ML19322B866

Cycle 3 Reload Rept.

ML19322B866
Person / Time
Site:	Oconee
Issue date:	12/31/1975
From:	BABCOCK & WILCOX CO.
To:
References
BAW-1427, NUDOCS 7912060696
Download: ML19322B866 (54)

Text

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3'4 267

[ 73 BAW-1427 Decerber 1975

' ;/ de.,4%'s/ gy /h i

///6I i 2603 OCONEE UNIT 1 l CYCLE 3 RELOAD REPORT 1

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Babcock &Wilcox

$/%M 1912060h9$

BAW-1427 Decuter 1975 i

i OCONEE UNIT 1 CYCLE 3 RELOAD REPORT i

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i BABC0CK & WILCOX Power Generation Group Nuclear Power Generation Division P. O. Box 1260 Lynchburg, Virginia 24505 Babcock & WNeon

. . , . _ . _ __ . , _ _ . . . _ . ~ . ~ , - . . - . - _ . . . _ - . , . , , - - . _ . , - - , - . . . . - - - . - . , , . , , , . - . , - -

CONTENTS P.s ce

1. INTRODUCTION AND

SUMMARY

. . . . . . . . . . . . . . . . . . . . 1-1

2. OPERATING llISTORY . . . . . . . . . . . . . . . . . . . . .. . 2-1

3. CENERAL DESCRIPTION . . . . . . . . . . . . . . .. . . .. . . 1- 1 4 EUEL SYSTEM DESIGN . . . . . . . .. .. . . . . . . . ... . . 4-1 4.1 Fuel Assembly Mechanical Design . . .. .. .. . . . . . . 4-1 4.2 Fuel Rad Design . . . . . . . . . . . .. . . .. . . . . . 4-1 4.3 Thermal Design . . . . . . . . . . . . . . . . . . .. . . 4- 3 4.4 Material Design . . . . . . . . . .. . . . . . . . .. . . 4-4 4.5 Operating Experiences . . . . . . . . . . . . . . . . .. . 4-4

5. NUCLEAR DESIGN . . . . . . . . . . . .. . . . . .. . . . . . . 5-1 5.1 Physics Characteristics . . . . .. . . . . . . . . . . . . 5-1 5.2 Analytical Input . . . . . .. . . . . . . . . . . . . . . 5-2

5. 3 Changes in Nuclear Design . . . . . . . . . . . . . . . . . 5-2

6. TilERMAI.-liYDRAULIC DESIGN . . . . . . . . . . . . . . . . . . . . 6-1 6.1 Thermal-Ilydraulic Design Calculations . . . . . . . . . . . 6-1 6.2 DNBR Analysis . . . . . . . . . . . . . . . . . . . . . . 6-1

7. ACCIDENT AND TRANSIENT ANALYSIS . . . . . . . . . . . . . . . . 7-1 7.1 General Safety Analysis . . . . . . . . . . . .. .. . . . 7-1 7.2 Rod Withdrawal Accidents . . . . . . . . . . . . . .. . . 7-1

7. 3 Moderator Dilution Accident . .. . . . . . . .. .. . . . 7-2 j

7.4 Cold Water (Pump Startup) Accident . . . . .. . . . . . . 7- 3 7.5 Loss of Coolant Flow . . . . . . . . . . . . .. .. . . . 7-3 7.6 Stuck-Out. Stuck-In, or Dropped Control Rod Accident . . . 7-4 7.7 Loss of Electric Power . . . . . . . . . . . . . .. .. . 7-4 7.8 Steam t.ine Failure . . . . . . . . . . . . . . . . .. . . 7-5 7.9 Steam Generator Tube Failure . . . . . . . .. . . . . . . 7-5

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CONTENTS (Continued) i'aze

!.1J Fuel Handling Accident . . .... . . . . . . . . . . . . 7-3 7.11 Rad Election Accident . .. . .. . . . . . . . . . . . . 7- 5

~ . I .' niximum Hypothetical Accident ... . . . . . . . . . . . 7-6 7.13 '*iste

.. Cas Tank Rupture . ... .. . . . . . . . . . . . . 7-6 7.14 LOCA Analysis . . . . . .. .. . . . . . . . . . . . . . 7-6 M. TROPOSED MODIFICATIONS TO TECllNICAL SPECIFICATIONS . . . . . . . 3-1

9. S TART-L'P PROGRAM . . . . .. . .... . . . . . . . . . . . . . 9-1

10. REFERENCES . . . . . . . .. . .. .. . . . . . . . . . . . . . 10-1 List of Tables Table 1

I a.1-1 Fuel Design Parameters . .... . . . . . . . . . . . . . 4-5 4.2-1 Fuel Rod Dimensions . . . ... . . . . . . . . . . . . . 4-5 l 4.2-2 Input Summary for Cladding Creep Collapse Calculations . . 4-6 l

4. }-1 Densified Fuel Temperature Analysis Parameters for Cycles 2 and 3 . . . . .. ......... . . . . . . 4-7

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5. t- 1 Cycle 2 and Cycle 3 Physics Parameters . . . . . . . . . . 5-4 5.1-2 Shutdown Margin Calculation Oconee 1 Cycle 3 . . . . . . 5-5 6.2-1 Cycle 2 and Cycle 3 Design Conditions . . . . . . . . . . 6-3 l 7.1-1 Comparison of Key Parameters for Accident Analysis . . . . 7-7 7.1*-1 Allowable LOCA Peak Linear Heat Rate . . . . . . . . . . . 7-8 i l

List of Figures ,

Figure '

1- 1 Oconee 1. Cycle 3 Core Loading Diagram . . . . . . . . . . }- 3 3-2 Oconee 1 Enrichment and Burnup Distribution for Cycle 3 . }-4 i

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_ List of Figures (Continued)

Flaure Page F- 1 skonce 1. Cycle } Cont rol Rod Locations Bef ore Interchange

. . . . . . . . . . . . . . . . . . . . . . . 3- 5 3- 4 Oconee 1. Cycle 3 Control Rod Locations After Interchange

. . . . . . . . . . . . . . . . . . . . . . . }-o

4. 3-1 Maximum Cap Size Vs Axial Position . . . . . . . . . . . . 4-8

4. 4-2 Power Spike Factor Vs Axial _ Position . . . . . . . . . . . 4-9 5.1- 1 BOC (4 EFPD) C)ile 3 'No-Dimensional Relative Power Distribution . . . . . . . . .. . . . . . . . . . . . .. 5- 7 S-1 Core Protection Safety Limit

... . . . . . . . . . . . . 8-2 6-2 Core Protection Safety Limits . . . . . . . . . . . . . .

S-3 5- 3 Protection System Maximum Allowable Setpoints . . . . . . 8-4 6- 4 Protection System Maximum Allowable Setpoints . . . . . . 3-4 o- 5 Rod Position Limits for 4 Pump operation Applicable to the Period From 0 to 230 t 5 EFPD . . . . . . . . . . .. 8-b 5-4 Rod Position Limits for 4 Pump Operation Applicable to the Period Af ter 23015 EFPD . . . . . . . . . . . . .. S-7 S- 7 Rod Position Limits for 2 and 3 Pump Operation Applicable t o t he Period f rom 0 to 2 30 1 5 EFPD . . . . . . . . . . . 3-8 d-8 Rod Posit ion Limits for 2 and 3 Pump operation Applicable to the Period Af ter 23015 EFPD . . . . . . . . . . . . . 8-9 d-9 Operational Power Irbalance Envelope for Operation From 0 to 230 t 5 EFPD . . . . . . . . . . . . . . . . . . .. 8-10 o-10 Operational Power Imbalance Envelope for Operation After 230 1 5 EFPD . . . . . . . .. . . . . . . . . . . . . . . 6-11 6-11 LOCA Limited Maximum Allowable Linear Heat Rate . . . . . 8-12 iv -

1. INTRODUCTION AND S'M.ARY This report justifies the operation of the third cycle of Oconee Nucicar Station, Unit I, at the rated core power of 2568 T4t. Included are the required analyses, as outlined in the USNRC document " Guidance for Proposed License Amendments Relating to Refueling" June 1975.

To support cycle 3 operation of the Oconce Nuclear Station, Unit I, this report c= ploys analytical techniques and design bases established in re ports which were previously submitted and accepted by the USNRC and its predece4sor (>ee ref.rences).

A brief summary of cycle 1 and cycle 3 reactor parameters that are related to power capability is included in Section 7 of this report.

All of the accidents analyzed in the FSAR have been reviewed for cycle 3 operation. In those cases where cycle 3 characteristics proved to be con-servative with respect to those analyzed for cycle 1 operation, no new analysis was perforned.

The Technical Specifications have been reviewed and the modifications required for cycle 3 operation are justified in this report.

Based on the analyses performed. which take into account the postu-lated effects of fuel densification and the Final Acceptance Criteria for l'nergencv Core Cooling Sv* tens, it haa been cencludsd that Oconec thit 1 Evele 1. can be safely operated at the rated power level of 2568 5't.

1-1

2. OPERATI;G llISTORY t he- ref erence cycle for the nuclear and thermal hydraulic analyses of ocen..e ?;ucle.ar Station Unit I i=i the presently operating cycle 2. Cvele 2 pw.cr c.c.'itto roceenced on March II,1975, following the complet ion of the 7ero pow.r physics testing. The rated power level of 2568 '"4t was achieved on

.T;> r i l 11, 1975. A control rod interchange was performed at 53 effective full poser days (1:FPD). The design fuca cycle of 29D EFPD 13. scheduled for rompletion in .lanuary of 1976 2;o operating anomalics occurred during the 1-second cycle which would adversely affect the fuel performance during the thir.1 cycle.

The nuclear and thermal-hyd7aulie analyses of cycle 2 utilized the B..-2 critical heat flux correlation and the me.asured core tiow. The cycle )

.an.ilvu s also enployed these features vliich have the combined or singul.or effect of increasing margin to D:;B.

Operatlon of cvele 3 is scheduled to begin in March of 1976 The d.si,n evcle length is 292 EFPD and one control rod interchange is planned

.i t 100 t 10 FFI'n.

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3. GTJ;ERAL LESCRIPTION The Oconee Unit I reactor core is described in detail in sect toti l of the Oconee Nuclear Station. I' nit I . Tin.n l Sa f e t y An.s ! . s i s deport ( c:ce ..

1).

The cycle 3 core consists of 177 fuel assemblics, each of .hich is a 15 by 15 array containing 208 fuel rods.16 control rod guide tubes. and one incore instrument guide tube. The fuel pin cladding is rold-worked 7.irc.elov-4 with an 03 of 0.430 inch and a wall thickness of 0.0265 inch. The fac 1 consists of dished end cylindrical pellets of uranium dioxide which are 0.700 inch in length and 0.370 inch in diameter. (See Table 4.1-1 and Table 4.2-1 for additional data.) The fuel assemblics in batch 3 have an average nominal f uel loading of 463.6 kg of uranium whereas the batch 4 an<i ~ assem-blies maintain an average nominal fuel loading of 463.6 kg of uranius. Ti.e undensified nominal active fuel lengths and theoretical densit ies also sary between batches and are presented in Tahics 4.1-1 and 4.2-1.

Figure 3-1 is the core loading diagram for (konee Unit 1. Cycle 1.

The init ial enrichments of batches 3. 4A and 45 were 2.15, 2.60 and 1.20 wt '

2 35 t '. respect ively. Batch 5 is enriched to 2.75 wt -

235'

t. All of the 5..t c h 2 a::semblies and 24 of the hatch 3 assemblies will be disch.irged at the end of cycle 2.

The remainder of batch 3 assemblies and the batch 4A ar.d 4B assembiles w!!! be shuf fled to new locations at the beginning of cycle 1.

f resh :.. tch 5 assemblics will occupy primarily the periphery of t he core a  :

4 m.slor axes posit ions slightly interior to the core. Figure 3-2 is an e4 Y core map showing the assembly burnup and enrichment distribution at the b. ,

of cycle 3.

Reactivity control is supplied by 61 f ull-length Ag-In-Cd ecntr. . r

.ad soluble boron shim. In addition to the full-length control rods, eigh; axi.

power shaping rods are provided for additional control of the axial ; .... t d i st r ibut tor.. The cycle 3 locations of the 69 control rods and the group designaticas are indicated in Fagures 3-3 and 3-4 The core locations of tlic total pattern (69 control rods) for cycle 3 are identical to those of the reference cycle indicated in the Ge onee I. Cycle 2 Reload Report (reference 2).

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The group designations, however, differ between cycic 3 .ind the refsrem u cycle in order to nininire power peaking. One control rod int e r e n.in re is pl. inned at 100 + 10 EFPD.

The nominal ta r. tem pra'sstire i n 2200 ps i.e .md t h. Jen s i f i. .' r.r'ni n s i he.it ra t s- is 5.78 kw/ft at the rated core power of 2568 .Wt .

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i i .; t.r e 3-1 tesnee 1. Cycle 3 FUEL. TRANSFER Cere Loa < ling Diagras CANAL k X

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5 5 5 5 5 4

5 5 5 3 3 3 3 i i ~>

3-7 G-8 B-9 0

i 3 43 '6

. 5 4h 43 3 5 5 C-5 B-4 3-5 B-Il S-12 C-Il i 5 43 3 3 43 3 4B 3 i 3 3 ad 5 5 g 11 - 3 C-4 I)-6 A-7 B-8 A-9 D-10 C-12 C-i 5 3 3 LB 43 4B '

4B 45 43 ah 3 l 3 5 E-3 li- 3 E-5 A-6 3-6 A-8 S-10 A-10 E-11 lr-13lE-13 5 5 43 3 4B 3 3 4A 3 I 3 4B 3 43 5 D-2 F-4 F-1 T-6 F-7 E-8 F-9 F-lO F-15 F-12  ;-14

, S 3 48 4h 4B 3 4B 48 i

' 3 3 4h 4B  ; eB 3 G-2 E-2 C-1 F-2 C-6 C-3 F-8 i

C-13 G-lO F-14 G-1 1 3 E-la G-14 5 3 5 3 4B 4A 3 4A 3 4A 43 3  ;

11 - 7 3

.g 11 - 2 11- 1 11 - 5 H-6  !!-8 li-10 H-il H-13 H-la 11 '1 5 3 4B 4B 45 3 4B 3 43 4B K-2 M-2 3 '. B 4's 3 K-1 L-2 K-6 0-3 L-8 n-13 K-10

- L-14 K-15I" ~~. K-14 i i 4h 3 4B 3 3 4A L 3 3 43 3 '- 5 N-2 L-4 L-1 L-6 L-7 M-8 L-9 L-10  !.- 15 I.-12  ;-14 i 3 3 43 43 M 43 4B 45 45 4B 3 M-3 3 $

5-3 M-5 R-6 P-6 R-8 P-10 R-10 i 1

1-! ! S-13 "-li i 5 5 43 0-8 3

0-4 3

N-6 4B R-7 P-8 3 4B 3 l3 48 3 5 k-9 N-lO 0-12 H-13 3 5 3, OB 43 3

5 4B 45 3 5 3 0-5 P-4 P-5 P-11 P-12 0-11 9 i

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, 5 5 5 3 3 3 5 5 5 I P-7 K-8 P-9 e

R 5 5 5 5 5 I

Z l 2 3 4 5 6 7 8 9 10 : 11 12 13 14 l' Batch Location Cycle 2 Core Location

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"IGU:=E 3-2 CCO:'EE I F_*;RICHMF.'.T AND 1,*1';Ui' DISTRIBC:0N FOR CYCLE 3 l

l 8 9 10 11 17 13 14 15 2...+ .' .1 5 2 . *>0 3.20 2.13 2.75 2.I3 2.75 l H 12152 13460 10877 9612 153 % 0 l'.357 0 3.20 2.15 3.20 3.20 3.20 2.15 2.75 E IF245 0 76 % 15755 11913 9330 10349 2.15 3.20

• 2.15 3.20 2.75 2.75 L

15788 7974 12B03 7273 0 0 3.20 2.15 2.15 2.75 M

13145 12720 16419 0 3.20 2.75 2.75 N

11010 0 0 2.75 0

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R EG Initial Enr.cle.ent XX:G 150C Burnup (L'D/MTC) 3-4

Figure 3-3 Ocons e 1, Cycic 3 Control Rod Locations Before *nterenange X

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5 3 2 3 C 4 7 7 i4 D 6 8 8 I 5 6

. l E 4 5 1 1 5 4 F 3 8 2 6 i  !

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8 3l C 6 7 1 3 3l 1 7l W-

!s 2 5 6 4 4 i

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1 3 3 1l 7 L 3 8 2 2 8 3 M 4 5 1 '

1 5 t.

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1 2 3 4 5 6 7 8 9 t 10 11 g 12 l 13 14 .. 5j Group Nunber of Rods Function 1 8 Safety 2 N Safety 3 12 Safety 4 9 Safety 5 8 Control 6 9 Control 7 8 2ontrol 8 8 APSR's

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Figure 3-4 Ocenee 1, cycle 3 control Rod Locations After Interchange X

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B d 3 6 C 2 5 5 2 D 7 8 7 8 7 E 2 5 1 1 5 2 F 6 8 4 4 4 8 6 C 5 1 3 3 1 5 W- 3 7 4 4 4 7 -Y H 3 K 5 1 3 3 1 3 L t' 8 4 4 4 8 6 y, 2 5 1 1 3 2 N 7 8 7 8 l7 y 0 2 5 5 2 ,

p 6 3 6 R

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1 2 3 4 5 6 7 8 9 10 11 12 13 14 ?5 i

Croup Nur.ber of Rods tuaction 1 S Safety X Group Number 2 S Safety 3 S Safety 4 9 Safety 5 12 Contr01 6

I 8 Control 7 3 Control 8 8 APSR',

3-6 TOTA 1. 69

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4. FUEL SYSTEM DESIGN 4.1 Fuel Assethly Mechanical Design Pertinent fuel design paramaters are listed in Table 4.1-1. All fuel asseaLifes are identical in concept and are mechanically interchangeable.

The new fuel assesblics incorporate minor modifications to the end fitt ings, primarily to reduce fuel assembly pressure drop and to increase holddown nargin.

All other results presented in the FSAR fuel assenbly mechanical discussion are applicable to the reload fuel assemblics.

i.2 Fuel !iod Desien Pertinent f uel rod dimensions for residual and new fuel are listed in Table 1.2-1.

The mechanical evaluation of the fuel rod is discussed below.

flajding Collapse:

.reep collapse analyses were performed for three-cycle assembly power histories for Oconee 1.

The batch i fuel is more limiting than batch a fuel due to the lever prepressurization ar.. Iower pellet density. A su - ry of the batches 3. 1. and 5 fuel ro1 designs is contained in Table 4.2-2.

The

b. itch 3 assensiv power histories were analyzed and the most limiting assembly for cvele 3 was determined. The predicted assembly power history for the nost limit ing assembly was used to determine the most limiting collapse ti=e as described in Ef.'-1003;P-A (reference 3). Measured power distribution dat.:

obtained during cycle 2 operation confirmed the accuracy of the cycle 2 design calculat ions used for t he collapse analysts.

The conservatisms in the analytical procedure are summarized below.

1.

The CROV computer code v.ss used to predict the time to collap:e. CROV conservit is ely predicts collapse times, as demonstrated in reference 3.

2.

No credit is taken for fission g.is release. Therefore, the net dif ferential pressures used in the analysis are conservatively high.

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3. The cladding thickness used was the LTL (lower tolerance limit) of the as-built measurements. The initial ovality of the cladding used was the. UTL (upper tolerance limit) of the as-built measurements. These values were taken

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trom a statistical sampling of the cladding.

4 Batch 3 cladding temperatures were calculated using assembly outlet temperatures. This results in cladding temperatures which are conservatively high when combined with the maximum axial peak.

The most limiting assembly was found to have a collapse time greater than the maximum pr,ojected cycle 3 life of 21,500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> (see Table 4.1-1). This analysis was performed using the assumptions on densification described in reference 3.

Cladding Stress:

Since the batch 3 fuel is the most limiting from a cladding stress point of view due to the low prepressurization and low density, the calcul.itions pertormed in the Oconce I Fuel Densification Report. (refe ence 4).

are the rust 1initing.

fuel pellet Irradiation Swelling:

The fuel design criteria specify a limit of 1.07. on cladding circum-ferential plastic strain. The pellet design is set such that the plastic cladding st rain is less than l' at 55,000 MWD /MTU. The conservatisms in this analysis are listed below.

1

1. The maximum specification value for the fuel pellet diameter 1 was used.

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

3. The cladding IC used was the lowest permitted specification tolerance.

4. The maximum expected three cycle local pellet burnup is l less than 55,000 MWD /MTU. l l

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4.3 Therm al Design The core loading for cycle 3 operation is shown in Figure 3-1. There

.ere n0 fresh (batch 5) fuel assemblies. 61 once-burned (batches 4a and 4B) assemblies and 56 twice-burned (batch 3) fuel assemblics. These asse=blies are thermally and geometrically sinflar. Limitations on the linear heat rate were established ut ilizing full fuel densification penalties. This results in a minimum linear heat rate capability of 20.15 kv/ft.

Fower Spike Mode 1_

The power spike model utilized in this analysis is identical to that presented in BAW-100555 except for two modifications. The modifications have been applied to Fg and Fk. 6 These probabilities have been changed to reflect add!tional data from operating reactors that support a somewhat different approach and yield less severe penalties due to power spikes. F was g changed from 1.0 to 0.5. Fk was changed from a Caussian distribution to a linear distribution. which reflects a decreasing frequency with increasing gap size.

The power spike and maximum gap size have been calculated both for ba t e t- 4 and batch 5 fuel. The maximum gap size versus axial position is shown for both batches in Figure 4.3-1 and the power spike factor versus axial length is shown in Figure 4.3-2.

For those analyses where centerline fuel melt is limiting, the higher power spike of batch 5 fuel has been used; however, for DNBR analyses (Section 6.2), the batch a power spike has been used. The factor, when combined with the shorter act ive length of batch 4 fuel, results in the worst DNBR densif ication penalt y.

Fuel Temperature Analysis Thernal analysis of the fuel rods assumed in-reactor fuel densifi-e.ition to 96.57. theoretical density (TDF). The basis for the analysis till:cd is given in BAV-10055 and BAW-100447 with the following codifications:

1. The opt ion in the code for no restructuring of f uel has been used in the .inalysis presented here in accordance with the NRC interim evaluation of TAFY.

2. The calculated gap conductance was reduced by 25? in accord-ance with the NRC interim evaluation of TAFY.

4-3

During Cycle 3 operation, the highest relative assembly power levcis occur in batches 4 and 5 fuel. Fuel temperature analysis for batches 1. 2, and 3 fuel is documented in the Oconce 1 Fuel Densification Report. This analysis is also applicable to batches 4 and 5 because they have the same linear heat rJte capabilitleS to centerline melt as batches I, 2 and 3 (26.15 kv/ft). The maximum hot spot centerline fuel tcmperature is predicted on .he basis of the reference design peaking conditions as shown in Table 4.3-1.

4.4 Matyrial Design The batch 5 fuel assemblies are not new in concept and they do not utilfre different component materials. The refore , the chemical compatibility of all possible fuel-cladding-coolant-assembly interactions for the batch 5 tuel assenblies are identical to those of the present fuel.

4 . _5 Operating Experiences B5W's operating experience has been demonstrated in the operation of six 177 fuel assembly plants utilizing this fuel assembly design.

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Table 4.1-1. Fuel Design Parameters Residual New Fuel Assemoly Fuel Assembly Batch 3 Batch 4 Batch 5

1. Fuel Assec.bly Type Mk-B2 Mk-33 Mk-B4

2. SunSer 56 6L 60

3. Initial Fuel Enrichment 2.15 3.20/2.60 2.75

4. Initial Fuel Density, 93.5 > 94.5 93.5

% Theoretical

5. Initial Fill Gas Pressure (Minimum specified), psia * * *

6. Batch Burnup, BOL, MWD /MTU 15076 9798 0 7 Clad Co!! apse Ti w , Effective > 26,000 > 30,000** > 26,000**

Ful1 Power Hours Table 4.2-1. Fuel Rod Dimensions Residual Fuel New Fuel Assembly Ccno nent Batch 3 Batch 4 Batch 5 ,,

1. Fuel Rods 0.D. I nches .4 30 .4 30 .430 I .D. I nches .377 .377 .377

2. Fuel Pellet o.D. Inches .370 .3685 (mean) .370 Density, % Theoretical 93.5 > 94.5 93.5

3. L'ndensi fied Active Fuel Length, inches 144 142 142.6 4 Flexible Spacers. Type Corrugated Spri ng Spring Spacer

5. Solid Spacers, Material Zr0 Zr-4 Zr-4 2

• PROPRIETARY

• • A conservative power history envelope was t. sed for bat ches a and 5 rat her than spectsic histories.

4-5

Taste 4.2-2. I rpu t _ _S umma ry for_ Cladding Creep Colla 2se Calculations-Batch 3 Batch 4, Bitch $

Pellet 03 (scan specified), in. .3700 .3685 .3700 Pellet Dec.Sity (cwas specified).

~ TD 93.5 94.5 93.5

ensified Pellet OD. in. .3663 .3661 .3663 Cladding ID (mean specified), in. .377 .377 .377 Cladding Ovality. (L*TL) . in. * *

• Cladding Thickness (LTL). in. * * *

"repressure (minimum specified), psia * *

• Post Densification Prepressure (cold), psia * *

• Reactor System Pressure, psia 2200 2200 2200 Stack lie f ght (undensified), in. 144 142 142.6

• PROPRIETARY w

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TABLE 4. 3-1 DENS!FIED FL*EL TEMPERATt!RE ANALYSIS P_ ARA.SETERS TT)R CYCLES 2 AND 3 Ecietor Core Power Level MW t 2568 Sy. stem Pressure, psia 2200 Reactor Vessel Coolant Temperature, F 579 Fraction of Ileat Cencrated in Fuel and Cladding .973 F . ,g 1.78 Fg 1.70 Fq (Nuclear) 3.03 Fq (Nuclear and Mechanical) 3.12 Average Therent Output kw/f t - Batch 3 5.742 Batch 4 5.805 Batch 5 5.799 Average Fuel Temperature. F 1350

'taximum Fuel Centerline Temperature at I ot Spot. F 4710 Densified Active Fuel Length, in. - Batch 3 141.S4 Batch 4 140.30  !

Batch 5 140.46 Linear !! cat Rate to Central Fuel Melt, kw/ft 20.15 l Initial Theoretical Density (TDI) - Batches 3, 5 93.5 l

t Batch 4 95.5 l

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l 4-7 l

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3.0 TOF = 96.5

,3 20 ,_

, i O f ej g\$9 ' g,gs .S' * )

' f j f pit? gatt"

! '8'nieus Let Average censity 00 l l I i 1 l

- 20 40 60 33 300 120 140 Areal P058t#0ft, intfit s 1

NAIINUN GAP $32[ ys AglAL P053fl0N Figure 4.3 1 4-6

-- _._ , _ -v-- _

I.09 TOF = 96.5%

1.07 -

1.06 -

9.*

3 1*05 8 **

~

\

C g0 2

%$g+% of 1.04 -

a m #

3 1.03 -

s' l.02 -

I.01 - Nansmus Lot average Density

)

i 1.00 l I I I I I O 20 40 60 80 100 120 140 A::al Positecn. Intnes POWER SPIKE FACTOR VS AIIAL Postil0N Figure 4.3 2 l

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5. NUCLEAR DESIGN 11 P_bysics Characteristics Table 5.1-1 coepares the core physics paramaters 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 para =eters are to be expected between the cycles.

The longer cycle 3 will produce a slightly larger cycle dif ferential burnup than that for the cycle 2. The accumulated average core burnup will br higher in cycle 3 than in cycle 2 because of the presence of the once-burned batch 4 fuel and the twiec burned batch 3 fuel. Figure 5.1-1 illustrates a representat ive relative power distribution for the beginning of the third cycle at full power with equilibrium xenon and normal rod positions.

The critical boron concentrations for cycle 3 are approximately the same as those for cycle 2 but vary slightly due to cycle length differences, radial power distributions. etc. The control rod worths for hot full pcser difier between cycles due to changes in group designat ions as well as changes in radial flux distributions and isotopics. The ejected rod worths in Table 5.1-1 are the naximum calculated values within the allowable rod insertion l imit s. It is difficult to compare values between cycles or between rod patterns since neither the rod patterns from which the CRA is ejected nor the isotopic dist ribut ions are ident ical. Calculated elected rod worths and their adherence to criteria are considered at all times in life and at all power levels in the development of the rod position limits presented in Section 8.

The maximum stuck rod worth for evele 3 is lower than for cycle 2 at the beginning of cycle but higher oc end of cycle. However, no adverse safety implicat ions are associated with this higher worth since the adequacy of the shutdown margin with cycle 1 stuck rod worths is demonstrated in Table 5.1-2.

For t he shutdown calculations the following conservatisms were applied.

1) Poison material depletion allowance

2) 10% uncertainty on net rod worth

1) Flux redist ribut ion persalty 51

Flux redistribution was accounted for since the shutdown analysis wa% calculated using a two-dimensional model. The shutdown calculation at

~

the end or cycle 3 is analyzed at approxiaately 230 EFPD'r. This is the latest t ir;c (+ $ days) in' core life in which.the transient bank is ne.arly fully inserted. After 230 EFPD's the transient bank will be almost fully withdrawn thus increasing the availabic shutdown margin. Reference fuel cycle shutdown margin is presented in the Oconce I, Cycle 2 Reload Report.

The cycle 3 power delicits f rom hot zero power to hot full power are higher than thow for cycle 2 due to a mvre nega .ve moderator coefficient in cycle 3.

The dif ferential bcron sorths and total menor, worths for cycle 3 are the same or lower than for cycle 2 due to depletion of the fuel and the associated buildup of fission products. Effective delayed neutron fractions for both cycles show a decrease with burnup.

5.2 Analvtical input The cycle 3 incore measurement calcul tion constants to be used for comput ing core power distribut ions were prepared in .he saec manner as the reference eyele.

5.3 Changes in *;uclear Design There were no relevant changes in core design between the reference and reload cycles. The same calculational methods and design information were used to obt ain t he important nuclear design parameters. In addition, no significant operational procedure chanyes exist from the reference cycle with reg.srd to axi.nl or radial power shape control xenon centrol, or tilt control.

The operat ional linits (Technical Specifications changes) for the reload cycle are shown in Section H.

A fuel melt limit of 20.15 kw/ft has been employed in calculating the Reactor Protection System (RPS) setpoints and is the same as in cycles 1 and 2. The batch 5 fuel assemblies will be loaded as in Figure 3-1. As-built data h.sve been used to ensure eighth-core syrset ry in 2 35 L ' loading. The three hatch I. assemb!!cs that had been assigned a maxi-um linear power rating of 20.02 kw/ft based on as-hu!!t d.ita will again be placed in lower power core locat tor.s. These locations (E-10. L-II, and M-6) have been investigated and l I

it has been determined th.it af ter the f uel has been shuf fled to these cycle 3 l l

5-2

. .y 4

1 core Iccations, they will not experience greater than 19.4 kw/ft throcch cycle 3.

Thus, a sufficient fuel melt margin will be =aintained through cycle 3.

See reference 2 for a detailed outline of the cethods involved.

In addition, assembly ID61, which contains s1=ulated f uel colu=n i

Raps, sill be placed in core location F-13 in conjunction with B&*s continuing progran to evaluate fuel performance. Contained in one fuel rod of assembly ID61 . ire three ceramic spacenswhich simulate fuel densification gaps. The de.crintion of the irradiation program for this special assembly in Oconee Unit I was presented in a letter (6/18/74) to Angelo Giambusso,1*SAEC.

Ccntinuation of the irradiation of assembly ID61 will not adversely affect fuel or r(actor performance during cycle 3.

l

'I l

4 h

E i ,

l i

I

, 5-3 e

g -  %

-r P

Table 5.1-1 Cych 2 .snd Cvele_ 3 Physics Parsmeters Cvele 2 Cvel 3

.yele length. EFPD 290 292 tycle burnup. :Wd/mtU 9000 9107 A.cr.ae core burnup - EOC. .W4/:rtU 14.550 17.254 Inittst core loading, atU 82.6 82.3 Critic.sl buron - EOC. ppm hzP* -

all rods out H2P 1285 1332 groups 7 and 8 inserted HFP -

1159 116) groups 7 and 8 inserted 1028 977 Critical boron - EOC. ppe H2P -

all rods out 285 36 4 HrP - group 8 (37.5% withdrawn, equil. Xe) 75 56 Control rod worths - HFP. BOC It,k/k Group 6 Group 7 1.12 1.39 Group 8 (37.5% ud) 1.14 1. 30 0.36 0.45 Control rod warths - HFP. EOC 1* k/k Group 7 droup 3 (37.5% wd) 1.97 1 . 31 0.41 0.46 Maximum ejected rod worth - HZP. 31.k/k Pad configuration 1 Rod config*sration 2 0.71 0.82 0.80 0.27

taxt=um situck rod worth - HZP. I'.k/k boC E9C 2.55 2.34 1.96 2. 71 Power deficit. HZP to HFP.' % ?k/k BOC (groups 7 .and 8 inserted) 1.34 E0C (groups 7 and 8 inserted) 1.45 1.99 2. f,a =

• Dopple r coe f f. - BOC. 10~ ?k/k/F 100*. power (O Xe)

-1.60 -s.60 Doppler coe f f. - E0C. 10" .k/k/F 100i power (equil. Xc) -1.62 -1.62 5-4 l

l 1

l

4 q.

Tjble 5.1-1 (Continued)

Cycle 2 Cycle 3

'*0 der..to r ew t f. - liFP. 10 1.k/k/F S h! . Ci L . ING pra. greups 7 and 8 inserted) - 0. 79 -0.89 i:th', ti ;ull . L , 17 ppa, group 8 inserted) -2.35 -2.42 30ron worth - !!FP. pral'?k/k EOC (1000 ppa) 97

. EOC (l 7 ppm) 103 91 92 i

Nonon vorth - HFP. **k/k 30C (4 el.ays)

EOC (equilibrium) 2.64 2.64' 2.69 2.67 Ef fec tive delayed neutron fraction (HFP) 4 BOC Eoc

.00602 .00562

.00520 .00518

• HZP denotes riot .'ero Pow. r/ilFP denotes Hot Full Power

) A*

Power deficit at 233 EFPD's j ..

1

+

e 4

l l

j 5-S '

i .i I

r k-E

Table 5.1-2 Shutdown Margin Calculation Oconee 1. Cycle 3

!. .Wallable Rod Worth BOC, ; k/k EOC*, i k/k

a. Total rod worth HZP 9.43 9.21 .

5. Worth reduction due to turnup of poison caterial -0.33 -0.37

c. ."axi=us stuck rod. H2P -2.34 -2.6*

d. Net worth 6.76 6.20

c. I.e s s 10; uncertainty -0.68 -0.62

2. Total available worth 6.08 5.58

11. hequired Rod Worth

a. P.ner deficit, EFP to HZP 1.45 2.00

3. 'taxinu:2 allowable inserted roJ vorth 1.53 1.39

. Flux redistribution 40 0.69

d. Total required worth 3.33 4.08

1. . S: utdown .".argin f l . f . minus II .d.) 2.70 1.50 NtiTE: Required shutdown margin is 1.03; k/k

• For :ioutdown =argin calculations, this is defined as

• 230 EFPD, the latest time in core life in'which the transient bank is nearly. full in.

1 5-6 d

FIGURE 5.1-1 S0C (a EFPD), Cycle 3 Two-Dimensional Rel.ative Power Distribution - Full Peser, Equilibriua Xenen,'?orm.al LM Positicas (Groups 7 .and 8 Inserted) 8 9 10 11 12 13 14 15 l l

H  : . .! i 1.16 1.28 1.47 1.00 1.21 .62 .59 7

g 1. D >  !.44 1.09 1.40 1.22 .62 .61 8

L .28 1.09 1.07 1.41 1.11 1.03 .57

.91 M  :.47 1.40 1.41 1.29 .97 40 .94 8

r( . 00 1.22 .97 1,29 1.23 .72

.91 7

0 'I l 11 90 1 23 53 63 P 92 .92 1.03 .94 ,

. 72 R .w . e,1 .57 N +-

Inserted Rod Group :;o.

. .u +

Relative Power Density 5-7

6. T!!ERMA1.-IlYDRAtt IC DESIG';

6.1 Thernal-liydraulic Design Calculations Ther al-hydraulic design calculations for support of cycle 3 operation ut i' i . cd the same analytical methods previously documented in referenees 1 and 2. Adjust ents to these calculations recognize the introduction of the Mh-h4 asse=blies in batch 5 and account for modifications in the use of the hLW-2 critical heat flux correlation.8 9 The B&W correlation was utilized in the licsnsing of the Oconee 1 Cycle 2 core. In the application of the B&W-2 CliF correlation to the oconee 1. Cycle 3 core, two modifications in the use of the correlation have been instituted. The following modifications have also been applied to the TMI-1, Cycle 2 core.

'. The limiting design DNBR of 1.30, representing a 95 percent cenfidence level for 95 percent population protection, was used in the analysis. A 11 citing DNBR of 1.32 representing a 94 percent confidence level for a 95 percent protection was used in previous design analyses. This change is consistent with industry practice and statistical standards associated with liniting design DNBR values as accepted by the NRC Staf f and ACRS.

2. The pressure range applicable to the correlation has been extended downward fron 2000 psia to 1750 psia. This revision 1% based on a review of rod bundle Cllr data taken at pressures below 2000 psia which shows that the B&W-2 correlation con-servatively predicts the data in this range, f.2 DNBR Analvsis la addition to the itens discussed above, the maximum design conditions cans!Jered in the FSAR and generie fuel assembl;* geometry based on total Mark B as-built data were taken into account. This resulted in a minimun DNBR of 2.0 at Ild percent power for undensified fuel.

The eticcts due to densificaticn can be divided into two categories:

(1) the result of reduced stack height and (2) power spiking caused by j densification induced gaps in the fuel column. As input to the DNBR analysis ter batch 4 fuel (nost limiting), the minimum lot average density and the l densified as-built stack height were used. Using this input and the carres-ponding power spike, the most liciting DNBR conditions were calculated for cycle 3 operation.

6-1

a

.t_

The axial flux shape which resulted in the maximum change in DN3R .

fret :he i

origiral design value was an outlet peak with .' core offset of

+11.5*.

The spike sagnitude and the maximum gap size are discussed in Sectien 4. 3 and the values used in the analysis are 1.07 and 1.96 inches, 4

rosnectively. The results of the two effects are -5.4% and -3.0% change in rfni.u= not channel DNBR and peaking margin, respectively. These numbers are- sun =arized in Table 6.2-1 which includes comparisons of other pertinent evele 2 and cycle 3 data.

i i

I  !

l i

l i

i s.

6-2 g v~ q re -

y- W --y -<.- yy y-- --my T- T- w

Table 6.2-1 Cycle 2 and Cycle 3 Design Conditions Cycle 2 Cycle 3 Pev r f. eve l, .%'t 2568 2568 System Pressure, psia 2200 2200

.% aetor Coolant Flew, ll Design Flew 107.6 107.6

.'.

• s . : Inlet Coalant Tenperature - 555.9 555.9 bro'. Pcwer, F

'ie.4se t out let Coolant Teciperature - 602.26 602.26 100 Pcwer , F Ret. Design Radial - 1.ocal Power 1.78 1.78

., Feaking Factor Ret. Design Axial Flux Shape l'.5 cosine 1.5 cosine Dsnsified Active Fuel Length

• 139.64 139.64 Aver.ne yeat Flux (100% Power), 176472 176472 P. t u / h- f t -

M.ixir t:n ifcat Flux (2002 Power), 471180 471180 ht u r h-i t-(for DNBR calculations) '

Ciff Correl.ition B 5'a*- 2 B &'.*- 2

?!!nicum DNBR (Max. Design Conditions, 2.0 2.0

0 Ds nsi f Ieat ton Penaltles) (112 power) (1122 Power) het Channel Factors

!:r.thalpy Rise 1.011 1.011 llea t Ilux 1.014 1.014 11.w Area 6.98 0.98 Dinstfle.ition Effects thinge in DNBR Margin, 2 -5.4 -5.4 Citange in Power Peaking .'fargin, 7. -3.0 -3.0 4

Nu-her ussd for DNBR analysis (batch 4 length). See Table 4.bl for values for batches 3 & 5.

1 6-3

7. ACCIDENT A';D TRA'*SIENT ANAI.YSIS 7.1 General Safety Analysis Each FSAR accident analysis has been examined with respect to changes in cy, le 3 paraneters to determine the effects of the cycle 3 reload and to ensure that thermal performance during hypothetical transients is not degraded.

Core thermal parameters used in the FSAR accident analysis were design operating values based on csIculated values plus uncertainties. A comparison of cycle 2 values of core thermal paraceters with paraceters used in cycle 3 analysis is given in Table 6.2.1. Cycle 2 and cycle I core thermal parameters are cocpared in reference 2. These are paraceters common to all of the accident analyses presented herein. For each accident of the FSAR, a discussion of the accident and the key parameters are provided. A comparison of t!.e key paraneters (See Table 7.1-1) from the FSAR and cycle 3 is provided with the accident discussions to show that the initial cor.ditions of the transient are bounded by the FSAR analysis.

The effects of fuel densfiication on the FSAR accident analysis results h1ve been evaluated and are reported in reference 4 Since batch i reload :uel assethlies do not contain fuel rods whose theoretical density is le er than those considered in reference 4, the conclusions in that reference are still valid.

Calculational techniques and twthods for cycle 3 analysis remain

.onsistent with those used for the FSAR. Additional DNER =argin is shown

or evele 3 due to use of the B&W-2 CilF correlation instead of the W-3 CHF correlation.

'1.2 new dose calculations were perforced for this reload report.

The dose censiderations in the FSAR were based on maximum peaking and burnup for all core cycles and therefore the dose considerations are independent of the reload batch.

1

7. 2 Rod Withdrawal Accidents

)

This accident is defined as uncontrolled reactivity addition to tne I core from withdrawal of contrcJ rods during startup conditiens or from rated power conditions. Both types of incidents were analyzed in the FSAR.

The ieportant parameters during a rod withdrawal accident are Doppler coefficient, moderator temperature coef ficient and the ratr e: shich reactivity is added to the core. Only high pressure and high flux trips l

\

7-1

)

are accounted for in the FSAR analysis, ignoring multiple alarts. interlocks and trips that normally preclude this type of incident.

For positive reactivity addition indicative of these events, the nost severe results occur for 301. conditions. The FSAR values of the key para =eters for BOL cenditions were -1.17x10 *

. k/k/F for the Doppler coef-ficient uJ.ix104

..k/k / F f o r t he rude ra t.' r tempe ra t u re- coefficient and red group worths un to and including ,t 10.0 ak/k rod worth.

Compa,rable cycle 3 para-netric values-are -1.60x10 ~#_ 2k/k/F for Doppler coef ficient. -0.39x10

• 2k/k/F ~

for naderator temperature coef ficient, and naximus rod bank worth of 9.41

?k/k.

Dae re f o re, cycle 3 parameters are bounded by design values assuced for the FSeR analysis. Th us , for the rod withdrawal transients, the consequences will be no more severe than those presented in the ISAR and the fuel densifi-cation report.

7. I "odera tor Di lut ion Accident Boron in the reactivity.

forn of boric acid is utilized to control exess The buron content of the teactor coolant is periodically reduced to conpensate for fuel burnup and transient xenon ef fects with dilution water supplied by the rakeup and purification system.

The coderator dilutlen transients onsidered .are the pucping er water with zt ro boron concentration f ro the r.ike ap t ank to the rc.ictor coolant system under conditions of full power oper.ition, hot .hutdown .ind during refueling.

The key p.araneters in this analysis are the initial boron cencen-tration. Soron reactivity worth, and moderator te=perature coefficent for powe r cases.

For positive reactivity addition of this type, the most severe rssults occur for ho!. conditions. The FSAR values of the key paraceters for 301. condit iens were 1400 ppm for the initial boron concentration. 75 ppa /17

• k/k buren reactivity worth and +0.9;x10" "k/k/F for the twderat or tecter.iture coefficient. Cor parable cycle 1 values are 997 ppm for the initial boron concentration. 78 ppn/ll. Ak/L baron reactivity worth and -0.89x10~ ak/k/F for the noderator tenperature coefficient.

The FSAR shows that the core and RCS are adequately protected during this event. Sufficient time for operator action to terninate this transient is also shown in the FSAR even wi th :uxi-nun dilut ion and minimun shutdewn nargin. The predicted cycle 3 parameter -

values of importance to noderator dilution transient are bounded by the FSAR design values, thus, the ana'.ais in the FSAR is valid.

7-2

7.4 Cold Water (Pump Startup) Accident The NSS does not contain any check or isolation values in the reactor coolant system piping. therefore, the classical cold water accident is not possible, liowever, when the reactor is operated with one or core pu.ps not runnine, and the idle pucps are started. the increased flew rate will cause the average core temperature to decrease. If the moderator temperature coefficient is negative, reactivity will be added to the core and a power increase will occur.

Protective interlocks and adsinistrative procedures exist to prevent the starting of idle pu=ps if the reactor power is above 222.

Ilowe ve r , these restrictions were not assumed and two pu=p startup from 50%

power was analyzed as the most severe transient.

To maximize reactivity addition, the FSAR analysis assumed the most negative moderator temperature coef ficient o f - 3. 0 x 10 ~

ak/k/F and least negative Doppler coef ficient of

-1. 3x10' Ak/k/F. The corresponding nost negat ive noderator tenperature coef ficient and least negative Doppler coefficient predicted for c;cle 3 are-2.42x10 ak/k/F and -1.60x10

~

~

ik /k /F, resp.etively.

As the predicted cycle 3 moderator temperat ure coef ficient is less negative and the Doppler coef ficient is more r'ega t i ve- than the values used in the FSAP.. the t rans i en t results would be less severe than those reported in the FSAR.

J,. ~) F]ow 1.o s s__ o f C o o I a n_

failure or A fromreduction a loss of in the reactor coolant flew can occur from mechanical electrical power to the pumps. With four indepen- I dent l pumps available, a mechanical f.nlure in one pump will not af fect cperation of others.

With the reactor at powe:r the ef fect of loss of coolant flow is a rapid increase in coolant tenpert.ture due to reduction of heat removal capability, th i s increase could result in DNB if corrective action were not

- taken innediately.

The key pan ameters for '.-pump coastd<wn or locked rotor incident are the flow rate, flow coastdown characteristles. Doppler coefficient, moderator temperature coefficient, and hot channel DNB peaking factors. The conservative initial conditions assumed for the densification report were:

FSAR values of flow and coastdown. -1.2x10' *

._k/k/F 42ppler coe f ficient .

+0.5x10' *

..k/k/F noderator temperature coef ficient, with densified fuel power spike and peaking. The results showed the LGBR renained above

1. 3 (W- 3) for 7-3

the 4-pump coastdown and the fuel cladding temperature re=ained belcw criteria limits for the locked rotor transient.

The predicted parametric values for cycle 3 are -1.60xlO ~5 li/m/F Doppler coef ficient. -0. 89 x if[ ak/k/F moderator temperature coefficient and peaking factors as shown in Table 6.2-1. Since tne B&W-2 CilF correlation was used for cycle 3 and the predictei cycle 3 values are bounded by those used in the densification report, the esults of that analysis represents the most severe consequences f rom a lo a of flew incident.

7.6 Stuck-Out. Stuck-In. or Dropped ontrol Rod Accident If a centrol rod is dropped into the core while operating, a rapid decrease in neutron power would occur, accompanied by a decrease in core average coolant temperature. In addition, the power distribution may be distorted due to a new cont rol rod pattern. Therefore, under these con-ditions a return to rated power may lead to localized power densities and heat fluxes in excess of design limitations.

The key parameters for this t ransient are n>derator temperature coef ficient, worth of dropped rod, and local peaking factors. The FSAR analysis was based on 0.46 ak/k and 0.36' ak/k rod werths with a moderator tenperature coef ficient of -3.0x10~ ak/k/F. For cycle 3. the maximum worta rod at power is 0.20% ak/k and the moderator temperature coefficient is

-2.42x10~ ak/k/F. Since the predicted rod worth is less and the coderator temperature coef ficient more positive, the consequences of this transient are less severe than the results presented in the FSA1.

7.7 tess of Elect ri - Power Tuo types of power losses were considered in the FSAR: (1) a loss of load condition, caused by separation of the unit f ram the t ransmission systen, and (ii) a hypothetical condition which results in a complete loss cf all systen and unit p owe r e x.:ep t the unit batteries.

The FSAR analysis evaluated the loss of load with and without turbine runback. When there is no runback a reactor trap occurs on hien reactor coolant pressure or temperature. This case resulted in a non-limiting accident. The largest offsite dose occurs for the second case, i.e., loss of all electrical power except unit batteries, and assuming operation with failed fuel and steam generator tube leakage. These results are independent of core loading and, therefore, the results of the FSAR are applicable for any reload.

7-4

_7. d Steam I.ine Fsilure-A steam line failure is defined as a npture of any of the stean lines f ro 2 the steam generators. t~pon initiation of the rupture, both stea:

unerators start to blowdewn, causing a sudden decrease in prie:ary e.ystem t er.pe ra t u re , pressure and pressurizer le ve l . The tecperature redaction leads to positive reactivity insertion and the reactor trips on high flux or lew RC pressure. Die FSAR has identified a double-ended rupture of the stean line between the steam generator and steam stop valve as the worst case situation at end-of-life conditions.

Tae key parameter for the core response is the a- derator tecperature coefficient which was assumed to be -3.0x10 ak/k/F in the FSAR. The evele 3 predicted value of moderator te=perature coef ficient is -2.".2x10 2k/k/F.

This value is bounded by the value used in the FSAR analysis and hence, the results in the FSAR represent the worst situation.

_7_. 9 Steas Generator Tube Failure A rupture or leak in a steam generator tube allows reactor coolant and ass sciated activity to pass to the seco1dary system. The FSAR analysis is based on complete severence of a steam gent rator tube. The pri=ary 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 llandline, Accident.

The mechanical damagt type of accident is considered the maxicu=

potential source of activity release during fuel handling activity. Die prieary conesrn is over raJiolozical releases which are independent of core loading anJ. therefore, the results of the FSAR are applicable to all reloads.

7.11 Rod Ejection Accident For reactivity to be added to t h e- core at a more rapi d ra t e than by uncontrolled rod withdrawal, physical f ailure of a pre %ure barrier component in the cont rol rod drive asse=bly must occur. Such a failure could cause a pressure differentist to act on a control rod asse-f>ly and rapidly elect the asse=bly f ron the core. This incident represents tire most rapid reactivity insertion that can be reasonably postulated. The values used in the FSAR and

'densification report at BOL conditions of -1.17xlu *k/k/F Doppler coef fielent.

+0.5x10 ?k /k/ I' naderator te=perat ure coef ficient , and i jected rod worth of 0.50 ak/k represent tiie maxinu= possible transient. The corresponding cycle )

l parametric values or -1.n0xlO-5 Ik/k/F Doppler and -0.s9xto-a 'k /k/F maderat er l

I l

7-5 i

I 1

I

teeperature coef ficient, both more negative than those used in reference /..

and a r.aximum predict.d ejected rod worth of 0.257. *.k/k ensure that the results will be less severe than those presented in the FSAR and densification report.

7.12 axinira Hypogt ical Acc ident There is no postulated nechanism whereby this accident can occur, since this would require a multitude of failures in the engineered t.afeguards.

The hypothetical accident is based solely on a gross release of radioactivity to the reactor building. The consequences of this accident are independent of core loading. There fo re , the results reported in the FSAR are applicable for all reloads.

7.13 Waste Gas Tank Rupture The waste gas tank was assumed to contain the gaseous activity evolved f rom degassing all of the reactor coolant following operation with 11 defective fuel. Rupture of the tank wois!d 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 loading and, therefore, the results reported in the FSAR are applicable to any reload.

7. !!. LOCA Analysis A generic LOCA analvsis for H6V 177 FA levered-loop NSS h.s. been performed using the Final Acceptance Criteria FCCS Evaluation .% del. Tais study is reported in BAW-10103 f reference 10). The analysis in BAV-l'J193 is generic in nature since the limiting values of ke parameters for all plants in th* category were used. Furthermore, the averate fuel temperat ure as a function of the linear heat r.ite and the lifetine pin pressure da t.:

used in the BAV-10lO 3 LOCA limits analysis are conservative compared to those calculated for this reload. Th us , the analysis and the IES I!cits reported in BAW-10103 provide conservat ive results for the operation of Oconee ! Cycle 3.

Table 7.14-1 shows the bounding values for allowable LOCA peak linear heat rates for oconee 1. Cycle 3 fuel.

7-6

6

_ TABLE 7.1-1 ' Comcarisen of Key Parareters for Accident Analysis FSAR &

1 Parameter Densification Report Value Predicted Cycle 3 Value Doppler Coefficient, BOL ~

-1.17x10 sk/k/F

, EOL. -1,33x10 ak/k/F

-1.60xlOSik/k/ F

-1.62x10 ak/k/ F i

.%derator Coefficient, BOL ~

+0.5x10fsk/k/F ~

EOL

~

-0.89x10 I k/k/ F

-3.0x10 ' sk/k/ F -2.42x10 *:k/k/ F

~

All Rod Bank Worth (!!ZP) 10.0% k/k 9. ".2 ?. k / k Initial Boron Concentration 1400 pps 977 ppr.

Boron Reactivity Worth 75 ppeiltak/k 78 ppr./17.*k/k Max. E*ceted Rod Worth (liFP) 1 0.50tik/k O.2 55. k /k Dropped Rod Worth.11FP 0.46* k/k 0.2r::.k/k

'l 1

I i 7-7 l

- , - - ~.- , y , , -

,w- ,

Table 7.14-1 A1.LL*.?/JLE LOCA ;-7?J LI : EAR liEAT RATE

_ Core Tlev.irfon, ft.

Allo *able Peck Linear IIca t it .t e , kt*/f t 2

15.5 4

16.6 6

18.0 8

17.0 10 16.0 7-8 e

P

3. PROPOSED !!O31FICATIONS TO TECllNICAL SPECIFICATIONS i

lhe Technical Specificatior.s have been revised for cycle 3 operation. The changes c.ade are as a result o,f:

(1) 1he use of a 95/95 confidence level rather than 99/95 as

, discussed in Section 6.1.

( .' ) The increase la range of applicability of the S&W-2 Clif correlation as discussed in Section 6.1.

(3) 1he use of the Final Acceptance Criteria I.0CA analyses i f 3r restricting peaks during operation as discussed in Section 7.I'..

( '. ) A revision to the assumptions upon which the flux / flow RPS setpoint is based. This setpoint now accounts for signal noise on the basis of data accumulated t ron operating BsW reactors.

Based upon the Technical specifications derived from the analysen.

presented in this report, the t'inal Acceptance Criteria ECCS linits will not be exceeded and the thert al design criteria will not be violated.

, 8-1

24:0 -

/

23:0 L ACCEPTABLE OPERATION

0 ,.

E t

=

@ 21:3 _

E

~

5 3 2203 0 !_

13:3 _

I 1903 r 1

553 560 6:0 620 640 660 Reactor Outlet Terperature.'F 1

1 CURVE REACTOR COOLANT FL0s (10 nri 1 141.3 106 DNBR Limit 2 105.6 a 106DNBR Limit 3 69.3 a 106 Quality Limit UNIT 1, CYCLE 3 CORE PROTECTION SAFETY LIMITS

,,, Figure 8-1

inermal Power Level '-

-,120

(,,,,,,,, UNACCEPTA8LE P ATION

(+3s.i 2) 110 ACCEPTABLE 4 PUMP OPERATION 100

(-*o.ss) 90

(-30.85.3) (es.3) (+3a.ss.3) (+5o,,5)

- 80 ACCEPTA8LE 3&4 PUMP CPERATION t-.o.6a.3) --

70

(-30.sa.2) - 60 (ss.2)* (+38.sa.2)

ACCEPTABLE 2.3 (*50 58 3)

-- 50 E4 OPERATION

( --o.

• s . 2 )

<- 40

(+so.3a.2)

-- 30

-- 20

- 10 40 -20 0 +20 +40 +60 Reactor Poser imoalance, i CURVE REACTOR COOLANT FLOW (in nri t 141.3 x 106 2 105.6 x 106 3 69.3 x 106 LNIT 1. CYCLE 3 "THE FLUX FLOW SETPOINT FOR 2 0 CORE PROTECTION SAFETY LIMITS PUWP OPERATION WUST BE SET AT 0.949 .

Figure 8-2

1 2t'l i

l P = 2355 psig T = S19 F l

2300 -

i 2200 -

T i I

E

.~

l ACCEPTABLE

$ 2100 OPERATION '

E l

c 8

E 2000 ~ M UNACCEPTABLE E [ OPERATION

J

~

Y 1900 s s

4 P = 1800 pssg I I t 1 540 560 580 SCO 620 640 Reactor Outlet Ter::erature, F l

UNIT 1. CYCLE 3 l PROTECT'CN SYSTEM MAXIMUM ALLCAtSLE SETPOINTS Figure 8-3 e-4

Poser Level. ',

120 UNACCEPraBLE OPERATION

,110(10s.5)

ACCEPTABLE ,

@o 100 = PowP s OPERATION b.

+'

l 90 I

t I i l 80 (7s.a) l ACCEPTABLE 3&4 PUMP I 70 0 "I"0" l

I I

e 60 I l l (51.7) l 50 AcctPIABLE 2.3 & u  !

l PUWP 40 0 ' ' " 0" l l 30

- 20 o o S 9 10

-40 -20 0 20 40 Reactor Power imoalance. >

-THE FLUX FLOW SETPOINT FOR 2 0 PUNP OPERATION UNIT 1. CYCLE 3 NUST BE SET AT 0.949 PROTECTION SYSTEM MAXIMJM ALLOWABLE SET POINTS Figure 8-4 3-5

R00 POSITICN LIMITS FCR 4 PLNP OPERATICN APPLICABLE TC THE PERICD FROM 0 TO 230 1 5 EFPC (196.iC2) (226.so2) 100 - 1 @

RESTRICTED REC 10%

POWER LEVEL CUTOFF 90 - e . . (226.92)

(896.92)

~

' * (" " # ' 4 OPERATION IN THIS REGION RESTRICTED 15 NOT ALLOWED 70 - REGIO".

60 .

E

=  ;

M 50 -

. (iss.50) l

\

E MININUM SHUT 00sN ,

GIN LIMIT PERNISSIBLE E OPERATING 33

~

REG 10N 20 -

e (es.es) 10 b (o. 6) 0 - - .

i . . , , .

0 100 200 300 Rod inder (',, Withdraen) 0 25 50 75 100 0 25 50 75 100 i i . . . . . . .

Group 5 Grcup 7 0 25 50 75 100 t a e f i Group 6 Rod index is tne percentage sum of the eithdraeal of Groups 5.6 ana 7. i UNIT 1. CYCLE 3 RCO POSITION LIMITS Figure 8-5 l

.s-6 l

l

I RCO PCSITION LIMITS FCR 4 PtM' OPERATICN APPLICABLE TC THE PERIOD AFTER 23015 EFPD (266. 002 )( 2 N.102 )

100 -

(2so.co2) G G POIER LEVEL CUTOFF h (266.92) ~ e _ 9, 80 - A goo,s p

~

OPERATION IN THIS REGION RESTRICTED IS NOT ALLORED GIONS

60 -

=

=

{ 50 -

(896 50) i O

40 -

T WININ!!E SHUTOOWN o 30 -

NARGIN LIMIT I PERMISSIBLE 20 - OPERATING (806.:5) ,

REGION 10 - '

b l (0.6)  !

0 1 . . . . . . t , , ,

0 100 200 300 Rod inden.

• Withdrawn 0 25 50 75 100 0 25 50 75 100 i e t I a f 3 e t Group 5 Croup 7 0 25 50 75 100 t I a f a Group 6 Roa inae as the percentage sum of the witnarasal of Groups 5.6 ana 7 UNIT 1. CYCLE 3 ROD PCSITION LIMIT Figure 8-6 e-7 J

1 1

ROC POSITICN LIMITS FOR 2 AND 3 P W OPERATICN APPLICABLE TO THE PERIOD FROM O TO 230 1 5 EFFD (ins.so2) (s75.so2) 100 -

OPERATION IN THIS REGION CPERATION IN THIS REGION 90 -

15 NOT ALLONED WITH 2 OR IS h0T ALLCeED WITH 3

@ 3 PUMPS PUMPS 5 80 -

2 70 "

PERMISSIBLE

$ OPERATING

$ 60 a:

REGICN w

0 50 -

('"5 50)

=

3 40 WININUN SHUT 000N WARGIN E ~

LIMIT 20 -

T (72.as) 10 -

0 (o.6) 0 ' ' ' ' r i . . . , ,

0 100 200 300 Rod inden. . Withdraen 0 25 50 75 100 0 25 I t a e 50 75 100 a 3 f g g g Grcup 5 Gr::up 7 0 25 50 75 100 t f I a a Group 6 Sco . mace es the percentage sua of tae .staorawal of Groups 5. 6 aad 7 UNIT 1. CYCLE 3 ROD POSITION LIwITS Figure 8-7

-e

RCO POSITICN LIMITS FOR 2 APC 3 PUMP OPERATION APPLICABLE TO THE PERICC AFTER 230 1 5 EFPD (205.102) (236.102) 100 -

9 .

~ OPERAil0N IN THIS REGION IS NOT t.LL0 SED WITH 2 OR 3 PUNPS OPERATION IN 5 THIS REGION IS 80 -

NOT ALLOWE0 j WITH 3 PUNPS 2; -

S S0 -

M 3 .

. (196.50) 5 40 -

a NININUM SHUTOOWN 3 -

NARGIN Lluti -

PERulSSIBLE CPERATING

. 20 -

u REGION j *

(806.15)

(0.6) 0 e i i i . . . i , , ,

0 50 100 150 200 250 300 ;

Roa inaex ', Es tnaraon 1

0 25 50 75 100 0 25 50 75 100 t f a e a 3 g 3 e e i Group 5 Group 7 0 25 50 75 100 Group 6 ses amoes es t*e serceatage sum of t%e =sthera.at of Groues 5. 6 and 7.

UNIT 1. CYCLE 3 RCO POSITION LIMITS Figure 8-8 o-9

Poser. ', of 2568 WWt RESTRICTED REGl0m (102.-i5) g , g (102.+5)

(92.-85) i n

_ (p (92.+5)

(80.-20) 80 -- > (80.+20) 70 PENMIS$18LE operating 60 . Region 50 .

40 .

30 . .

20 10

-30 -20 -10 0 +10 +20 +30 Core imoalance. ?

UNIT 1. CYCLE 3 OPERATIONAL POWER IMBALANCE ENVELOPE FOR OPERATICN FRC)4 0 TO 230 1 5 EFPD F.tgure 8-9 e-10

Pomer. ; of 2558 wet R[$1RICTED GEGlos

(-10.102) 100 (+ 5.8o2)

(-12.92) 90 (+ss.92)

(-20.so) 80 (+20.so) 70 PERMi$$18LE oP[ RATING aEGion 50

- 40 30 20 10

-30 20 -10 0 + 10 +20 +30 Core imoalance.

UNIT 1. CYCLE 3 OPERATIONAL POER IMBALANCE ENVELOPE FOR OPERATION AFTER 230 1 5 EFPC Figure 8-10 d-11

20 , , , , , ,

18 ,-

2

.x

/

~

/

. / -

• /

5 a: 16 -/ /

t 5

I -

3 Sc 14 3

CEhERIC e

n UNIT 1. BATCH 4 -~ ~

h_ 12 i

10 ' ' ' ' ' >

0 2 4 6 8 10 12 Axial Location of Peak Poner From Bottom of Core, ft LOCA LIMITED MAXIMLN ALLCWABLE LINEAR HEAT RATE Figure 8-11 3-1 1

i l

t

9. STARTT.~P PROGRAM The planned startup testing associated with core performance are provided below. These tests verify that core performance is within the assumptions of the safety analysis and provide the necessary data for continued safe plant operation.

Pre-Critical Tests

1. Control Rod Drive Trip Tim Testing Zero Power Tests

1. Critical Boron Concentration

2. Temperature Reactivity Coef ficient

3. Control Rod Group Worth

4. Ejected Rod Worth Power Tests

1. Core Power Distribution Verification at Approximately 40.

75. and 100: FP Norcal Control Rod Group Configuration

2. Core Power Distribution Verification at Approximately 40 FP With Worrt Case Dropped Rod Fully Inserted 3.

Incore/out-o f-Co re Detector Imbalance Correlation Veri-fication at Approxt=ately 75 FP

4. Power Doppler Reactivity Coef ficient at Approximately 1001 FP

5. Temperature Reactivity Coef ficient at Approximately 100: FP 9-1

10. REFEPINCES

1. Oconee Nuclear Station-Unit 1. Final Safety Analysis Report, Docket No.

50-269.

2. Oconee 1 Cycle 2 Reload Report, BAW-1409 (Rev. 1), Babcock & Wilcox, October 10, 1974.

3. A. F.J. a.eker t , !!. W. Wilson, and K. E. Yoon, "Progra:s to Determine In-Reactor Performance of B&W Fuels-Cladding Creep Collapse," BAW-10094P-A, Babcock & Wilcox, January 19 75.

4 Oconee 1 Fuel Densification hport, BAW-1388 (h v. 1), Babcock & Wilcox, July 1973.

5. Fuel Densification Report, BA* i -10055 (Rev. 1). Babcock & Wildox, June 1973.

6. "DensifieJtion Kinetics and P.ver Spike 2 del," Neting with USAEC.

July 3, 19 74; J. F. Harrison (B&W) to R. lobel (USAEC) Telecon, "P(wer ,

Spike Factor," July 18, 1974

7. C. D. N rgan and H. S. Kao, " TAIT - Fuel Pin Temperature and Gas Pressure Analysis ," BAW-10044, Babcock & Wilcox, Ny 19 72.

8

" Correlation of Critical lleat Flux in a Bundle Cooled by Pressurized Water "

BAW-10000, Babcock & Wilcox, .v. arch 1970.

9

" Correlation of Critical Heat Flux in a Bundle Cooled by Pressurized Water "

BAW-100 36 Babcock 6 Wilcox, February 1972.

10. "ECCS Analysis of B&W% 177 FA towered-Loop NSS." 3AW-10103 (Rev. 1).

Babcock & Wilcox, September 1975.

10-1