

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 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 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 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 j I
  • 4 ** 0 1 ... '* --It -0 +
  • 4 . 4 \ * ** --* .. ** __L 4 \ * \. Cl> '* .. * ..... "' l"O. 10: .. """ """. ,..., -...... -..... * ** 4 8 8 8 0 ---_, *
  • 0 -I., ... .,---0 -f"t WI ., ., N -I I I I ._ _,9--C U I .................

'° ... 4 u 0*+4-""""" .... ii n ..Ii-0 -"' 0 0 0 .. § ...

  • I .., .. I ... .. I I .; -:I ... .. -§
  • a. I! N u .... i u .. .. -N u .. c § .... *--... N .. ! I 0 =--I -§ 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)

.... --..-......

...... I -3M D1y1*IVIU a-of-Life IU/ii .... ** I

.......

....... -...........

--..... +---.... -+-..........

--t -Core l*tt* I 20 40 &O I Fu* I lidp I 1n1 I I Core Top 144*

... 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 '\. (D1111n Li*it) ' " \ \. lin 111u* DN*

  • 1. 55 \ \ \ \. ...........

\ / \ I "' \ v I I Calcul1t1d Surt1c1 \ Miit Flu1 l J v 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 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 0. 7 ... :I -.. -0.6 ... ., % -IQ u a 0.4 0.3 0.2 0. 1 0 ' \ \ '. 2.51 I 1 QI lb/tu-f t2 \. I I I I l r l l , 1-3 DNI Hiit Flux \ (Du 11n L 1*1 t) \ \ \ \ \

DNIR. 1.12 \ \ .-............

\ I / " v ' I \ , J Calculated Surface \ Nut F lu1 \, 560 510 600 &20 &40 6&0 680 700 720 Local Enthalpy, ltu/lb p. 3.FIG-43 18 ijJ I Nuclear TMI Unit 1 Update

  • 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 :;uRvE ILLANCE SPECIMEN HOLDER TUBE LOWEP, GRI 0 FLOW rJJSTR18UTOR
p. 3.FIG-44 CONTROL p,oo DRIVES STUDS INTERNALS VENT VALVE CONTROL ROD GUIDE TUl3E CORE SUPPORT SHIELD INLET NOZZLE FUEL ASSEMBLY REACTOR VESSEL THERMAL SHIELD GUIDE LUGS INSTRUMENT GUIDE TUBES INSTRUMENT NOZZLES 0il?]Nuclear TMI Unit 1 Update* 5 7/86 Reactor Ve111l 1nd lntem1l1 *Gen. Arnngement Fig. 3.2-44
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

p. 3.FIG-46 l?HJJ I Nuclear TMI Unit 1 Core Flooding Arrangement CORE FLOODING NOZZLE Update -5 7186 Fig. 3.2-46 z--, .......

rn p. 3.FIG-47 Cllf """' AllLI Id 5JlJNucl*r TMI Unit 1 Internals Vent Valves UllCIH lll *uct* HlllL llU Update

  • 5 7/86 Fig. 3.2-47
p. 3.FIG-48 SEE SECTION BELOW ldijlJNuclear TMI Unit 1 Internals Vent Valve Clearance Gaps Update
  • 5 7/86 Fig. 3.2-48

) ) p. 3.FIG-49 DID ,.ITTIM LOWER ENO FITTING ,............

1; . I ::r--'---1 INST!tUMt:NTATION TUH _J CONNECTION ' . l: l I\ ' I 1r; I . a .... /-' i I I I I TOP VIEW I INSTRUMENTATION I TUH I . I' I I ' T. J . l COHT!tOl.

ROD l(JJ i ;JI. GUIDE TUH I ! l r A ' I I I I/ [] /!. I ;:: ,, . i I FUEL ROO ASSf:Mlll.Y i I i I . J'I a a Ci I

  • I I I I I I I I I
  • l CltOSS SECTION [!m!Jtlucleer TMI Unit 1 Fuel Aaembly Update
  • 5 7186 Fig. 3.2-49 TOP VIEW NEUTRON MATERIAL CONTROL.ROD
p. 3.FIG-50 COUPL.INt Id 5JJJNuclear TMI Unit 1 Control Rod Assembly Update -5 7/86 Fig. 3.2-50 COUPLING SPIDER TOP VIEW NEUTRON ABSORltNG . CONTROL ROD p. 3.FIG-51 l<l ijl I Nuclear TMI Unit 1 Axial Power Shaping Rod Assembly Update -5 7/86 Fig. 3.2-51 SPIDER----

IUftNAILI POISON ft OD TOP VIEW IUftNAILI POllON MATllUAL p. 3.FIG-52 1?1 ijlJNuclear TMI Unit 1 Burnable Poison Rod Assembly Update

  • 5 1186 Fig. 3.2-52 SPIDER----

TOP VIEW ORIFICE ROD p. 3.FIG-'53 111 ijl I Nuclear TMI Unit 1 Orifice Rod Assembly Update -5 7/86 Fig. 3.2-53

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 POllTION INDICATOR AllDllLY p. 3.FIG-56 STATOR AIHMLY l" / MACTOll VllUL MUD I I CCUILINI AllDllLY ldijllNuclear TMI Unit 1 Control Rod Drive
  • Gen. Arrangement Update -5 7/86 Fig. 3.2-56

'\ ) SECllll J.l SICTlll Y.Y 1 *ITH Jiil llllU ASSEllLY LUI SCHW Hiil 1111111 ucnu T*T SlCllll S.S UCTIH 1111 p. 3.FIG-57 SICllll J.I SICllll l*I LUI SCHI HT USHILY SlCflll ... L. l,l, HCTlll 11 HCTlll l*I [!iil!INucl*ar TMI UnH 1 Updltl

  • 5 7111 Fig. 3.2*57
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