ML19312C806

Tech Spec Revisions for Cycle 3

ML19312C806
Person / Time
Site:	Oconee
Issue date:	06/21/1977
From:	DUKE POWER CO.
To:
References
NUDOCS 8001090606
Download: ML19312C806 (46)

Text

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TECHNICAL SPECIFICATION REVISIONS FOR OCONEE UNIT 2, CYCLE 3 6/21/77 S + .2 2 0 g?,"[ # WY?

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. 1 10F llATED THElHIAL PCsER DN84 Lluli . 120 f.33 S.ii2> \b '112' (21.3.112) 110 lACCEPIABLF 4 PCuP gGPERAT10N l \ ~N MeFTLIWii As FT L14ti ^ l - 100 y33.lC2)

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60 50 40 30 20 -10 0 10 20 30 40 50 60 A al Power imoalance.

• CURVE REACTCR COOL ANT FL0s (GP4i I 3T4.880 2 280'035 UNIT 2 CORE PROTECTION $AFETY LIM:TS s

3 183,690 ,

sn rowie, OCONEE NUCLEAR STATION Figure 2.1-2B 9

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THERMAL POWER LEVEL, i UN ACCEPT ABLE OPERA TION

(-23.1.105.5) kl05.5) (1,1.3.105.5)

M 2

.897 100 M, 1.259 l '

i (23 96)

ACCEPTABLE l

' (-37.89) l4 # UMP _ 90 g l OPERATION I I I l ( 7 6. 8 ) _ 80 g i I 70 l (23.68.3)

ACCEPTABLE l

(-37.68.3) l3 & 4 PUMP _ 60 l l OPERATION l (58.7)g 50 I I

- 40 l (23.41.2)

(- 3 7, 3 4. 2 ) ACCEPTABLE  !

2,3, &4 - 30 l l PUMP OPERATION l e l I k 2- l - 10 l ".

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-60 -50 -40 -30 -20 -10 0 10 20 30 go Power lobalance. .1 PROTECTIVE SYSTEM MAXIMUli ALLOWABLE SETP0litTS UNIT 2 ,

{Y OCONEE NUCLEAR STATION Figure 2.3-2B l

2.3-9 -

,l:

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CPERAil0N IN THi! (174 2 102) i o (22L B 102) gg REGION l$ NOT RESTalCTED ALLONE,u (174 ; 92) (225 8.92) 4 POWER LEVEL 80 - t 160. 2.80 '

(239.8.80)

SHUTOOWN

, MARGIN LIMli (146_2 70) (253.8.70) 60 -

(132.2.60s

.a (267.8.60) 0

- (45.50) (118 2.50) PERWISSIBLE (281.8,50)

OPERATING l - REGION 40 (300.37) 20 (0,15) (70.15) 0 e i , , i i , , , ,

O 50 t00 150 200 250 300 Roa inces. ' Witnoraen 0 25 50 75 100 0 25 50 75 100 I I 1 I f I f  ? f f Group 5 Group 7 0 25 50 75 100 I f f f f Groug C R00 POSITION Lb,ITS FOR 4 PUMP OPERATION CROM 0-100 + 10 EFPD,

. UNIT 2 ,

pe row k OCONEE NUCLEAR STATION we Figure 3.5.2-1B1 4

3.5-14 J

(181.102) ,,(225.8,102) 100 -

OPERATION IN THl3 RESTRICTED REGION REGION IS NOT ALLOWED W 5.8,92)

POSER LEYEL CUTOFF SHUT 00tN WARGIN 80 - LiglT ISO 2.80) ( 239. J. 80 )

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_ (146.2,70) '253.8,70)

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(281. 8.50)

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(56.15) (70,15)

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0 50 100 150 200 250 300 Roa inlex, 5 Witnarawn 0 25 50 75 100 0 25 50 75 100 i i i 4 e i i , , ,

Group 5 Group 7 0 25 50 75 100 t  ! , , l Group 6 R0D POSITION LIMITS FOR 4 PUMP OPERATION FROM 100 + 10 TO 250 + 10 EFPD UNIT 2 -

I OCONEE NUCLEAR STATION Figure 3.5.2-182 l

3.5-14a .

RESTRICTED REGION 100 ~

6.5.102) (31.4.102) RESTRICTED REGION (4 1.92) d

'( 3 3. 8,9 2 )

90 .

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80 . (1 7,80) < >(33.8,80) 70 (0,70) , ( 33. 8,70) b 60 - ~

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PERMISSIBLE 2 OPERATING 30 REGION 20 -

10 -

0 - . . . . .

0 10 20 30 40 50 60 70 80 90 100 APSR, "s Withdrawn APSR POSITION LIMITS FOR OPERATION AFTER 250 + 10 EFPD

~~

UNIT 2 OCONEE NUCLEAR STATION Figure 3.5.2-4B2 l

I 3.5-23g i

i (98,102) (132 2.102) 100 STRICID (160.2.102) (W'O

- OPERArl0N IN THl3 EGOi RESTRICTED

, REGION IS NOT FGi 2 & 3 PUNP REGION FOR 3

= - ALLC#ED WITH 2 OR 3 PM TEUTION (146.2,89) (253.8,89) pggp E RESTRICTED OPERArl0N 5 80 -

IN THl3 0 (I32 2 73) (267.8,76)

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REGION O

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RESTalCTE0 OPERATION IN THIS RES10N FOR

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AMD THREE PUMP CPERATION FROM 100 + 10 TO 250 + 10 EFPD

!! NIT 2 eau rowra OCONEE NUCLEAR STATION

' - Figure 3'.5.2-2B2

3. 5 - 19a e

Poner, ", of 2568 48t RESTRICTED REGION

, (-20.5.102) ( +14. 2,10 2 )

100

(-19.0.92) (+13.4,92) 80 -

70 -

60 -.

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l OPERATIONAL POWER IMBALANCE ENVELOPE FOR OPERATION FROM 100 + 10 TO 250 + 10 EFPD

~

UNIT 2

[

OCONEE NUCLEAR STATION Figure 3.5.2-382 3.5-22a

i l

i 100 -

OPERATION IN THIS REGION 15 NOT RESTRICTED REGION 3pp

= - ALL0tED slTH 6 59)

I 2 09 3 PU4PS FOR 2 & 3 PuuP yggyim L

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REGION O

j 40 -

~

a J

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g 40 -

(69.15) 0 i , . , , , , , , ,

0 50 100 150 200 250 300 Roa inces, *, fitnaraan 0 25 50 75 100 0 25 50 75 100 t ' i i i . , , , ,

Group 5 Group 7 0 25 50 75 100 i r , , ,

Group 6 R0D POSITION LIMITS FOR TWO AND THREE PUMP OPERATION AFTER 250 + 10 EFPD

~

UNIT 2 i OCONEE NUCLEAR STATION Figure 3.5.2-2B3 3.5-19b

100 _

(198.102) .'(251.6.102)

A

_ OPERAil0N IN THl3 t (25l.6.92)

EGiuN !S NOT ALLOWE0 CUTOFF 80 - (237.6,80)

SHUT 00eN 4ARGIN (223.6.70) j LIMIT RESTRICTED 60 - REGION (209.6.60) g C

o -

(121.50) (195.6,50)

( 40 -

E PERMISSIBLE OPERATING 20 -

(69,15) REGION (146.15) 0 ' '

t i , , , , , , ,

0 50 100 150 200 250 300 Roa index, *. setnaraan 0 25 50 75 I n 0 25 50 75 100 Group 5 gray, 7 0 25 50 75 100 t , . , ,

Group 6 R00 POSITION LIMITS FOR 4 PUMP OPERATION AFTER 250 + 10 EFPD UNIT 2

{w OCONEE NUCLEAR STATION Figure 3.5.2-1B3 3.5-15

Amend.ent tc 3A'. -14 5 2.

u, . v.,st- - t .. i _ , Cs C.ut 3

- Reload Pcport --

Jun: 1977 E.'.ECOCK & k'II.COX Power Generation Group

?!uclear Power Generation Division P. O. Box 1260 Lynchburg, Virginia 24505 Babcoch & Vlilcox

Amendment 1 (6/15/77)

INTRODLT JT ON This docuncat auends the Oconee thi t 2, Cycle 3 Reload Report , BAU-1452 (April 1977), to account for modifications to the Cycle 3 desi,;n parar.eters necessi-tated by a lover Cycle 2 CFPD burnup than was assumed in the Reload Report and by the loading of 12 once-burned assemblies from Oconee I'n i t 3 into the core. The revised Cycle 2 burnup 12, 277 EFPD, which in 29 EFPD shorter than the Cycle 2 design length of 306 EFPD used in the Reload Report. The Cy-cl. 3 length, however, was extended f rom 292 to 300 EFl'D.

This artendment consists of the follouing revised pages f rom the Reload Report:

2-1 4-5 5-7 7-8 8-10 3-1 4-7 7-2 8-3 8-11 3-2 5-1 7-3 8-6 8-12 3-3 5-4 7-4 8-7 8-14 3-4 5-5 7-5 8-8 8-15 4-1 5-6 7-6 8-9 8-17 4-4 e

e, ee e a be % e j

l

Amendment 1 (6/15/77)

2. OPERATING HISTORY The reference fuel cycle for the nuclear and thermal-hydraulic analyses of the third cycle of Oconce Nuclear Station, Unit 2 is the presently operating Cycle

2. Cycle 2 achieved initial criticality on July 8,1976, and power escalation began on July 12, 1976. The 100% power level of 2568 MWt was reached on July 18, 1976. No operating anomalies occurred during Cycle 2 operation that would adversely affeet the fuel performance in Cycle 3 during the design length of 300 IIPD. No control rod interchanges are planned for Cycle 3. Control rod l1 group 7 will be withdrawn at 250 (110) EFPD of operation.

s i

2-1 Babcock & Wilcox l

Amendment 1 (6/15/77)

3. GENERAL DESCRIPTION The Oconee 2 reactor core is described in detail in Chapter 3 of the Oconee Nuclear Station FSAR.I The Cycle 3 core comprises 177 fuel assemblics (FAs), l1 173 of hich have a 15 by 15 array containing 208 fuel rods, 16 control rod guide s, and one incore instrum.cnt guide tube. The cold-worked Zircaloy-4 fuel rod cladding has an OD of 0.430 inch and a wall thickness of 0.0265 inch. The f uel consists of dished-cad, cylindrical pellets of uranium dioxide (UO2 ), which are 0. 370 inch in diametet . (See Tables 4-1 and 4-2 for addi-t ional data.) The other four FAs in Cycle 3 are demonstration 17 by 17 assem-blies - two Mark C and two Mark CR. All FAs in Cycic 3 except the 17 by 17 demonstration assemblics maintain a constant nominal fuel loading of 463.6 kg of uranium. However, the undensified nominal active fuel lengths and theoret-ical densities vary between batches; these values are given in Tables 4-1 and 4-2.

Figure 3-1 is the core loading diagram for Oconee 2, Cycle 3. All the batch 2 assemblies will be discharged at the end of Cycle 2. Five once-burned batch 1 assemblies with an initial enrichment of 2.06 wt % 235U will be reloaded into the central portion of the core. Batches 3 and 4, with initial enrich-ments of 3.05 and 2.64 et % 235 U, respectively, will be shuf fled to new loca-tions. Batches 5 and Sa, with ini tial enrichme.its of 3.03 and 2.53 wt % 235 U, will oc.cupy primarily the core periphery and four interior locations. Twelve once-burned assemblies from Oconce 3, Cycle 1, with an initial enrich', ment of 2.01 we % 2 35 U will be loaded into the core and have been designated as batch 1-X. Figure 3-2 is an eighth-core map showing the assembly burnup and enrich-ment distribution at the beginning of Cycle 3.

, Reactivity control is supplied by 61 full-length Ag-In-Cd control rods and sol-uble baron shim. In addition to the fu]]-length control rods, eight partial-len.;th axial pescr shaping rods (APSRs) are provided for additional control of axial power distribution. The Cycle 3 locations of the 69 control rods and the group designations are indicated in Figure 3-3. The core locatlons of the 3 -I E~ *N #' Y'D

Anendment 1 (6/15/77) f total pattern (69 centrol rods) for Cycle 3 are identical to those of the ini-Lial cycle described in Chapter 3 of the F5AR.I However, the r,roup designa-2 to minimize power peaking.

tions dif fer be t..cen Cycle 3 and the ref erence cycle The nc=Iral systen pressore is 2200 psia, and the core average densified nemi-nal heat rate is 5. 79 ':'J/ f : at the rated core power of 2<68 hWe.

Babcock & W;lcox 3-2

Amendcient 1 (6/15/77)

Figiare 3-1. Core Loading Diagran, Oconee 2, Cycle 3 Fi'El. IFJ5SiE2 C/d/.L l>

I A 5 5 5 5 5

.i B '

s- <>

. ..i lex S'k C 3 3 c? 4 3 4 07 3 5 4 C

4.0 13 CYI ?10 73 1 P6 CYL l7 . F1 D 5 5 3 3 4 3 54 3 4 3 3 5 5 r, n!- ai; pi  ; n, n.. .c 7 E 3 3 3 . 4 4 1 4 4 4 3 3. 5 re +. st1 1- w ." 2 i P4- R7 R6 N3 .; 6

' .: 1-x l .s t .t F 5 ) si3 4 4 .12 4 4 4 N4 4 5 K1 5 5 t"t ni. trl 9 4 cyl cl '2 (yl G $ 3 i, 3 4 4 .  ; ..  ; 3 4 3 5 ui it- .. i 7 ;ts et? n13 ryl .i t v4 a:2 v '. t? M3 i l 1

) 3 3 5A 3 4 :1* "l3 l'i 4 3 5A 3 3 5 -

H 3,37 ,,,,, 73 , 3 yy.. Ov1 rY1 CY1  !!I 12 1 H2 18 4 1

5 3 4 3 4 4 I 4 4 4 3 4 3  %

K a I' l l s'14 E12  !)t h C12 ( i i Cil C3 C4 D2 F. i F2 E3 1-X l-X l-K 1-X L 5 5 G13 4 4  !>l 2 '. 4 4 D4 4 4 C3 5 5 rY! r14 c i ', rvi n9 A8 47 rc. : c1 r? i:y1 M 5 3 3 4 4 4 3 4 4 4 3 3 5 010 Db AID Y# B12 C8 B4 A7 1.1 D3 c.6 Mk CR N 5 5 3 3 4 3 5A 3 4 3 3 5 5 04 Cl? Bil D11 D5 R$ C '. G7 1-X l-X .M K C 0 4 5 3 C9 4 3 4 c1 3 5 4 f 1 's F3 rvi M10 pg RA cyl F7 A r, P 5 5 5 3 3 3 5 5 5 C11 us c5 5 5 5 5 5 -

R l

z 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15

- l Lit. li l'rs v iaan C. ore 1.ocat ion l

l l

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i 3-3 Babcock & Wilcox

Amendment 1 (6/15/77)

Figure 3-2. Enriel. ment aad T.itrau;, Distribut. ion, Oconec 2, Cycle 3 e 9 10 11 I? 13  ! .. 15 2.% 2.06 2.6 . J.05 .'.33 3.05 3.05 3.03 11 12,'f 6 3 10,931 6. 3 % 21, M1 0 18,563 21,711 0 2.64 2.6'4 '

. 65 s.05 2 . 6 '. 3.05 3.03 K (

6.5E 10,671 5,991 21,200 F , 9')5 18,637 0 2.01 2.64 2.64  ?.01 3,03 3.03 1.

12.531 6,157 8,360 16,281 0 0

_.64 3.03 3.05 3. fi 24 5,290 17,2(.2 23,085 0 3.05 3.03 3.03 22,384 0 0 2.64 0

5,302 P

R ,

x.xx In i t. h] Enrichment.

xxxxx P.0C. P.urnup, !NJ/etU L

3-4 ,

Dabcock & Wilcox

e t

Amendment 1 (6/15/77) 9 I

4. FUEL SYSTEM DESIGN 4.1. Fuel Assembly Mechanical Desigi

- The Cycle 3 core cceprises the normal resident and reload Mark B FAs plus five batch 1 assemblius, 2 Mark C (17 by 17 array) demonstration assemblics (part of batch 4), 2 frcsh Mark CR (17 by 17) demonctration assemblics, and 12 Mark B FAs from batch 1 of Oconce 3. The pertinent fuel design parameters and dimensions are listed in Table 4-1. The five FAs from hatch 1 and the 12 from Oconee 3 t batch 1 (referred to as batch 1-X in this report), which are all once-burned assemblies, are mechanically similar to those in batch 3. The fresh FAs (batch

5) incorporate minor design modifications to the spacer grid corner cells, which reduce spacer grid in:craction during handling. In addition, improved test methods (dynamic impa t testing) show the spacer grids to have a higher seismic capability and thus 2n increased saf ety margin over the values in reference 3.

The two Mark CR assemblies included in batch 5 are mechanically identical in function to the Mark C demonstration assemblies of batch 4. The Mark CR assem-blies are different because they have reconstitutable lower end fittings; how-ever, no reconstitutability tests are scheduled for the Oconee 2 pool. The mechanical design of the Mark C demonstration assemblies and the comparison be-tween the Mark C and B assemblics are given in reference 4. All fuel assemblies in the Cycle 3 core are mechanically interchangeable, but the Mark C and CR dem-onstration assemblies cannot be located in a rodded location. The static and dynamic structural characteristics of the Mark C and CR demonstration assemblies are compatible with the Mark B assemblies and have been designed to maintain their mechanical integrity throughout the three cycles of operation and to i

-k successfully withstand all seismic and LOCA loads postulated for the Oconee 2 L

reactor.

All other results presented in the FSAR discussion of the fuel assembly mechani -

l cal design are applicable to the reload FAs.

I 4-1 Babcock s.Wilcox

Amendment 1 (6/15/77) i 1. The maximum specification value for the fuel pellet diameter was used.

2. The maximum specification value for the fuel pellet density was used.

j 3. The claddin?, ID used was the lowest permitted specification tolerance.

4. The maximum local pellet burnup (expected three-cycle) is less than 55,000 mwd /mtU.

4.3. Thermal Design All fuel assemblies in this core are thermally similar. The fresh batch 5 fuel inserted for Cycle 3 operation does not introduce any significant dif-ferences in fuel thermal performance relative to the batch 2 fuel discharged at the end of Cycle 2. The linear heat generation rate capability of the batch 1-x, 4 and 5 fuel (Table 4-4) is greater than that of batches 1 and 3 (20.15 kW/ft versus 19.8 kW/ft). These linear heat rate limitations were established using the TAFY-37 code with fuel densification penalties.

The two Mark CR assemblies in batch 5 are thermally identical to the two Mark C (batch 4) demonstration assemblies described in reference 4. The four demon-stration assemblies have been placed in nonlimiting core locations.

4.3.1. Power Spike Model (Densification)

The power spike model used for Cycle 3 analysis is the same as that used for Cycle 2.2 Figures 4-1 and 4-2 show the maximum gap size and power spike fac-tor, respectively, versus axial position. The power spike factor and gap size are based on unirradiated batch 3 fuel (92.5% TD) with an assumed enrichment of 3.0 wt % 235U. These values are conservatively high for all batch 1, 4, and 5 fuel.

4.3.2. Fuel Temperature Analysis Thermal analysis of the fuel rods assumed in-reactor densification to 96.5%

theoretical density (TDF). The analytical methods utilized are the same as those documented in references 2 and 6 for Cycles 1 and 2, respectively. The average fuel temperatures shown in Table 4-4 are taken from the analyses uti-lized to define the linear heat rate (LHR) capability for the fuel (references 2 and 6). This analysis is based on the lower tolerance limit of the speci-fication fuel density and assumes isotropic diametral shrinkage and anisotropic axial shrinkage (consistent with reference 8) resulting from fuel densifica-tion.

l t

4_3 Babcock & Wilcox l

o Amendment 1 (6/15/77) 4.4. Material Design The batch 4 fuel assemblies are not new in concept, <nd they do not utilize different component materials. Therefore, the chemical compatibility of all possible fuel-cladding-coolant-assembly interactions for the batch 5 fuel as-semblics are identical to those of the present fuel.

4.5. Operat ing Experience B&W's operating experience with the Mark B,15 by 15 FA design has verified the adequacy of this design. As of December 31, 1976, the following operating experimce has been accumulated for the six B&W 177-FA plants using the Mark B fuel assembly:

Max. assembly Cumulative Current burnup net electrical Reactor cycle mwd /mtU output, MWh l1 Oconee 1 3 23,000 15,232,533 Oconce 2 2 22,200 10,584,123 Oconee 3 2 19,800 9,933,642 TMI-1 2 23,700 11,854,960 Arkansas onc 1 19,253 8,957,632 Rancho Seco 1 10,761 3,531,597 4_4 Babcock & Wilcox

>z Table 4-1. Fuel Design Parameters and Dimensions Twice-burned once-burn _d assemhlies Fresh I ocl a .sc% i is ..

batch 3 Batch I-X("} Batch 4 Batch 1 Ma r k C < h"w l;atch 5 Mark ri: a no FA type Mark B-3 Mark B-3 Mark B-4 Mark U-2 17<17 array Mark li-i 17 17 i r e ..v No. of assemblies 48 12 54 5 2 50/4 2 Fuel rod OD, in. 0.430 0.430 G.430 0.430 0.379 0. 4 10 0.379 Fuel rod I D, in. O.377 0.377 C. 37 7 0.377 0. 332 0.177 0.332 Flexible spacers, type Corrup,ated Spring Spring Corruga w! Spring Spring Spr itw, Right space rs , type Zro j Zr-4 Zr-4 Zro; Zr-4 Zr J. Zr-4 Undensified active 144.0 144.0 142.6 144.0 143.0 !4.2.25 141.0 fuel length, in. I Fuel pellet OD (mean 0.370 0.36S2 0.370 0.370 0 . 12 '4 0.lb95 0.12 '.

specified), in.

s- s 95.3 93.5

~

& Fuel pellet initial 9.' . 5 92.5 94.0 94.0 94.0 density, % TD Initial fuel enrich, 3.05 2.01 2.64 2.06 '.64 3.01/2.53 1.01 wt % 2 "' U Initial fill y,as ;) res- (b) (Sc } (b) (b) (b) Sa ve as ';.me as Kirk sure (min spec), psia hatch 4 C dano Burunp, BOC, N.Jd/mtU 20,434 16,281 7,415 11,338 5,102 0 0 Cladding collapse 230,000 >30,000 >30,000 >30,000 30,000 230,000 s30,000 g time, EF?:1

• h a

Design life, 1:FFil 24,408 18,672 20,8% 17,760 20,3% 21,216 21.213 $

tn r>

ca 3 ty rt n -

a Q (a)From Oconee 3. m es y (b) Refer to proprietary informat. ion in reference 2, Table 4-1. h

s rs M O N x v

D .

Amendment 1 (6/15/77)

Table 4-3. Fuel Thermal Analysis Parameters Batch 1 Batch 1-X("I Batch 3 Batch 4 Batch 5 12 48 56(b) 56(C)

Nc. of assemblies 5 Initial density, % TD 92.5 95.3 92.5 93.5 94.0 ,

Pellet diameter, in. 0.3700 0.3682 0.3700 0.3700 0.3695 .

Stack height, in. 144.0 141.0 144.0 142.6 142..'5 Densified Fual Parameters Pellet dinneter, in. 0.3632 0.3649 0.3632 0.3645 0.3646 Fuel stack height, in. 141.1 140.2 141.1 140.5 140.5 Nominal LilR at 2568 5.77 5.80 5.77 5.80 5.80 1

MWt, kW/ft Avg fuel temp at 1335 1310 1335 1320(*I 1320(*)

nominal LilR, F LilR capability (cen- 19.8 20.15 19.8 20.15 20.15 terline f uct melt), KW/f t (a)From Oconec 3.

(b) Includes two Mark C (17 by 17) demonstration assemblics.

(c) Includes two Mark C (17 oy 17) demonstration assembifes.

(d)Densification to 96.5% TD assumed.

(" Matt C and Mark CR fuel will operate at a Inwer average heat rate and a corresponding lower average fuel temperature.

e 4-7 Babc0Ck & Wi!Or. .

Amendment 1-(6/15/77) 1

5. NUCLEAR DESIG:1 5_.1. Physics Characteristics Table 5-1 compares the core physics parameters of Cycles 2 and 3. The values for both cycles were generated using PDQ07. Since the core has not yet reached an equilibrium cycle, differences in core physics parameters are expected be-tween the cycles.

The shorter Cycle 3 will produce a smaller cycle dif ferential burnup than that f r Cycle 2. Figure 5-1 illustrates a representative relative power distri- y bution for the beginning of the third cycle at full power with equilibrium xe-non and normal rod positions.

4 The critical boron concenttations for the beginning of Cycle 3 are icwcr in all cases than those for Cycle 2. End of cycle conditions vary between Cyc1 s 2 and 3 causing lower critical boron concentrations. As indicated in Table l'1 5-2, the control rod worths are sufficient to maintain the required shutdown margin. However, due to changes in isotopics and the radial flux' distribution, the hot, full-power control rod worths will be less than those for Cycle 2.

The Cycle 3 ejected rod worths for the same number of regulating banks inserted are lower than those in Cycle 2. It is difficult tocomparevalues[between cycles or between rod patterns since neither the rod patterns from which the CRA is assumed to be ejected nor tne isotopic distributions are identical.

i

. Calculated ejected rod worths and their adherence to criteria are considered at all times in life and at all power lerels in the development of the rod in-I sertion limits presented in section 8. The maximum stuck rod worths for Cycle 3 are lower than those in Cycle 2. The adequacy of the shutdown margin with l1 Cycle 3 stuck rod worths is demonstrated in Table 3-2. The following conserva-tisms were applied for the shutdown calculations:

5-1 Babcock & Wilcox

i Amendment 1 (6/15/77)

Tabic 5-1. Oconce 2. Cycle 2 and 3 Physics Parameters 4

Cycle 2(#) Cycle 3( }

Cycle length, EFPD 306 300 Cycle burnup, tEd/mtU 9582 9392 Average core burnup - EOC, SMJ/utU 18,606 18,554 l Initial core loading, mtU 82.1 82.0 i

4 Critical boron - BOC, (no Xe), ppm l

! Il2P( }, group 8 (37.5% wd) 1445 1335  !

{ ll2P, groups 7 and 8 inserted 1330 1250 HFP, groups 7 and 8 inserted 1140 1070 Critical boron - EOC (eq Xc), ppm I!ZP9 434 290

, group 8 (37.5% wd, eq Xe)

, IIFP s 87 35 ,

Control rod worths - HFP " BOC, % ak/k Group 6 1.20 1.12 Group 7 0.96 0.88 Group 8 (37.5% wd) 0.54 0.33

Control rod worths - IIFP (250 EFPD), % ak/k Group 7 1 1.33 1.12 j Group 8 (37.5% ud) 0.51 0.40 lbx ejected rod worth - IlZP, % Sk/k BOC I

250 EFPD 0.59(5) 0.58 0.49((d) 0.47 d) i Fbx stuck rod worth - HZP, % ak/k

, BOC 2.16 1.94

250 EFPD 2.22 2.04 Power deficit - liZP to llFP, % ak/k '

BOC 1.65 , 1.63 l 250 EFPD 2.49 2.17 Doppler coeff. - BOC, 10-5, ak/k/*F 100% power (no Xe)- -1.51 -1.47 Doppler coef f. - EOC,10-5, ak/k/*F .'

i 100% power (eq. Xe) -1.55 -1.54 i Moderator'coeff. - HFP, 10~4, ak/k/*F BOC -1.03 -0.49 EOC - -2.60 -2.58 Boron worth - HFP, ppm /%4k/k #

BOC (1000 ppm) '

109 106

-EOC (17 ppm) 101 96 i

j 5-4 Babcock & Wilcox -

?~

i .

-  : .-- - i ,2 , , _ , _ _ . , , . , _ _ - . . --

M.e ndmcat I (6/15/77)

TabI. 3-1. Ir1..' ' d T Cycle 3(

cv. e le_2 _

Xenon uorth - !iFP,

, .; / k 2.60 2.65 EOC (4 days) 2.66 2.75 EOC (e q u i li o r lu. .)

I:f fect Ive delayed neutron fraction  !!rP 0.00577 0.00590 {1 Uuc 0.00522 LOC 0.00516 l (n)lla.5ed on Cyc le 1 lengt h o f '.40 1:I PJ and a cycIc 2 desi>;n length of 306 EFPD.

(b) Cycle 3 data are for t iu- conditionr. stated and are based on a Cycle 2 length of 177 EEPD; C' le 2 ma:. not b ._ at t he s arae conditions as Cycle 3. t ,

(c)IlZP - hot zero power; HTP - hot in11 power.

(d) Ejected rod value for groups 5, 6, 7, and 8 1.iserted.

Babcock & Wilcox 5-5

Amendment 1 (6/15/77)

Table 5-2. Shutdosa Margin Calculaticn for Oconec 2, Cvele 3(*}

EOC, ': ak/k 250 EFPD, ' 2k/k Availabic 'd 'lo r t h tlZP 8.30 S.43 Total rod wo "

L' orth reduction due to burnup -0.25 -0.32 of poison material

-1.94 -2.04 Maximum stuck rod, ifZP 6.11 6.07 Net worth Less 10% uncertainty -0.61 -0.61_

5.50 5.46 Total available worth Fequired Rod k' orth 1

Power deficit, liFP to llZP 1.63 2.17 1.19 1.20 Max allowable inserted rod worth Flux redistribution 0.40 0.69_

Total required worth 3.22 4.06 Shutdown Margiu Total avail, worth minus 2.28 1.40 total required worth (a) Required st.utdown mar,;in i s 1. 00ll, Ak/k.

(b)IlZP denotes hot zero power; !!FP denotes hot full power.

5-6 B.Shcock t. Wi!ccx

. . . -. -~ . - . .

i e

t Amendment 1 (6/15/L7) s Figure 5-1.  ?,00 ('. EFPD), Cycle 3 Two-Diraensional ?.clative Po.rcr Distribution - Fall Potter, Equilibrien Xenon : Tor ~ al 1

' Rod Positions (Groups 7 and 8 Inserted) 4 12 13 14 15 8 9 10 11 7

' l 1.07 1.14 1.40 1.24 1.34 3.91 -

0.46 0.66 H }

l i

~'

. _ _ 1 1.33 1. .' 3 1.30 1.05 0.97 0.50 0.7?

K 1.21 i

\SN 1

.'g7 I'  !.'0 4 1.23 0. f. 5 N 1.1 'i 0.97 0.79 1.17 0.70 l

1.24 1.30 1.20 f.23 1.02 0.94 1.C3 3

N 1.34 1.05 0.97 1.02 1.00 1.23 0.77 1

1 I

t 0 0.91 0.97 0.80 0.94 1.23 0.73 7

P 0.47. 0.80 1.17 1.03 0.77

t I

0.66 0.74 0.70 x Inserted Rod Group No.  !

R x.xx Relative Power Deasity 1

5-7 Babcock & Wilcox

Amendment 1 (6/15/77)

The important parameters during a rod withdrawal accident are the Cappler co-efficient, the moderator tempera ore coef ficient, and tha rate at which re-activity is added to the core. Only high-pressure and high-flux trips are interlocks, accounted f or in the FSAR analysis, which ignores multiple alarms, and trips that normally preclude th.fs type of incident. Fcr positive reac-tivity additions indicative of these events, the most severe results occur for EOL conditions. The FSAR values of the key parameters for EOL conditions were

-1.17 x 10~ (ak/k/*F) for the Doppler coef ficient, 0.3 = 10-Sk/k for the moderator temperature coefficient, and rod group worths of up to and including a 10% Ak/k rod bank worth. Comparable Cycle 3 parametric values are -1.47 x 10- (ok/k/*F) for the Doppler coef ficient. -0.49 ^ 10 * (ak/k/*F) for the i moderator temperature coef ficient, and a maximum rod bank worth of 8.30%

Ak/k. Therefore, Cycle 3 parameters are bounded by design values assumed for the FSAR analysis, and thus for the rod withdrawal transients, the consequences will be no more severe than those presented in the FSAR. For the rod with-drawal from rated power, the transient consequences are also less severe than those presented in the densification report.6 7.3. Moderator Dilution Accident The Boron in the form of boric acid is used to control excess reactivity.

boron content of the reactor coolant is periodically reduced to compensate for fuel burnup and transient xenon ef fects with dilution water supplied by the makeup and purification (MU5P) system. The moderator dilution transients considered are the pumping of water with zero boron concentration f rom the shutdown, makeup rank to the RCS under conditions of full-power operation, hot and refueling. The key parameters in this analysis are the initial boron concentration, boron reactivity worth, and the moderator temperature coeffi-cient for power cases. .

For positive reactivity addition of this type, the most severe resalts occur for BOL conditions. The FSAR values of the key parameters for BOL conditions were 1400 ppm for the initial baron concentration, 75 ppe/1.07. (Ak/k) boron reactivity worth, and +0.94 x 10 " ak/k/ *F for the moderator tcmperature co-ef ficient. Comparable Cycle 3 values are 1070 ppm for the initial boron con- 1 centraticn, 81 ppa /1.0% (ak/k) boren reactivity werth, and -0.49 x 10 (2k/k/*F) for the moderator temperature coef ficient. ~he FSAR shcus that the core and RCS are adequately protected during this event. Sufficient tima for 7-2 Babcock & \Vilcox

- - - . . . . . ~. - . _ . .-. - -- --

A l

Amendment 1 (6/15/77 I

i i nost conservative initi al conditienc -ucre accumed for the densifi:stien re- '

-5 portG FSN4 values of flow and cuastdoun, -1.17 x 10 (_k/k/ *F) Doppler coefficient, 0.5 x '10 " (ak/k/ *F) moderator temperature coefficient, with I

densified fuci power spike and peakinF. The results showed that the DNBR re-

mained above 1.3 ,(W-3) for the four-pump coastdown, and the fuel

ladding temperature remained b: low criteria limi

• s f or the locked-rotor transicnt.

1

-S The predicted paramet ric values for Cycle 3 are -1.47 x' 10 (ik/k/*F) Doppler coef ficient, 0. 4 9 x 10 " (ak/k/ *F) moderator temperature coef ficient, and

~

! l1 j peaking factors as shown in Table 6-1. Since the predicted Cycle 3 values are bounded by those used in the densificatien report, the results of that j analysis represent tha most severe consequences from a loss of flow incident.

1 7.6. Stuck-Out, Stuck-In. or Dropped Control Rod If a control rod was dropped inte the core while it was operating, a rapid l

decrease in neutron power would occur accompanied by a decrease in the core I

+ average coolant temperature. The power distribution might be distorted duc

! to a new control rod pattern under which conditions a return to full power

might lead to localized pouer densities and heat fluxes in encess of design

! limitations.

The key parameters for this transient are moderator temperature coefficient, dropped rod worth, and local peaking factors. The FSAR analysis was based on 0.46 and 0.36% ak/k rod worths with a moderator temperature coefficient of -3.0 x 10 " (ak/k/

• F) . For Cycle 3, th'e maximum worth rod at power is

~

! 0.20%.ak/k with a moderator temperature coefficient of -2.58 x 10 -4 (ak/k/*F).l1 f Since the predicted rod worth is less positive and the moderator temperature .

coefficient more positive, the consequences of this transient are less severe

than the results presented in the FSAR. .

{

i 7.7. Loss of Electric Power

! Two types of power losses considered in the FSAR were a loss of Iqad condi-tion caused by separation of the unit from the transmissien systen and a hypothetical condition resulting in a complerc loss of all system and unit l power except that from the unit batteries.

The FSAR analysis evaluated the loss of load with and without turbine runback.

[

k' hen there is no runback,- a reactor trip occurs on high reactor coolant pres-sure or temperature. This case results in a nonlimiting accident. The 7_4 Babcock r. Wilcox

l l

Amendment 1 (6/15/77) l Jargest off-site dese occurs for the second case, i.e., loss of all electrical p< tze r < 2ept unit Utter!<, . suming operation with failed fuel and steam gen-erator tube leakage. These results are indepenJ nt of core loading; therefore, 9

the results of the FSAR are applicabic for any reload.

7.8. Stcan Line Failure A stean line failure is defined as a rupture o f au:; of the steam lines f rom the steam generators. I'p o n initiatioa of the rupture, both steam generators start to blew down causing a sudden decrease in the prir..ary system tempera-ture, pressure, and pressuriser level. The temperature reduction leads to positive reactivity inaertion, and the reactar trips on bigh flux or low RC pressure. The FSAR has identified a double-ended rupture of the steam line between the steam generator and stean stop valve as the worst-case situation at LOL conditions.

The key parameter for the core response is the noderator temperature coef fi-

~

cient, which was assumed in the FSAR to be -3.0 x 10 " (Lk/k/"F). The Cycic 3 predicted value of moderator temperature coef ficient is -2.58 x 10 (ak/k/ ll

• F). This value is bounded by those used in the PSAR analysis; hence, the results in the FSAR represent the vorst situation.

L9. Steam Gener' tor Tube Failure A rupture or leak in a steam generator tube allows reactor coolant and asso-ciated activity to pass to the secondary system. The FSAR analysis is based on complete severance of a steam generator tube. The primary concern for this incident is the potential radiological release, which is independent of core loading; hence, the FSAR results are applicable to this reload.

7.10. Fuel llandlinydecident The mechanical damage accident is considered the maximum potential source of activity release during fuel handling activities. The primary concern is radiological releases that are independent of core loading; therefore, the FSAR results are applicable to all reloads.

7.11. Rod Ejection Accident For reactivity to be added to the core more rapidly than by uncontrolled rod withdrawal, physical failure of a pressure barrier component in the control rod drive assembly must occur. Such a failure could cause a pressure 7-5 Babcock & \Vilcox

Amendment 1 (6/15/77) dif ferential to act on a control rod assembly and rapidly eject the assembly from the core region. This incident represents the most rapid reactivity f r-sertion that can be reasonably postulated. The values used in the FSAR and densification report at LOL conditions, -1.17 x 10~ (ak/k/*F) Doppler ccef-ficicat, 0.5 x 10 " (ak/k/*F) moderator temperature coefficient, and an ejected

~

rod worth of 0.65% ak/k, represent the naximum possible transient. The cor-responding Cycle 3 pananctric values of -1.47 x 10~ (ak/k/*F) Doppler, -0. 4 9 l1 x 10-4 (ak/k/*F) noderator tenperature coefficient (both more negative than those used in reference 6 and a maximum predicted ejected rod worth of 0.55% l1 ok/k ensure that the results will be less severe than those presented in the FSARI and the densification report.6 7.12. Maxinum Hypcthetical Accident There is no postulated mechanism whereby this accident can occur since it would require a multitude of f ailurcs in the engineered safegt.ards. The hy-pothetical accident is based solely on a gross release of radioactivity to the reactor building. The consequences of this accident are independent of core loading; hence, the results reported in the FSAR are applicable for all reloads.

7.13. Waste Cas Tank Rupture The waste gas tank was assumed to contain she gaseous activity evolved from degassing all of the reactor coolant following operation with 1.0% def ective fuel. Rupture of the tank would result in the release of its radioactive contents to the plant ventilation system and to the atmosphere through the unit vent. The consequences of this incident are independent of core loadin;;

therefore, the results reported in the FSAR are applicable to any relo9d.

7.14. LOCA Analysi_s .

A genecic LOCA analysis for the B&W 177-FA lowered-loop NSS has been performed using the final acceptance criteria ECCS evaluation model reported in BAW-10103.10 The analysis in BAW-10103 is generic in nature since the limit-ing values of key parameters for all plants in this category were used. Fur-thermore, the combination of average fuel temperature as a function of linear heat rate and the lifetime pin pressure data used in the BAW-10103 LOCA limits analysis is conservative compared to those calculated for this reload. Thus,

)

7-6 Bobcock & W'i!ccx j

A:ncndment 1 (6/15/77j T. i b l .' 7- 1. Co a r i .~ , n o f Mc. pa rar:i t a r;, f or Acc ident .\na i v s is TSAR, dcn:if Predicted Parareter value Cycle 3 value r

COL Doppler coeff, 10 ', Jk/k/*/ -1.17 (a ) -1.47 e

1:0L Doppler coef f , 10 ', n/k/*E -1.33 -1.54 1:0L nedera tor coef f , 10~ ', 2/k/ 'F +0.5 I) -0.49 Eol. modera tor coef f , 10 ", ak/h/*r -3.0 -2.58 All rod bank wort h (ll7.P),

ah/h 10.0 8.30 1 Initial boron cone (l!I P), ppm 1400 1070 Boron re.ctivit'. worth, 70F, 75 81 ppn/1% t.h/h Max. ej ec ted rod worth (HFP),

• ak/k 0.65 0.55 Dropped rod worth (l!PP), ak/k 0.46 0.20 (a) -

-1.2 x 10,5 ak/k/F was used for stean-line failure analysis.

-1.3 x 10 " tk/k/F was used for cold water analysis.

(b)+0.94 x 10~'+ ak/k/F was used for the nodcrator dilution accident.

7-3 Babcock 8. Wilcox

Amendment 1 (6/15/77)

Fir;ure 8-2, Cure Pro tec t i en S.if e ty Li:ait.s, Oconee 2, Cycle 3

. OF UTED rhEP"1L PDER CNSR t I"'l - ., 123

1. 112)

. 33 6.ii2) ,

(~1 3.112)

~I j aC .i ..t 4 T' .' P g

• FT Llul f l

r e ti Listi . I 100 \ (33.102)

I i l

l I

( '0

, H) l ,

l I 1

@)

4 95 31 l l I I l

- B ,,'a

/l?%rrit:E3 g , p . ...,,

Dri p;; i a ;

I I

(75 3) j I

' l . 7 t. l 1

(CE 3, l g 1 l l n l l l

1 Q,,"d; . St> ISS 2)l l l l attt Pit;L E I I I l 2. 3 & 4 FLvP l l 1

-~

5' (46 ?>

OPER:Il01 l

! l l

(39 2) -- 40 l l

[

1 I I I I

- 30 l l

l I l I ' I 29 i i l I i I i I f -- 10 l l l

l l l l u I t i i il ii i . .

60 50 40 -30 20 10 0 10 20 30 40 50 50 Anal Po er Imualance, i CURVE REaCIOR CCCLA',r FLO* (CP4i I 374 ff0 2 720 035 3 183.C]O 4 204 310 T*E ILU? Ft04 SETFS.f,I FOR 2 0 OPERATIM, W ST BE O 943 l

l l

8-3 Babcock 8. Wiicox

Araendmen t 1 (6/15/77)

Figure 8-5. Prot ec tive syst em :Ir.imum A11cwable Setpuints, Oconce . ' , Cycle 3 THERv8L POWER LivEL.

U I. 3 C C E P T A 3 L C O l' E R A T I C )

IIC

(-23.1.805 5). f(125.5) . . , . _ , .

(11.3.105.5)

I I H : . 897

~ '

u,=i,259 l l

ACCEPTABLE l (23.95)

,37,, )

l4 g

PUNP - 00 e P E R .9110 h l I (73.1) .. FC l

\

l .i

<(23.68.3) l A c c ! P ' t !' L E I

(.3t 6 8 . 3 ; '

l3 & e t. n p

_ 60 l l opt

• Alich g l (58.7) g, i- 33 l l l _ go l -

(23.41.2) l A L C EP I A llL C l

(-37,3 4. 2 )

l2.3, I & 4 _ 30 l l PUMP OPIkA110NU l

- 20 9 I

-I

- 10 l- a S

n' li l ,2 m c< l l f I f*l!  !

ln II " l 1 l

-60 -50 .u0 -30 -20 -10 0 10 20 30 40 Power imbalance. '

I *THE FLUX FLO'll SETPOINT FGR 2'O FUJ.' OPE.*JTION f.!UST EC SET AT 0.949 l

8-6 Echcock W.4 Ice" e

Amendment 1 (6/15/77)

Figure 8-6. Rod Position Limits f or Ecur-Pwap Opeation From 0 to 100 1 10 EFPD, Oconee 2, cycle 3

' e225.B.10D r> U : ?t 0' P. T9 t S ' 3 7# 2 IC2 ' <- o j ,' ,' ~

*ie ' - te j; DESI

tCTED

.',[E,

FISTRitiED tia 2 32, # m b S26 ,3 REGICN p;g i ;p, FGhER LEuEL Sc _ ,

i IGO 2.50 i i 233 8 E0, Sw)703nN .-

- "ARG!N,LIMli < (146 2 Tij (253.8 70)

[ 60 -

(132 2.008 t;G7.8.00)

E

-

(45.50) (118 2 301 P E U?l SSI BL E f2C1 8.50)

OPETAfi'.G 4g AfGION f _.

g (300.37s i 20 y40 15) <70 15) 0 - t , i i , e i , , , ,

0 50 100 150 200 250 300 Rod Innes. elinartan 0 25 50 75 100 0 25 50 75 100 1 I t f f f f i f }

Group 5 G r ct:p 7 0 25 53 75 100 1 0 1 t e Group G 8-7 Babccck & Wi!ax

l l Amendment 1 (6/15/77)

Fi;;ure 8-7. Rod Posit ion I.imits f or Four-Pu.ap Operation From 100 t 10 to 250 : 101:Ft>D, oconeo 2, Cycle 3 t i ; t 10 2 , ,, , 223.8,132)

IC3 -

c l a u t o . U; N 6 PECICN 15 f)! ill'inE0 f REM O C MD SEGION te74.2,0/>

. '225.0,02)

' PC* LR Lis'EL SM'icus vaRGIN C'JT0f f 50 -

Lieli ._ ___, f 16J.2,50) r239 8.60)

I

/

_ i!4C.2,70) 253.8,70)

~

RFSTRfCTED f

a 60 -

RE G U;'a

• $132.2.60, (257 S.00)

_ (100 501 'l18.2,50)

(281 8,50 )

40 -

j (300,37, .

PERTISSIELE OP! R AI i t.;

gg REGION (56 ;5, (70,15) a f ___l t .  ! '

! f f f  ? 9 0 50 100 150 200 250 300 Rou inces, titnnra n 0 2'. 50 75 100 0 25 50 75 100 i i , , i , , , , ,

Group 5 Group 7 0 25 50 75 100 3

Group 6 1

l l

l t

l

- l l

8-8 Babcock & Wilcox

. o Amendment 1 (6/15/77)

Figure S-8. Red Positlen 1.imits for Four-Pump Oper:ttion Mter 250 A 10 El PD, Oconee 2, Cycle 3 i;g ,

'IG5.102s =

o(25l 6.102i

'~- (251 G,92).

CPErailG** ts inl5 i I Ritt0N IS 'df SLL0=[: CufGFF i

~

80 _ (237 C.20) s ;a;;,% 4

. . v,gis f223 6.70)

likli *EifaiCTED REGifA 2C3 G,60)

[ CD -

, =

. . 2t so , i195.0 50 s

'l i I 40 -

PERVISSIBLE OPERATING i ION (69 15)

<145.i5)

Y 0' ' ' ' ' ' ' ' ' ' ' '

O 50 100 t50 200 250 300 RJJ I ,itle s W i lli'J f J a" 0 25 50 4 100 .0 25 50 75 tCD

' ' ' ' f f 1 i I  :

Grouw 5 Group 7

! o 25 50 75 100

! i , . , ,

3 Group 6 J

d i

i i

e 4

I 8-9 Babcock 8.V!kox

.h ndment 1 (6/15/77) 4 Figure 8-9. Rod Position 1.imits for Two- and Three-Pump Operation Fron 0 to 100 2 10 ErPD, Oconce 2, Cycic 3

.05.102) '132.2 102)

( 233 3,102) igg ,

,cro 7 '100 2,102; OP E R .',i t C s . ?. fels  %.c RESTRICIED TEGIO

. IS '.0 f '.s 2 f. '3 > P.'?

PEGION FOR 3 e . AtLL*UJ rite 2 C4 3 Pl" yg rie. 414C 2.PG, (253 8.89) pg.jp PJ'- 7 5 JEU it'.

U ;f.:Tri1CTEC OPERATION I} LO -

I'4 .'els a gget. (132 2,70) (257.8,76) 9 .

,, u l N t 'et*

= SHJT Ger 63

. L 1::l T [

j ~

~:45.50 ) T Ei >'I S S i gt ! (200.47) i 5 CP!4Hiras 40 .

.=.

a REG 10'.

U.,

d. 20 .

(0.15) 0 ,  ;  ! t . , t t , y 0 50 100 . 150 200 250 300 Rou inces, estntitaan 0 25 50 75 100 0 25 50 75 100 i f  ? f I l l f f I Group 5 Group 7 0

I 25 f

50 f f 75  :::f Grouu 6 1

1

~

8-10 Babcock & Wilcox 1

I l

Amendment 1 (6/15/77) 1' Figure 8-10. Itod Positten Liuits for Two- and Three-Ptap Operation

From 100 : 10 to 230 : 10 F.FPD, Oconec 2, Cycle 3 l

'33' 'M v233 S 102) i 100 -

0H RAfth a Twt 3 RESTRIOTED

& Rf 010N 15 '/); Au r;E REGION FGR (253 8.63) 3 PU"?

OPERATION j

S3 -

sne!0c.m nuco. ( 20 7. a. 76) 2 s Livi; ---

, g U

, a E3 -

+

0 -

100 50)

(300,47) 40 _

i  ; P E R'4l Sil et E CFleATINC

~

- DEGION 1

70 -

55 15) d 0 ' ' t '

i , , , , ,

0 50 10D 150 200 250 300 l R3'I Inder. #itndra.n 0 25 50 75 100 0 25 50 75 100

' ' ' ' i i , e , j Group 5 groug 1 0 25 50 75 100 I ' f f f.

Group 6 I

1 1

1 l

l l

s 8-11 Babcock a. Ylilcox -

t

~

, u -

, o Amendraent 1 (6/15/77)

FI;;ure 8-11. Rod Position Limits for Two- and Thr:e-Pump Operation After 250 1 10 EFPD, Oconce 2, Cycic 3 0

1 CP[Ea t ' $'. lf; Till $

REGict. is *GI

. CE;falCTED kEGIC'. i E p ..,!. , FOT, 2 f, 3 o c/ 'P d.p3;;ty 2 '223 G II) tc .

muTic.

/. eswa it. '..l; 5 N' u / ' (203 6.761 s

G 5 e.!liM 4'.

93 GIN L 191T~

E r.c -

" /

i l 21. 50 ) PERMISSIBLE CPERATING 7 Rf GICt.

a

,2 40 -

t j 20 .

t09'155

/

C s e i , , , , , , , , ,

0 50 100 150 200 250 300 Ro:i ince: Witnoraen 0 25 50 75 100 0 25 50 75 100 E i I I I f f

f g g Group 5 Group 7 0 25 50 75 100 1 f f  ? e Group G I

I I

! 8-12 Babcock & Wilcox l

I t

e e6 A:nendraen t 1 (6/15/77)

Figure R-13. Operctrional Power labalance Envelope for Operation F r or, 100 1 10 to 250 : 10 EFPD, Ocenee 2, Cycle 3 P o w e r , ', o f 2508 i,"ll t RESTRICTED I;EG10l4

(-20.5.102) 5 --

(+14:2,102) 100-(-19.0.02) c

( +13. 4,92 )

90 --

80 - -

70 -

SD --

50 --

40 --

30 --

20 --

10 --

0

-1 -

-30 20 -0 0 410 +20 Care Imaalance, G l

i l

8-14 Babcock & V!itecx

-, m.,

e s Amendment 1 (6/15/77)

Fi;;itre 8-14. Opera tional Power Imbalance Envelope for Operation After 250 10 EFFD, Oceace 2, Cycle 3

Power, 5 of 2503 " t 4

RESTRICTED REGlot

( +11.1,10 2 )

( 2G.7.102) - - __

100 - _!

(-24.5,92) . ( +10. 7,92) 90 80 -

70 -

1 60 -

50 _

40 -

30 20 -

10 - ,

1 e . . 0 ,[ ,

-30 -20 10 0 +10 +20

- Core imoalance, "4 l

l

I 8-15 . D3bcock & Wilcox f

,- S Amendment. 1 (6/15/77)

Figure 8-16. APSR Position Limitu for Oper:ttion After 230 10 EFPD, Oconce 2 Cyclo 3 RESTF.iCTED RECIC:.

100 ~

':6.5.102) e (31.4.102) RESTRICTED

./ \ REGl0ti l \o(33.8,92) l 90 '(4.l.02) j r

/  :

80 .'lI 7.En)

,, ( 33. 8, sil )

1 1g , (0.70) t(33.0,70)

N.'

3 CD -

(84.8.60)

(103.60) i  ; 50 -

40 "

PER!.!I SSI BL E f OPERATitJG 30 REGI0tt 20 -

1 10 -

t 0 s .t. ._ . , , , ,

0 10 20 30 40 50 00 70 00 00 100 APSR '. t!s tnd raen 8-17 C. *.ccck & Wi!cox

-t

,,