ML19322B872

Cycle 4 Reload Rept

ML19322B872
Person / Time
Site:	Oconee
Issue date:	03/30/1977
From:	BABCOCK & WILCOX CO.
To:
References
BAW-1447, NUDOCS 7912060703
Download: ML19322B872 (59)

Text

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BAW-1447 March 1977 d

OCONEE UNIT 1 CYCLE 4

- Reload Report -

9 267 7

i*sfik r....,.r~uases CC2.Y'.13GC11VUs /

Babcock &Wilcox 77cp>o y,,-

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'b rc h 19 7 7 OCOSEE L' NIT 1 CYCLE 4

- Reload Report -

BABCOCK & WILCOX Power Generation Group Nuclear Power Generation Divi.sion P.O. Box 1260 Lynchburg, Virginia 24505 Babcock & Wilcox

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l CONTENTS i

l Page 1.

INTRODUCTION.

1-1 i

2.

OPERATING F: STORY 2-1 3.

GENERAL DESCRIPTION 3-1 4

FUEL SYSTEM DESIGN.

4-1 4.1.

Fuel Assembly Mechanical Design 4-1 4..'. Fuel Zed Design 4-1 4.2.1.

Cladding Collaps.:

4-1 1

4.2.2.

Cladding Stress 4-2 4.2.3.

Fuel Pellet Irradiation Swelling.

4-2

4. 3.

Thercal Design.

i 4-2 4.3.1.

Power Spike Model (Densification) 4-3 4.3.2.

Fuel Tenperature Analysis 4-3 4.4 Material Design 4-3 4.5.

Operating Experiences 4-3 i

5.

NUCLEAR DESIGN.

5-1 5.1.

Physics characterist ics 5-1 5.2.

Analvtimal Input 5-2

5. 3.

Changes in Nuclear Design 5-2 6.

T!!ERMAl.-IlYDRACLIC DESIGN.

6-1 6.1.

Eva' ation.

6-1 6.2.

DNBR Analysis 6-1 6.3.

Dressure-Temperature Limit Analysis 6-2 6.4.

Flux / Flew Setpoint Evaluation 6-2 7.

ACCIDENT AND TRANSIENT ANALYSIS 7-1 7.1.

General Safety Analysis 7-1 7.2.

Rod Withdrawal Accidents.................

7-2 7.3.

Moderator Dilution Accident 7-2 7.4.

Cold Water (Pump Startup) Accident 7-3 j

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 Line Failure....................

7-5 7.9.

Steam Generator Tube Failure 7-5

- Babcock & Wilcox

Contcars (Cont'd)

Page 7.10.

Fuel Handling Accident 7-5 7.11.

RoJ Ejection Accident 7-6 7.12.

Maximum Hypothetical Accident 7-7 7.13.

Waste Gas Tank Rupture 7-7 7.14.

LOC \\ Analysis 7-7 8.

PROPOSED MODIFICATIONS TO TECHNICAL SPECIFICATIONS 8-1 9.

STARTl!P PROGRAM.

9-1 RFFERENCES A-1 1.ist of Tables Table 4-1.

Fuel Design Parameters.

4-4 4-2.

Fuel Rod Dicensions 4-4 4-3.

Input Sunnary for Cladding Creep Collapse Calculations 4-5 4-4 Fuel Thermal Analysis Parameters.

4-5 5-1.

Oconee 1, cycles 3 and 4 Physics Parameters 5-4 5-2.

Shutdown Margin Calculation for Oconee 1, Cycle 4 5-6 6-1.

Thermal-Hydraulic Design Conditions 6-3 7-1.

Comparison of Key Parameters for Accident Analysis 7-8 List of Figures __

Figure 3-1.

Core Loading Diagram for oconee 1, Cycle 4 3-3 3-2.

Enrichment and Burnup Distribution for Oconee 1 Cycle 4..

3-4 3-3.

Control kod Locations for Oconee 1, Cycle 4 3-5 4-1.

Maximum Gap Size Vs Axial Position -- Oconee 1 Cycle 4.

4-6 4-2.

Power Spike Factor Vs Axial Position -- Oconee 1, Cycle 4...

4-7 5-1.

BOC (4 EFPD), Cycle 4 Two-Dimensional Relative Power Distribution - Full Power, Equilibrium Xenon, I;ormal Rod Positions (Groups 7 and 8 Inserted) 5-7 8-1.

Oconce 1 Cycic 4 -- Core Protection Safety Limits 8-3 8-2.

Oconee 1. Cycle 4 -- Core Protection Safety Limits 8-4 8-3.

Oconee 1 Cycle 4 - Core Protection Safety Limits 8-5 H-4 Oconee 1, Cyele 4 - Protective System Maxiuum Allowable Setpoints 8-6 8-5.

Oconce 1 Cycle 4 -- Rod Position Limits for Four-Pucp Operation From 0 to 100 (-10) EFPD.

8-7

- 111 -

Babcock & Wilcox

F_ilu re s _ (Con t ' dj),

j Figure Page 6-6.

Ocence 1 Cy'cle

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- Rod Por.ition Limit s fcr Four-fuep (Teration Frca 100 (:10) to 250 ( 10: EFPD.

d-8 S-7.

Ocence 1. Cycle 4 - Rod Fosf t ton Linits for Four-Pucp cperation After 250 (710) EFPD.

S-9 H-8.

Oconee 1. Cycle 4 - Rad Position Limits for Two-and Three-Pump Operation Fron 0 to 100 (:10) EFPD 8-10 i

8-9.

Oconee 1. Cycle 4 -- Rod Position Limits for Two-and Three-Pump Operation From 100 (!10) to 250 (210) EFPD 8-11 l

H-10.

( conee 1. Cycle 4 - Rod Position Limits f or ?.ro-and Three-Putip Operation Af t er 250 (:lth EFPD 8-12 4

M-11.

Oconee 1. Cvele 4 - 9perational Power Ir. balance Envelope for Operation From G to 100 (110) EFPD.

8-13 8 - 1.'. Oconce 1. Cycle 4 - Operational Power imbalance Envelope I

for Operation Fron 100 ( 10) to 250 (:10) EFPD.

8-1 1

6-13.

Oconee 1 Cycle 4 - Operational Power Imbalance Envelope for Operation After 250 (210) EFPD.

8-15 l

8 - l '.. Oconee 1. Cycle 4 -- APSR Position Limits for Operation Fren 0 to 100 (210) EFFD.

8-16 i

S-15.

Oconee 1 Cycle 4 - APSR Position Linits for Operation l

Fron 100 (t10) to 230 (;10) EFPD.

8-17 8-16.

Ocor.ee 1. Cycle 4 - APSR Position Linits for Operation

]

After 250 (-10) EFPD.

8-18 8-17.

LOCA-1.inited Maximum Allowable Linear Heat Fate for i

Oconec Unit 1 8-19 i

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Babcock a.Wilcox

1.

INThu'!CTION This report justifies the operation of the fourth cycle of Oconce Nuc1 car Sta-t ion, L*n i t 1 (Oconee 1) at the rated ccre power of 2568 ?Nt.

Included are the required an.tlyse6 as outlined in the USNRC document " Guidance for Proposed License Am.ndrents Relating to Refueling," June 1975.

1 ro support Cycle 4 operat ion of Oconee 1, this report employs analytical tech-niques ond design bases establish:d in reports that were previously subsitted and accepttd by the USNRC and its predecessor (see references).

A brief so nsry of Cycle 3 and 4 reactor parameters related to power capability is included in section 5 of this report. All the accidents analyzed in the i

FSAR have been reviewed for Cycle 4 operation.

In those cases where Cycle 4 characteristics proved to be conservative with respect to those analyzed for Cycle 3. no new analyses were performed.

The Technical Specifications have been reviewed, and the modifications quired for Cycle 4 operation are justified in this report.

Based en the analyses performed, which take into account the postulated effects os fuel censifi ation and the Final Acceptance Criteria for Emergency Core Cool-ing Systens, it has been concluded that Oconee 1, Cycle 4 can be safely eperated at the rated pewer level of 2568 PMt.

1-1 Babcock &Wilcox I

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2.

OPERATING P.ISTORY The reference cycle for the n'iclear and thermal-hydraulic analyses of Oconee 1 i

is Cycle 3, which commenced power escalation on April 6, 1976, following com-l pietion of the zero pc.wer physics testing. The rated power level of 25e8.Wt was cchieved on June 2. 1976. A control rod interchange was performed at 100 effectivo full power days (EFF3). No operating 77omalies occurred during Cycle 3 that would adversely af fect the feel performance during the Cycle 4 desigr. length of 292 EFFD. 50 control rod interchange is planned during Cycle i

4 aperation. Control tod bank 7 vill be withdrawn at 250 210 EFPD.

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3.

CENERAL DESCRIPTION The Oeenee Unit I reactor core is described in detail in section 3 of the a:once Nuclear Station, L' nit 1. Final Safety Analys is Re;' ort !.

The Cycle < core, comp rising batches 4, 5, and 6, contains 177 fuel assemblies, each of which is a 15 by 15 array containing 208 fuel rods, 16 centrol rod guide tubes, and one incore instrument guide tube.

The fuel pin eladding is c.*1d-worked Zircaloy-4 with an OD of 0.430 inch and a wall thicknesa of 0.0263 lach.

The f uel censists of dished-end. cylindrical pellets of uranium dioxide which are 0.370 inch in diameter.

(See Tables 4-1 and 4-2 for additional data.) The fuel assemblies in batches 4, 5, and 6 have an averace nominal fuel loading of 463.6 kg of uranium. The undensified nominal active fuel lengths and theoretical densities vary between batches and are presented in Tables 4-1 and 4-2.

Figure 3-1 is tne core loading diagram for Oconee 1, Cycle 4.

The initial e,richnents of batches 4A and 4B were 2.60 and 3.20 wt.'. / ' 5U, respectively.

Entches 5 and 6 are enriched to 2.73 and 2.795 wt,1 235', respec t ivel y.

All 1

the batch 3 assemblies will be discharged at the end of Cycle 3.

The batch 4A. Jan, and 5 assemblies will be shuffled to new locations at tie beginning of Cycle 4.

The f resh batch 6 assemb'.ics will occupy the periphery of the core.

Figure 3-2 is an eighth-core map showing the assechly burnup and en-richment dist r ibution at the beginning of Cycle 4.

Reactivity cont rol is supplied by 61 full-length Ag-In-Cd s ontrol rods and by soluble baron shim.

In addition to the full-length control rods, eight axial power shaping rods are provided for additional cont rol of the axial power dis-tribution. The Cycle 4 locations of the 69 control rods and the group desig-nations are indicated in Figure 3-3.

The core locat ions of the total pattern (69 control rods) for Cycle 4 are identical to those of the reference cycle indicated in the Oconee 1, Cycle 3 Reload Report.2 The group designations, hewever, differ between Cycle 4 and the reference cycle in order to minimize pever peaking.

3-1 Babcock & Wilcox i

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The nominal syste= pressure is 2200 psia, and the densified naninal heat rate is 5.80 L/ f t at the rated core po er of 23o8 St.

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3-2 Babcock r.Wilcox i

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Core Leading Diagra: f or ocu-see 1. Cycle Tutt las%5FER 4

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Fig 2re 3-2.

Enrichment and bur nup Distribu: fon f or oconee 1, Cycle -

2 9

10 11 12

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14 15 2 50 2.75 3 20 I 2.60 2.75 3 20 3 20 2 755 s

22042 11756 16734 2l063 7370 21432 18734 c

3.20 2 75 2.75 3.20 2 75 3.20 2.195 K

19E45 6411 7323 19708 6363 21009 0

3 20 3.20 2 75 2 75 2 795 2 755 L

12904 20970 6383 SE03 0

0 2.75 2.75 3 20 2.795 a

7117 10314 23533 0

3.20 2.755 2.795 N

23898 0

0 2.755 0

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L. : t re.1 Red Iee.it ic:.3 for Oconee i..

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4 a.

FUEL SYSTEM DESIG!;

4

.. _1.

Fuel Assembly Mechanical Design Fertinent f uel design parameters are listed in Table a-l.

All fuel assemblies are identleai in concept c.na are rechanical ly in t e rc han.;eabl e.

The reload t uel assermlies (batch n) are ths; sace as the Cycle 3 reload assemblics except f or ninor desi.:n nodif icat ions to the spacer grid corner sells, which reduce

-r. cr s;rtit interaction during handling.

In aedition, inproved test methods td;aanic f r. pac t testing) show the spacer grids to have a higher seismic capa-bilit. and th...- an increased safety margin over the values reported in ret-ence 3.

Isll ot her result s presented in the FSAR fuel assembly mechanie 1 lis. tu.,i.'n are appl icable to the reload fuel assechlies.

.2.

Fuei Red Den 1g 1 -

? rtinent fuel rod almensions for residual and new fuel are listed in Table

- 2.

ihe neehanical evaluation of the tuel rod is described below.

4.2.1.

Claddine Coll.ipse Cr. ca. ollapse analvses vert performed for three-eyele assembly power histories fer Oe.nec 1.

The hatch a f uel is more limiting than batch 5 and 6 fuel owing te ita lon.:cr precieus incore exposure time.

Table 4-3 is an input si m ry for tN h.at. a a creep analy,Is.

The hatch 4 assembly power histories were

.ut. ! ).:ed, an d t h e nos t limiting assembly for Cycle 4 was determined. The pre-dieted.oesembly power history for the mest limit ing assembly was used to de-tsrmine the rhist licit ing cellapse t ime as described in 3.W-1008aP, Rev. 1.

1 Measured power Ji8tribut ten data obtained during Cycle 3 operat ion cont irned the accuracy of the Cycle 3 design calculat ions used for the collapse ar.alysis.

The conservatisna in the analytical procedure are summarized below.

I.

The CKOV comput er cede was used to predict tae time to a.oilapse.

CROV conservat ively predict a collapse times, as demonstrated in-reference "..

2.

?;o credit is taken for fission gas release. Therefore, the net differen-tial pressures used in the analysis are censervatively high.

4-1 Babcock s,Wilcox

C 3.

The cladding thickness used was the LTL (lower tolerance limit) of the as-built neasurements. The initial ovality of the cladding used was the 1:TL (upper tolerance limit) of the as-built ceasurements. These values ere taken f rom a statistical sampling of the cladding.

The nost 1initing assembly was found to hace a collapse time greater than the maximum ;)rojected three-cycle life of 20,976 hours0.0113 days <br />0.271 hours <br />0.00161 weeks <br />3.71368e-4 months <br /> (see Table 4-1).

This anal-ysis was pertorned ising the assumptions on densificatien described in refer-ence 4.

4.2.2.

Cladding Stress The tatch 3 fue! ha.e been analyzed and documented in the 0cenee 1 Fuel Densi-f f eat ion 1:enort.

This fuel had lower prepressurization and lower densLty than the hatth 4 5, and 6 f uel t h2.t nakes up the Cycle 4 core.

Therefore, the stre..ses reperted fr reference 5 represent conservative values with respect to the Cycle 4 eere.

4.2.>.

Fuel Pellet Irradiation Swelling Th. fi.el.ic31;;:i cr iter ia.'specif y a limit of 1.00 on cladding circumferential j

plastic str.ain.

U.e pellet desig,n is set so that the plastic cladding strain is le s than l' at 3 5,000 'Nd/mtt.

The conservatisms in this analysis are li3 red below.

.l ine m aximu.. S; ecif ication value for fuel pellet diameter was used.

lnc na -:1 :o P'. c if icat ion value for fuel pellet density was used.

1.

The ela,lding ID used v.ss the lowest permitted Specificatian tolerance.

Ihc max twi ex:iceted ti.ree-t yele local pellet burnup is less than 53,000

• L'd /m t t.

3.

• berna! Desig

.41 1 fuel assemblies in this core are thermally and geomet rically similar. The fresh batch 6 fue! inserted for Cycle 4 operation introduces no signif ie. int differences in fuel thermal perfornance relative to the batch 3 fuel disenarged at the end of Cycle 3.

The design ninicum linear heat rate capability is 20.15 kW/it, as shown in Table 4-4.

Linear heat rate capabilities are based on 6 with fuel centerline fuel melt and were established using the TAFY-3 code dens i f ica t io: penalt ies based un the assumpt ion of instantaneous densit icat ion to 96.St of theoretical density.

4-2 Babcock s, Wilcox w

Power Spi'<e Model sjfnsification)

The pcwer spike model used for Cycle 4 analysis is t!.e same as that used for Cycle 3."

Figures 4-1 and 4-2 show the maxieum gap size and power spike fac-tor, respectively, versus axial position. The power spike f actor and gap size are based on unirradiated oatch 5 fuel (93.5% TD) with an assumed enrichment of 3.0 wt i 2}i*L.

These factors are conservat ively high for batch 4 and 6 fuel.

4.3.2.

Fuel Tenperature Analysis Ther ul analysis of the fuel reds assu=ed in-reactor densification to 96.5*.

theoret ical density (IDF). The analytical methods are the same as those docu-ment ed in ref erences 5 and 2 f or Cycles 1 and 3, reepect ively. The average fue' tt=mperature (1320F) shown in Table 4-4 for all three fuel batches is t sta trom the Limiting (batch 5) analysis used to define the linear heat rate (LriRi capability for the fuel.

This analysis is based on the LTL of the Spec-it i; at ion f uel density and assumes isotropic diametrical shrinkage and aniso-t ror e axial shrinkage (censistent with reference 7) resulting from fuel den-sif. cation.

4.a.

Sterial Design 1he natch 4 fuel.issemblies are not new in concept, nor do they utilize dif-ferent cemponent materials. Therefore, the chemical compatibility of all possidle fuel-cladding-coolant-assembly interactions for the batch 6 fuel as-4cr.S l is s a re identical to those of the present fuel.

4.5

@erating Experiences n&W's operating experience with the Mark B, 15 by 15 fuel assembly design has verified the adequacy of the fuel assembly design. As of November 30, 1976, the fel10 wing operating experience has been accumulated for the six B&W 177-fuel asser.bly plants using the Mark B fuel assembly:

Cumulative Current Max assembly net electrical Reactor cycle burnup, M'a'd /mt U output, mwd Oconee 1 3

23.079 15,019,689 Oconee 2 2

21,779 10,408,446 Oconee 3 2

19,365 9,288,458 TMI 1 2

23.037 11,329,905 Arkansas One 1

18.278 8,380,351 Rancho Seco 1

10,621 2,872,577 4-3 Babcock 8.Wilcox

Table G-1.

Fuel Design Paraseters Residual FA

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Batch 4 3atch 5 batch 6 Fuel assembly type Mark B-3 Mark B ~.

Mark B-4 No. of FAs 56/5 60 56 Initial fuel enrich., wt 235U 3.20/2.60 2.75 2.795 Initial fuel density, % TD

>94.5 93.5 94.0~

Initial fill gas pressure (min (a)

(a)

Sa=e as specified), psia batches 4, 5 Batch burnup, BOC, S*d/mtU 20,612/

S220 0

21,333 Cladding collapse time, EFPH

>30,000 30,0000).30,0000}

(a) Refer to proprietary information in reference 2, Table 4.1-1.

(b)A detailed three-cycle collapse analysis will be performed for sub-sequent reload reports. A cladding collapse time of %30,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> is an engineering estimate based on comparison of the batch 5 ar.d 6 designs with batch 4.

Table 4-2.

Fuel Rod Dimensions Residual FA New FA, Compg gt Batch 4 Batch 5 batch 6 Feel rods OD, in.

0.430 0.430 0.430 ID, in.

0.377 0.377 0.377 Fuel pellets OD, in.

0.3685(mean)

'O.370 0.3695 Density, % TD

>C4.5 93.5 94.0 l~nd ens. active 142 142.6 142.25 fuel length, in.

Flexible spacers.

Spring Spring Spring type Solid spacer ma-Zr-4 Zr-4 2r-4 terial 4-4 Babcock & Wilcox

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Table 4-3.

Inpu: Su= mary fer Cladding Creep Collapse Calculations Batch 4 Pellet OD, in.

0.3935 Pellet density, O TD 94.3 Densified pellet OD, in.

0.3e61 Cladding ID, in.

0.377 Cladding ovality (L'TL), in.

(a) l Cladding thickness (LTL), in.

(a)

Prepressure,(min spec), psia ta)

Post-densif'n prepressure (cold), psia (a)

Reactor system pressure, psia 2200 Stark height (undens.), in.

142 (a) Refer to proprietary information in reference 2, Table 4.2-2.

Table 4-4.

Fuel Thermal Analysis Parameters Batch

-A(a)

• B(a) 5(a) 6 Nominal pellet density, % TD 95.5 95.5 93.5 94.0 i

Pellet diameter, in.

0.3665 0.3685 0.370 0.3695 Stack height, in.

141.0(

141.O 142.6 142.25 I}

Densified fuel parameters (densif'n to 96.3% TD assumed)

Pellet diameter, in.

0. 3 o* 0 0.3640 0.3645 0.364'6 Fuel stack height, in.

140.50 140.30 140.46 140.47 Nominal LilR at 2563 MWt, kW/ft 5.80 5.80 5.80 5.80 Avg fuel temp at nominal LilR, F 1320 1320 1320 1320 LiiR to centerline fuel melt, kW/ft 20.15 20.15 20.15 20.15 (a) Data frem reference 2.

(b) Conservative calculational paraceter.

4-5 Babcock & Wilcox I

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l'over Sp ike F.6c t or Vs Ax i.il l'osi t ion - Oconee 1. Cycle 4 1.09 l

TOF = 96.5%

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5.

NUCLEAR DESIGN 5.1.

Physics Characteristics Table 5-1 compares the core physics parameters of Cycles 3 and 4.

The values for 'ooth cycles were generated using PDQ07. The accumulated average core burnup will be higher in Cycle 4 than in Cycle 3 because of the difference in initial core loading. Figurs 5-1 illustrates a representative relative power distribution for the beginning of Cycle 4 at full power with equilibrium xenon and normal rod pesitions.

The critical boron concentrations for Cycle 4 are higher than those for Cycle 3 because of a higher feed enrichment, different radial power distributions, etc.

The control rod worths for hot full power dif fer between cycles due to c hanges in group designations as well as changes in radial flux distributions and isotoples.

The ejected rod worths in Table 5-1 are the maximum calculated values within the allowable rod insertion licits.

It is difficult to compare values between cycles or between rod patterns since neither the rod patterns f rom which the CRA is ejected nor the isotopic distributions are identical.

Calculated ejected 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 ejected red worths given in Table 2

5-1 for Cycle 3 are for rod configurations 1 and 2, respectively, whereas the ejected rod worths for Cycle 4 are for the beginning of Cycic 4 and 250 EFPD, respectively. The maximum stuck rod worth for Cycle 4 is lower than i

that for Cycle 3 at BOC and at EOC.

All safety criteria associated with these worths are met.

The adequacy of the shutdown margin s'ith Cycle 4 stuck rod worths is demonstrated in Table 5-2.

The following conservatisms were applied for the shutdown calculations.

1.

Poiscn material depletion allowance.

2.

10 uncertainty on net rod worth.

3.

Flux redistribution penalty.

~

5-1 Babcock 3.Wilcox

Tlux redistribution was accounted for since the shutdown arslysis was calcu-lated using i two-dimensional model. The shutdown calcalation at the end of Cycle 4 is analyzed at approxicately 250 EFPD.

This is the latest time (-

10 days) in core life at which the transient bank is nearly fully inserted. After 250 EFPD, the transient bank will be almost fully withdrawn. thus increasing the available shutdowa margin. The reference fuel eyele shardown e.argin is presented in the O-onee 1 Cycle 3 lieload, Report.2 The Cycle 4 power deficits from hot zero power to hot full power are higher than those for Cycle 3 owing to a twre negative moderator coefficient in Cycle S.

The differential boron worths and total non worths for Cycle a are lower t han titose for Cycle J Lecause of fuel depletion and the associated buildup of fission products. Effective delayed neutron fractions for both cycles show a decr(-ase wit h burnup.

1 2.

Analytical Input The Cycle 4 incore measurement calculation constants to be used for computing power distribations were prepared in the same manner as those for the core reference eyele.

5,. 3.

Changes in Nuclear Design There were no relevant char.ger. In core design between the reference and reload cycles.

The same calculat ional methods and design informat ion were used to obtain the leportant nuclear design parameters. The only significant opera-tional procedure change f rom the reference cycle is the specific-* ion of APSR pos it ion limits in addition to the usual regulating control rod and icbalance

!!mits for ECCS.

The APSR position limits will provide addit ional control of peser peaking, and assurance that 1.0CA kW/tt limits are not exceeded. The op-

< rational linits (Technical Specification changes) for the reload cycle are shown in section 8.

Three Oconee I hatch 4b fuel assemblies have centerline fuel melt limits be-tween 19. i5 and 20.14 kW/ f t.

These assemblies have been selectively loaded in Cycle 4 (core loca t f ens N-7, G-4, and D-9) to ensure that these licits are not exceeded during the cycle lifet ime.

In addition, assembly ID61, which contaf"s simulated fuel column gaps, will be

{

placed in core location H-6 in conjunction with B&W's continuing program to evaluate fuel performance. Contained in one fuel rod of assembly ID61 are three i

3-2 Babcock a,Wilcox i

1 i

i

ceramic spacers which simulate f uel densification gaps.

The description of the irradiation program for this special assembly in Ocon 4

ee Unit 1 was presented in a letter (6/18/74) to Angelo Ciambusso, USNRC.

of asse=bly ID6L will not adversely affect Continuing the irradiation Cycle 4 fuel or reactor performance during 1

1 1

i 1,

n 5-3 Babcock & Wilcox

Tabic 5-1.

Oconee 1. Cycles 3 and 4 Physics Parar.eters gycle 3 Cycle 4 Cycle Icngth. EFPD 292 292 Cycle buraup, trid/stU 9107 9136 Average core burnup - 292 EFPD,

!k'J /mt U 17,254 19,034 Initial core loading, mtU S2.3 82.1 Critical boron - BOC, ppm (no Xe)

!!Zp(a ), group 8 inserted 1332(b) 1415 llZP, groups 7 and 8 inserted 1169 1333 ilFP, groups 7 and 8 inserted 977 1145 Critical boron - 292 EFPD, ppm b) it:P( group 8 (37.5% ud, equil. Xe) 364 373

,;pg 56 88 Control rod worths - IIFP("

BOC, %

2k/k Group 6 1.30 1.07 Croup 7 1.30 0.93 Group d ( 37.5% wd) 0.45 0.50 Control rod worths - IIFP, EOC ',

i Ak/k i

Group 7 1.31 1.16 Group S (37.5 wd) 0.46 0.47 Fax ejected rod worth - IIZP, % Ak/k Banks 5-8 ins., N-12 ejected g

,6 Max stuck rod worth - IlZP, % Ak/k BOC (N-12)(d)

2. 34 1.74 EOC (1.-14)(d) 2.71 2.02 Power deficit, IlZP to IIFP, % Lk/k BOC (groups 7 and 8 inserted) 1.45 1.49 EOC (groups 7 and 8 inserted) 2.00 2.07 Dop'pler coeff - BOC, 10'S( Ak/k) / *F 100% power (O Xe)

-1.60

-1.45 Doppler coef f -- EOC, 10-5(ak/k)/*F 100% power (equil. Xe)

-1.62

-1.55 Moderator coeff - IIFP, 10-4 (Ak/k)/*F BOC (O Xe, 1000 ppm groups 7, 8

-0.89

-1.00 ins.)

EOC (equil. Xe, 17 ppa, group 8 ins.)

-2.42

-2.55 Boron worth -- IIFP, ppm /% t.k/k BOC (1000 ppm) 103 109 D)C (17 ppm) 92 101 5-4 Babcoc4 & Wilcox

d t

i Table 5-1.

(Cont'd) l Cycle 3 Cycle 4 Xenon worth - HFP, 7. Akik i

j EOC (4 dcys) 2.64 2.60 EOC (equilibrium) 2.67 2.61 Effective delayed neutron fraction -

HFP l

BOC) 0.00582 0.00593 I"

EOCJ 0.00518 0.00530 1

(a)ll2P: hot zero power, IIFP: hot full power.

a 5

Calculated for HZP banks 1-8 out.

(c)EOC denotes 250 EFPD with bank 7 withdrawn.

(d) Cycle 4 positions.

i

)

4 1

1 Babcock &Wilcox 5-5

,-e-p

,-v.,

7-,--

-,w-,

v, y,

e

Table 3-2.

Shutdown.% rgin C.sicularion for 0.;once 1, Cvele 4 EOC, * *k/k EOC'#}

% *.k/k Available Rod *' orth Total rod worth, HZP 3.36 8.77 Worth reduction due to bure n of poison material

-0.34

- 0. ~.1 L:d=nn stuck rod, !!ZP

-1.74

-2.02 2;et worth 6.28 6.34 Le.ss 10% uncertainty

-0.63

-0.63 Total available worth 5.65 5.71 Required Rod Vorth P.wer deficit, liF? to HZP 1.49 1.07 L x allcat Lle ins (rted red wi.:th 1.12 1.08 Flux redistribution 0.61 1.05 Tet.nl required worth 3.22 4.20 Shutdown M r:in (total avail.

worth minus total required worth) 2.43 1.51 hte: Required shutdown margin is 1.00% t_k/k.

i

" For shutdown margin calculations this is defined as s250 EFPD, the latest time in core life at which the transient bank is nearly full-in.

I j

i i

i i

5-6 Babcock & Wilcox i

i i

l l

J

Fi g.:re 5-1.

EOC (4 ETP3), Cycle ? No-Dimensional Relat ive Power Distribution - Full Power. Equilibrinus Xenon, Normal Rod Positions (Groups 7.and 3 Inserted) 8 9

10 11 12 13 14 15 l1

.92 I 11 1.12

.97 1.37

.90 49 s

.61 I

g I.11 1.11 1.16 1 I7 1 08 1.07

.83

.69 7

8 L

I.12 1.16

.65 99 3 C8 1.16 1 21

.67 l

M 97 1 17

.99 1.17 1.14 1.05 1.05 I

8 N

1 17 1 08 1.08 1.14 1.08 1.27

.89 t

0

.90 1.07 1.16 1.05 1.27

.89 7

P 49

.83 1.21 1.05 77 R

.61

.69

.65 Inserted Rod Group NumDer X II Relative Poser Density 5-7 Babcock & Wilcox

6.

THERE\\L-HYDRAULIC DESIGN 6.1.

Evaluation The thermal-hydraulic design evaluation in support of Cycle 4 operation uti-lized the methods and models described in references 1, 2, and 5.

Cycle 4 analyses have been based on 106.5% of the (first core) design system flow 2

rate.

Cycle 2 and 3 analyses had used 107.6% of design flow, based on a measured flow value of 108.6%.

The reduced tiow rate has been selected for the Cycle 4 analyses to provide increased margin to the actual measured values.

The core configuration for Cycle 4 dif fers slightly f rom that of Cycle 3 in tnat the depleted batch 3 fuel removed at the end of Cycle 3 is the Mark 3-2 fuel assembly design, while the fresh batch 6 fuel inserted for Cycle 4 is the Mark B-4 assembly design. Mark B-4 fuel assemblies exhibit a slightly lower resistance to flow than do the Mark B-2 assemblies, resulting fron a revised end fitting design. This change has been considered in the Cycle 4 core flow distribution analysis. No credit has been taken for the increase in system flow that results from the reduction in total core pressure drop.

~

p.2.

DNBR _ _Ana l ys i s The BN4-2 CilF correlation sas used for thermal-hydraulis analysis of Cycle 4 Tais correlation, which has been reviewed and approved f or use with the Mark-B 9

fuel assembly design, has been previously used for licensing of Cycles 2 and 2

3 of t ho Oconee 1 core,

The effect of fuel densification on minimum DNER is primarily a result of the reduction in active fuel lcngth, which increases the average 's flux. The Cycle 4 DNBR analysis was based on the cold densified active length of batch 4 fuel (140.3 inches from Table 4-4).

This is a conservative method of ap-plying the densification ef fect because the hot assembly during Cycle 4 always cccurs in either batch 5 or batch 6 fuel, both of which have slightly longer densified lengths, and because no credit is taken for axial thermal expansion Babcock s Wilcox #

6-1 l

1 1

t' the fuel colu=n.

This analysis dif f ers f rom t hat of Cycle 3 in two re-sctst First. the effect of the c'ensification power spise is no longer con-

- r i tered tor ^5SR analysi 6 on the basis..f infer =ation preserted in references S-!L.

Secend, the densified active length is ir.corporated directly in o the h analv is, resulting in a calculated nini=un DNBR of 1.91 at 112% power i ; abl e e>- 1..

The C.ycle 3.inalvsis had been based on a 144-inch active length

-itte the e: t e; s of educed act i ee lenath and the densification power spike al ulated se; arat. ly.

p.itenti.1 effect of fuel rod Imv on DNBR is considered by incorporat ing

. a r ab le =.u sin < into 353-limitect RPS setpoints. The maximut rod bow magni-

=e.aa cal slated fro: the equation g = 11.5 + 0.069 viiti. where a is the b

• bew nigni t. ate (mils) and RU is the burnup (%'d/mt19.

The resultant DNBR

.!tv. Saee.1 on a burnup equivalent to three cycles of operation, is 5.9.

Pre 4nrt-Tepper..ture f.imit Analysis a-a su re-t empe..t ure limits for Cvele 3 were based on the assumption that one

.ctor i 1 n... a.ent ea ls e f a iled o; en, thus reducing the effective coolant

'cw tor nest trsnsfer by a.6*;.

An NKC staf f evaluation!S of operating data t : a= li&V p la n t s - relieved h&W f roa having to include a vent valve flow pen-a!t:. In sa:etv analyses. Itawever. for Cycle 4 operation the Cycle 3 pressure-limits have been retait..=d as the basis for the variable low-tv.p.rature rv uore trip 'unttion.

The Cycle 3 li=its are conservative with respect to i.e Cycle e acalvsis. which assumes closed vent valves and incorporates a

m K pent!tv due to rod bcw.

n.~.

Fl us / F'.=w Se tmi n_t_Eva l ua t_lgn i!. fl*.s/ flew trip setpoitit proposed for Cycle 4 operat ion is the same as h at used early in Cycle 3 (1.055).

The basis for this setpoint has been

r. vised to delete the densification power spike and vent valve penalties and ta redace (f rom 107.6* to 106.5% of design) the initial-condition system flow rate assumed for analysis of the two-pump coastdown. A suitable DNBR raargin la provided in this analysis to of f set the reduction in DNBR resulting from the maximum predicted end-of-cycle fuel rod bow.

Babcock s.Wilcox 6-2

Table 6-1.

Thermal-llydraulic Design Conditions Cycle 3(a)

Cycle 4 Powe r leve l. %*t 2568 2568 System pressure, psia 2200 2200 Reactor coolant flow, % design flow 107.6 106.5 Vessel inlet coolant temp, 100%

555.9 555.6 power, F Vessel outlet coolant temp, 100%

602.2 602.4 power, F Ref design radial-local power 1.783 1.783 peaking factor Ref design axial flux shape 1.5 cosine 1.5 cosine Active fuel length, in.

See Table 4-4 See Table 4-4 Average heat flux, 100% power, 175,640(b) 175,640(b)

Btu /h-ft2 Cl!F correlation BAW-2 B,W-2 Minimum DNBR (cLex design conditions, 2.0 NAIC) no densif'n penalties)

(112% power)

Ikit channel factors Enthalpy rise 1.011 1.011 ih at flux 1.014 1.014 Flow area 0.98 0.98 Minimum DNER with densif'n penalty 1.892 1,914 (a) Reference 2.

Lused on densified length of 140.3 inches.

(c)See section 6.2.

6-3 Babcock & Wilcox I

4m

k I

7.

ACCIDENT AND TRANSIENT ANAL.YSIS f

_7. 1.

General Safety Analysis Each FSAR accident analysis has been examined with respect to changes in h

Cycle 4 parameters to determine the effects of the Cycle 4 reload and to en-sure that thermal performance during hypothetical transients is not degraded.

Care thermal parameters used in the FSAR accident analysis were design oper-ating values based on calculated values plus uncertainties. A comparison of the Cycle 3 values of core thermal parameters with parameters used in the Cycle 4 analysis is given in Table 6-1.

Cycle 3 and 2 core thermal parameters are compared in reference 2.

Cycle 2 and I core thermal parameters are com-pa red in reference 12.

These parameters are common to all the accident anal-yses presented herein. For each accident of the FSAR, a discussion of the accident an.1 the key parameters are provided. A comparison of the key param-eters (see Table 7-1) from the FSAR and Cycle 4 is provided with the accident discussions to show that the initial conditions of the transient are bounded by the FSAR analysis.

The effects of fuel densificatien on the FSAR accident analysis results have been evaluated and are reported in reference 5.

Since batch 6 reload. fuel as-semblics do not contain fuel rods whose theoretical density is lower than.those considered in reference 5, the conclusions in that reference are still valid.

Calculational techniques and methods for Cycle 4 analysis remain consistent with those used for the FSAR.

Additional DNBR nargin is shown for Cycle 4 because the B&W-2 CliF correlation is used instead of the W-3 CHF correlation.

No new dose calculations were performed for this reload report. The dose con-siderations in the FSAR were. based on maximum peaking and burnup for all core cycles; theref ore, the dose considerations are independent of the reload batch.

- s.-

cr,_

-- ra:

it' 3

Babeccoa nttox 7-1 Babcock & Wilcox

7.2.

Rod 'a'ithdrawal Accidents This accide nt is defined as uncontrolled reactivity adiltien to the core due to wit hdrawal of control rods durir.;; startup conditions or f rom rated p.ver condition..

Both types of incidents were.ina yzed in the FSt.R.

The important parameters during a rod withdrawal accident are Deppler coef-ficient, coderator temperature coefficient, and the rate at which react ivity is added to the core.

Only high-pressure and high-flux trips are accounted for in the FSAR analysis; the cenitiple alar:s, interlock. and t rips t hat nor-call.s precluste this type of incident are ignored.

For positive react ivity ad<t it ion indicc 'ive of t hese ever.t s, the most severe result s occur f or BOL cond it ions.

The FSAR values of the key paraceters for LOL c ondit ion.s were -1.17 10 ' /k/k *F for the Doppler coefficient. 0.5 10 -

?.k /'

• F fe r the nodcrator temperature coefficient, and rod group worths up to and inchading a 10.07. 1.k/k red worth.

Comparable Cycle 4 paramet ric values are -1.43

_10"

.'k/k *F for Doppler coefficient. -1.00 - 10 ~ f.k/k *F for mod-c r..t e r t et.pe r.a t u re coefficient, and maximum rod bank worth of 8. Mi f.k/k.

'.herefore, Cycle 4 pararacters are bounded by the design values assu:ned for the FSu analwis.

Thus, for the roe withdrawal transients, the consequences will Se r.o n.o r s-severe than t hose presented in t he FSAR and the fuel densification report 7.1

??.uf = /a ti r M lution Accident ooren (in the form of boric acid) is used to control excess reactivity. The boron content of the reactor coolant is periodically reduced to compensate for fuel burnup and transient xenon eft'ects; dilution water is supplied by the

=akeup and purification systen. The moderator dilution transients considered are the pu= ping of water with zero boron concentration from the makeup tank to the reactor coolant system (RCS) under conditions of full-power operation, hot shutdown, and refueling. The key parameters in this analysis are the initial boren concentration the boron reactivity worth, and the moderator temperature coefficient for power cases.

for positive reactivity addition of this type, the most severe results occur for LOL condit ions.

The FSAR values of the key parameters for BOL conditions sere 1400 ppm for the initial boron concentration, 75 ppn/l* Ak/k boron reac-tivity worth and +0.5 10

• Ak/k *F for the moderator temperature coefficient.

7-2 Babcock & VVilcox

Comparable Cycle 4 values are 1145 ppm for the initial boron concentration, 84 ppm /1% t.k/k cold boron reactivity worth and -1.00 = 10 ' ak/k *F for the-moderator temperature coefficient. The FSAR shows that the core and RCS are adequately protected during this event.

Sufficient time for operator action to terminate this transient is also shown in the FSAR, even with maximum di-lution and minimum shutdown margin. The predicted Cycle 4 parametric values of importance to moderator dilution transient are bounded by the FSAR design value; thus, the analysis in the FSAR is valid.

7.4.

Cold 'n'a t er (Pump Startup) Accident The hSS contains no check or isolation valves in the RCS piping; therefore.

the classic cold water accident is not possible. However, when the reactor is operated with one or nore pumps not running, and these pumps are then started, the increased flow 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.

There are protsetive lu*erlocks and administrative procedures to prevent the starting of idle pumps if the reactor power is above 22I.

However, these re-strictions were not assumed, and two-pump startup from 50% power was analyzed as the most severe transient.

To naxintze reactivity addition, the FSAR analysis assumed the most negative coderator temperature coef ficient of -3.0 = 10-' ak/k *F and the least negative Doppler coeffiElent of -1.2 = 10'E ak/k *F.

The corresponding most negative moderator temperature coefficient and least negative Doppler coefficient pre-dic:cd for Cycle 4 are -2.55 a 10~4 ok/k *F and -1.45 = 10-5 Ak/k *F respec-tively. Since the predicted Cycle 4 moderator temperature coefficient is less negative and the Doppler coef ficient is more negative than the values used in the FSAR, the transient results would be less severe than those re-ported in the FSAR.

7.5.

Loss of Coolant Flow A reduction in reactoc coolant flow can be caused by mechanical failure or a loss of elect rical power to the pumps. With four independent pumps available, a mechanical failure in one pump will not affect the operation of the others.

With the reactor at power, the effect of loss of coolant flow is a rapid in-crease in coolant temperature due to reduction. of heat removal capability.

7-3 Babcock & Wilcox

This increase could result in DNB if corrective act ion were not taken immedi-ately.

The key parameters for a four-pump coastdown or i locked-rotor inci-dent are the flow rate, flow coastdown characteristles, Doppler coefficient, moderator temperature coefficient, and hot channel DNB peaking f actors.

The conservative initial conditions assumed for t 5e densification report were FSAR values of flow and coastdown. -1.2 10'; 2k/k *F Doppler coef ficient,

+0.5 = 10' Ak/k *F moderator tenperature coefficient, and densified fuel power spike and peaking. The results showed that the DNBR remained above 1.3 (W-3) f or t he four-pump coastdown, and the fuel cladding temperature re-nained Selow criteria limits for the locked-rotor transient.

The predicted parametric values for Cycle 4 are -1.45 = 10-i ?.k/k *F Doppler ccefficient, -1.00

• 10 ak/k *F coderator temperature coef ficient, and peak-ing factor s as shown in Tab!c 6-1.

Since the SiW-2 CIIF correlation was used for Cecle 4. ani the predicted Cycle 4 values are bounded by those used in the densitic.ition report, the results of that analysis represent the most severe tonsequences from a loss-of-flow incident.

7.6 S t uck-thet, Stuck-In, or Dropped Control Rod Accident if a control rod is dropped into the core while operating, a rapid decrease in i

neutron power would occur, accompanied by a decrease in the core average cool-uit t e ;ier at ure.

In acidition, the power diste lbution cuy be distorted due to a new c..n t r.1 roti pattern.

Therefore, under these conditions a return to rated power nay lead to localized power densities and heat fluxes in excess of de-sie Ilmitations.

ISc key ;>arameters for this transient are moderator temperature coefficient,

w. Ih of the dropped rod, and local peaking factors.

The FSAR analysis was basco on.).46 arid 0.367 ak/k rod worths with a noderator temperature coef-ficient of -3.0 - 10 Ak/k *F.

For Cycle 4 the maximum rod worth at power is 0.20 nk/k, and the moderator temperature coef f icient is -2.55 10 ak/k *F.

Since the predicted rod worth is less and the moderator temperature coefficient j

is more positive, the consequences of this transient are less severe than the results presented in the FSAR.

7.7.

less of Electric Power Two types of power losses were considered in the FSAR: a loss of load condition caused by separat ion of the unit from the transmission system, and a i

7-4 Babcock & Wilcox l

l l

hypothetical condition that results in a complete loss of all system and unit r

power except the unit batteries.

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

When there is no runback, a reactor trip occurs or. high reactor coolant pres-sure or tenperature. This case resulted in a non-limiting accident. The largest offsite dose occurs for the second case, i.e.,

loss of all electrical f

power except unit batteries, and assuming operation with failed fuel and steam generator tube lea kage.

These results are independent of core loading; there-fore, the results of the FSAR are applicable for any reload.

7.8.

Steam Line Failure A steam line fa12are is defined as a rupture of any of the steam lines fron the steam generators. Upon initiation of the rupture, both steam generators start t.> blow down, causing a sudden decrease in primary system temperature, pressure, and pressurizer level.

The temperature reduction leads to positive reactivity insertion, and the reactor trips on high flux or low RC pressure. The FSAR has identified a double-ended rupture of the steam line between the steam generator and steam stop valve as the worst-case situation at EOL conditions.

The key parameter for the core response is the moderator temperature coeffi-cient, which was assumed to be -3.0 m 10-4 Lk/k *F in the FSAR.

The Cycle 4 predicted value of moderator temperature coefficient is -2.55 = 10-4 Ak/k *F.

Thra value is bounded by the value used.'. the FSAR analysis; hence, the re-sults in the FSAR represent the worst situation.

7.9.

Steam Generator Tube Failure A rupture or leak in a stear generator tube allows reactor coolant and asso-ciated activity to pass to the secondary system. The FSAR analysis is based on complete severence 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 Handling Accident The mechanical damage type of accident is considered the maxieum potential source of activity release during fuel handling activities. The primary con-cern is over radiological releases, which are independent of core loading; therefore, the results of the FSAR are applicable to all reloads.

7-5 Babcock & Wilcox

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 assecbly must occur.

Such a failure could cause a pressure differ-ential to act on a control rod assecbly and rapidly eject the assembly from the core.

This incident represents the most rapid reactivity insertion that can be reasonably postulated. The values used in the FSAR and densification report at BOL conditions

-1.17 = 10-5 Ek/k *F Doppler coefficient, +0.5 = 10~"

ak/k *F noderator temperature coef ficient, and ejected red worth of 0.50 ak/k represent the maximum possible transient.

The use of a 0.05Z Ak/k raximum ejected rod worth can be justified for Cycic 4 by comparing the key nuclear and thermal-hydraulic parameters pertinent to the eject ion accident between Oconee 21? and Oconee 1, Cycle 4.

The transient cal-culations perfor=ed for 0.65% Ak/k HFP ejected rod worth limit for Oconee 2 are equally applicable to Oconee 1. Cycle 4.

The values of -1.17 w 10~5 Ak/k *F Doppler coef ficient, +0. 5 4 Ak/k *F moderator coef ficient, and a radial local power peaking f actor of 1.78 were used in the analysis of the 0.65?. Ak/k rad ejection accident of Oconce 2.33 The corresponding Oconee 1, Cycle 4 para-metric values. -1.45 e 10-5 ak/k *F Doppler and -1.00 a 10-4 *k/k *F moderator temperature coef ficient s, are both more negative than those used for Oconee 2.

The radial-local power peaking facter is 1.783 for Oconce 1. Cycle 4, which is essentially the same as that used for Oconee 2.

The use of a G.652 ak/k maximum ejected rod worth for Oconee I has also been evaluated by comparing significant Cycle 4 thermal-hydraulic parameters with those used for the analysis of Unit 2.13 This comparison reveals the following:

1.

The initial minimum DNBR is significantly higher for Oconee 1. Cycle 4, primarily because of the use of the BAW-2 CHF correlation, as discussed in section 6.2, the re-ev aluation of RCS flow rate discussed in section 6.1, and the elimination of the

• ant valve penalty, discussed in section 6.3.

2.

The initial het spot fuel temperature is lower than that assumed in ref-erence 13, resulting primarily f rom the reduction in the power spike factor discussed in references 2 and 12.

7-6 Babcock s.Wilcox

f 3.

Other thermal-hydraulic parameters used in reference 13 are applicable e

and conservative for evaluation of the 0.65: 'k/k ejected rod worth for Cycle 4 operation of Oconee Unit 1.

Therefore, the =axicum ejected rod worth limit of 0.65 Ak/k for Cycle 4 is con-sidered acceptable, based on comparisons of key nuclear and thermal-hydraulic parameters.

7.12.

Maximum Hypothetical Accident There is no postulated necnanism whereby this accident can occur since it would require a multituJe of f ailures in the engineered safeguards. 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.

Therefore, the results reported in he FSAR are applicable for all reloads.

7.13.

Waste Gas Tank Rupture The waste gas tank was assumed to contain the gareous activity evolved from degassing all the reactor coolant following operation with 1% defective fuel.

Eupture 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 loading; therefore, the results repcrted in the FSAR are applicable to any reload.

7.14.

I.nCA Analysis A generic LOCA analysis has been performed for B&W's 177-FA lowered-loop NSS, using the Final Acceptance Criteria ECCS Evaluation Model. This study is re-1 ported in BAW-10103 ". The analysis in BAW-10103 is generic since the limiting values of key parameters for all plants in the category were used.

Furthermore, the average fuel temperature as a function of the LHR and the lifetime pin pressure data used in the BAW-10103 LOCA limits analysis are conservative com-pared to those calculated for this reload.

Thus, the analysis and the LOCA limits reported in BAk'-10103 provide conservative results for the operation of Oconee 1, Cycle 4.

7-7 Babcock & Wilcox

The bounding values for allowable LOCA peak linear heat rates for Oconee 1 Cycle 4 fuel are shown below.

Allowable peak Core elevation, ft LHR, kW/ft 2

15.5 4

16.6

[

6 18.0 8

17.0 10 16.0 Table 7-1.

Comparison of Kev Parameters for Accident Analysis FSAR and Predicted Parameter densif'n report value Cycled 4 value Doppler coesficient, BOL, 4k/k *F

-1.17 x 10-5

-1.45 x 10-5 EOL, ?.k/k *F

-1.33 x 10-5

-1.55 x 10-5 Moderator coefficient.

BOL, 4k/k *F

+0.5 = 10-4

-1.00 x 10-"

EOL ak/k *F

-3.0 10-'

-2.55 x 10 -

All rod 'aank worth, % 2k/k 10 8.36 Init ial laron conc, ppm 1400 1145 Boron reactivity worth, cold 75 84 ppa /1% 4k/k Max ejected rod worth.

0.50 0.37

% t.k/k Dropped rod worth, ilFP, 0.46 0.20

% Ak/k Babcock & Wilcox 7-8

4 I

l

{

S.

PROPOSED MODIFICATIONS TO TECliNICAL SPECIFICATIONS The Tec.hnical Specifications have been revised for Cycle 4 operation. Changes were the results of the follcuing:

1.

L' sing an. jected rod sc rth of 0.65% ik/k rather than 0.50*. 2k/k, as dis-cussed in section 7.11.

2.

Specifying APSR position limits in addition to usual regulating control r(d and imb.ilance lir.alts for ECCS. The APSR position limits will provide

, additional control of power peaking and assurance that LOCA kW/ft limits are not exceeded.

t* sing 106.')% of desip flow rather than 107.6%, as discussed in section 3

6.1.

4.

The FIEll; tenputer codeJ = 19 used in setting the Technical Specification linits.

3.

D*;BR-related RPS ll:21i bases have been revised to incorporate a projected fuel red boe penalty as discussed in sections 6.2 through 6.4 6

An analysis incorporating the ef fects of fuel rod bow on core parameters.

The rod bow evaluation was perforned using the methods and procedures de-scribied in ref.rence 19 with the power spike calculated so that at least W. of the fuel rods will not exceed a given power spike factor at a 95%

confidence level.

7.

The penalty on core coolant flow due to an assumed open vent valve has been eliminated based on a vent valve surveillance program performed dur-ing each refueling shutdown.

8.

The generic IDCA kW/f t limit curve shown in Figure 8-17 was used to set all limits as.13 iated with ECCS criteria.

Based on the Technical Specifications derived from the analyses presented in this t i. port, the Final Acceptance Criteria ECCS limits will r.ot be exceeded.

g_i Babcock & Wilcox

nor will the therral design criteria be violated. Figures H-1 through 8-13 and 8-17 illustrate revisions to previous Technical Specification safety limits:

Figures 8-14 through 8-16 illustrate limits not previously included in the Technical Specificat! con.

8-2 Babcock 8. Wilcox

Figure d-1.

Oconec 1, Crcle '. - Care Protection j

Safety Lir.its 2.t00 2200 a

i 5

0 E

2003 l

5 e

1800 7

1600 500 580 600 620 640 Re3Clor C00lant Outlet Temperature, F 1,

8-3 Babcock s.Wilcox

l l

1 Figure 6-2.

Oconee I, Cycle 4 - C.$re Pratection Safety Limits THERNAL P0eER LEVIL. %

_ _ 120

(-17.2.132)

__,g (20.1.112)

ACCEPTAaLE 4 t 40.98) 100

.taF I

(FEhAleth

__ 90 80 t53 201

( 40.721

-- 70 3 & 4 PJIF OPERAil0N

__ 60 ISO 53 31

-- 50 ACCEPTABLE i 40.44 2) 2.3 & 4

-- 40 PulF OPERAil0N

-- 30 (50.25 Op

__ 20

-_ 10 I

I I

l t

g

-60

-40 20 0

+20

+40

+E3

~

Reactor Poser innatance.,

CURVE RE ACTOR COOL ANT FLos (GPu)

I 374880 2

280035 3

183690 4*

204310

'THE FLUI FL0s SETPolNT l0R 2/0 PUNP OPERAil0N uuST BE SET AT 0.949 Babcock s. Wilcox

8_f,

d Tieure ti-3.

Ocence 1. Cycle '. - core PrettetIan S.sietv !.inits 2400

1. EIII AIII

{22C3~ CPERAfl09

{

i 0

?

.I

tt

/

s-2 e5 I

3 I

- se :

I

!C'0 -

550 520 600 620 640 Reacter Ccolant Gatlet Te peratuse F Cho.I aiact:a C00' ANT FLOS iGPut PueER PuePS OPERATING titPE OF Llutil 8

U 436C 4 00 5-

112, 4

Dwel

!!3"E 4 74 7 i 66 y, 3

gggg, 3

!!3590 i49 0 i Sg ga 2

,gggggy, 105 5 Di fir 51 CORE DE$1GN Flee l

i l

i i

l 6-5 Babcock & Wilcom

+

t i gi;t.

5-?.

Oi. ir.t v 1. Cyc's -. - I ret ce t !ve 3" stem "tixi..:.

4lowt.*le 5etpcints THERNAL poser LEVEL. $

_ _ 120

__ 110 (105.5

.g g

__ 100

+7

• N l

__ 90 t-40.84 ACCEPTAELE I

(30.87i 4 PUNP l

OPERATION

__ 80

,7g g, 1

i

__ 70l l

I CCEPTABLE

. 40 57 3 3 & 4 PUWP l

-- 60 l (30.60 3 OPERATION l

f a 51. 7 6*

l

-- 50 1

I g

__ 40l ACCEPTABLE l

(-40.30 23 23&4 l

(30.33 3)

P P

-- 30l g

OPERAfl0N l

l

__ 20 g

~l o

o "I

__ 10 I.

el ir l l

-40

-20 0

20 40 Reactor Poser 1:Dalance. ',

• THE FLUX FLOW SETPOINT FOR 2 0 PUNP OPERATION MUST BE SET AT 0.949 s _ (,

Babcock & Wilcox

9 7

Figure 6-5.

Oconee 1, Cycle *. - Rod Pcsition Limits f'or Four-Pu=p Operation Frs= 0 to 100 (:10) EFPD l

I 100

- CP[sait0N sk (94 102 (124.3028,,

,,4225 9 102) 1M15 f[G404 t$

POIEI kEIII 90

- MI A* LOS[0 CUTOTF t 174. 98 8 225 9.90)

PESTRICTfD ED RIGlog (163 80)

(230 9.60) ej 7;g 3

Ri$i8tCTE0 REGlou waqG g 10

\\

(164.78)

(235 9*70) l g

Lluli h 60 (159.60 v240 I 60:

N (300.60)

E 50

'42.53

. f D 50 6 PERulSSIBL E 3 gg CPigail % PEC40N 2

30 20 E

e3.35e 39 15 p 10 -

.0 01

'I I

I I

f I

00 O

100 200 300 Pos inse... Witnor.en 0

25 50 75 100 0

25 50 75 160 i

f t

i e

s I

soup 5 Grous 0

25 50 75 100 l

f I

t i

Sr:.a G R w. e..

., i,... t, i u, su, oi i.,. sin.....

., s,.u,, 5.s.. 2 M7 Babcock A Wilenw

- - n Figure 8-6.

Oconee 1. Cycle 4 - Rod l'osit ion Limit s for Four-Pump Operation Frc:s 100 (:10) to 250 (-10) EFPD I' '

103 REST 54CTED gg _ CP[R4110N 84 IMll pge[R ((v[L g (t,4 t.

< 4225 3 SDn

1. EGION 15 =CT Cu1CFF REG'C4 83 _ atttetD il65 I 80 t232 9 Boa 73 s+J100sh maasin (tult i164 3 70s i:35 9,7De a

3 C3 L RistalcifD REGION

'959 I 33*

4 30D 6Di

[ 50 a 126.50 s d'

PERul15tBLE OPf 611tNG eEGIO4 30

C J iS4 th e IC

.CC' a

f t

8 t

t i

f I

i i

C 100 200 333 Roc Inces ; eatnsraen 3

25 50 75 100 0

25 50 Th 108 I

f f

e I

t

?

I f

Croas b Group 7 0

25 SC F5 100 I

e f

l Grow 6 Ras neces,5 Ine 3,er:entage saa of tie estMraent at GtGdp1 $ 6 ant 7 a-3 Babcock & Wilcox

Figure 8-7.

Oconee 1, Cycle 4 - Rod Position Limits for Four-Pump Operation After 250 010) ETPD

,,4258.I 102) 100 P0ett LivEt CattFF

,t251 8 90 90 CPERaficia lh THl1 REG 10il 15 NOT atLOSED 80 RESTRICTIS

'246 8.801 n 'EIGie 10

.241 8 70)

E EO e236 8 68 50 e 12 7. 50 )

j d'

SituiD004 NARGIN PERul55tsti DP8atilNG REGl0N 30 1Iuli 10 et00 15 IS 0 Os C

1 0

103 200 300 Ros Inces. % setndraen 0

25 50 15 I00 0

25 50 75 100 t

I i

1 e

i t

i i

i Group 5 Grcup 7 0

25 50 75 100 i

t t

I g

Group 6 11:3 enies is tne percentage sus et tne estparaeal el Greams 5 E ans 7 B-9 Babcock & Wilcox

4 0

Figure S-8.

Oconce 1, Cycle 4 - Rod Position Limits for Two-and Three-Pump Operation From 0 to 100 (r10) EFPD i34 102)

(159.102)

(169.1028 (2' l. '02) 130

~ CPERATION IN THl3 y agg 3 pgy RE!!RICTED REGION is hai Att0eED CPERail04 REging

~ 881H 2 O!I 3 PuePS ei5fGICifD I (236.891 1864 236 Il0N

=

] BD IMIS REGIC AEsta:CTED IN 8241.76)

T115 him 8 ISS. 76 n f 10 - ul%tuur (300 76s Sm:IDC9N LINii b EL

e

, $3 - 444.50s 43 -

=

PERIIISS'PLE PPEstilmE REGIP%

33 -

~o

, 23

{

h,. 0 15 10 0

_t i

t I

t t

C 10 700 300 403 Inses,, setn3raen 0

25 50 75 800 0

25 50 15 ICO I

I I

I e

i I

e Group 5 Group 7 0

25 50 15 800 I

f f

I l

Grcup 6 RJ3 anuts as Int peittalagt 549 at the eithdrasal of Grawas 5.6 aPO 1 8-10 Babcock s. Wilcox

w Figure 8-9.

Oconee 1. Cycle 4 - Rod Posit ion Lic:it s for Two-and Three-Pump Operation Frea 100 (210) to 250 (:10) EFPD (188.102)

(231.102) 100 OPERAil0N IN THIS REGION RESTRICTED 15 NOT ALLO.E0 REGION FOR 3 PUNP 90

,235 89) e 3

OPERATION l 88 Suur00 N =ARE:N timir 24: 76)

E 70 (300.76:

m E 60 w

50 (126.50)

Q*

5#

~

PE4WISSIBLE OPERAil%G E

REGION 30

• 20 e.

10 (84 On O

I I

t t

t 9

l 9

g 9

g 0

100 200 300 Roa inces 1 Withdraert 0

25 50 75 300 0

25 50 75 100 E

I I

I I

I I

t t

Group 5 Croup 7 0

25 58 75 lb0 I

I I

i i

Grous &

Roa inden is the percentage sus of tee setndrasal at Groups 5.6 aa: 7 8-11 Babcock & Wilcox

)

Fir,ure S-10.

Oconee 1. Cycle le - Acd Positien Li: sits for Tw-and Three-Pu=p operation After 250 (:10) EFPD

. ' 8.102e 104 IN W HM LD.M

. Jag g 102, 100 15 NOT Att0eED slin 2 CR 3 4GO PuuPS FJt 2 & 3 90 M

pump 281 8.89 i e

?

IWT:

3 gg

. [R4T104 y

RESTRICIED

'235 8 76l a

tw idts

,3 Ib'

[-

$4cf 00e% s4RGiu Listi y

60 1

g

.327 50s 2

2 40

1. 0

~

PEPeSSIStf CPER4Te%G REGt04 23

.13G 15, 10 -

6 0 Di C

f f

f i

i

?

I i

i i

O 102 2c3 30C Roo ince.t W e tr.$ r aan C

25 S3 F5 500 0

25 50 15 100 I

f I

f t

i g

g g

Grcca 5 Group I o

25 5e 75 ioO L

f t

e i

Geen 6 R n e1res is tme percentage sue of tr.e esteceaeas at Groups 5 6 ans 7 3-12 Babcock s. Wilcox

}

Figure 8-11.

Oconee 1 Cycle 4 - Operational Po er 11ance Envelope for Operation From o co 100 (:10) El'T POWER, i of 2568 Met RESTRICTED REGl0H 110

(-17.3.102)

(.0.102) 100 --

l

( 15.3.90) 90 (8.8.90) 80 70 PERMISSIBLE OPERATING 60 RANGE 50 40 30 20 10 I

1 0

I I

-20

-10 0

+10

+20 Core Imnalance. '

8-13 Babcock & Wilcox r

Figure 8-12.

Oconee 1 Cycle

- eperatietal Pe er

!= balance Enzelope fer Operrition Fres 100 (?t0) to 250 (r10 EFP)

POWER. t 0F 2568 nt RESTRICTED REGION 110 G.J,102)

(+33.102) 100 1

(-17.0,90) 90

(+3.8.90) 50 70 60 ---

50 I

1 40 I

l 30 20 10 i

i l

L I

I I

-20

-10 0

+15

+20 Core imoalance. 5 I

l Babcock 8. Wilcox 8-l '.

\\

l Figure 8-13.

Oconee 1, Cycle 4 - Opera;ional Pc.er Inbalance Envelope for Operation After 250 (-10) EEP3 i

POWER, 5 0F 2568 185t I

110 j

RESTRICTED REGION

(-23.1.102) 00 f(15.8,102) f

(-21.2.90) 90 (14.9,90) f PERWISSI8l.E 80 OPERATING REGION 70 60 SC 40 30 --

20 10 i

l 1

1 I

I

-30

-20

-10 0

+10

+20 t

Core Imoalance. 5 8-15 Babcock a. Wilcox

t Figure 8-14.

Oconee 1, Cycle 4 - APSR Pesition Limits for j

operation Frorn 0 to 100 (210) EF?D i

100 i(37.0.102) 90

, (39.4.90)

RESIRICTED REGION 80 (39.4.80) 5 70 (46.4.70)

M (90.4.60)

[

60 50 T2 40 PERMISSIBLE 30 20 10 -

0 l

I I

I I

I I

I I

O 10 20 30 40 50 60 70 80 90 100 APSR. G Withdrawn 3-16 Babcock & Wilcox

Figure 3-15.

Ocor.ee 1. Cycle 4 - APSR Position Limits for Operation Fron 100 (:10) to 250 ( 13) EFPD (14.0.102)

I 2 IU2I 100

[RESTRICTE g

REGION

,(34.6.90)

RESTRICTED 90 (11.6,90)

REGION (34.6,80) 60 W (9.2,80) 70 (3.6,70)

(41.6,70)

(85.6,60) g

[

60 w (3.6,60) b0.60)-

(100.60)

L E

53 l

PER111SSIBl.E y

40 OPERATING REGION E

30

~

l

~

20

~

10.-

~

O I

I I

I I

I I

7 l

0 10 20 30 40 50 60 70 80 90 100 APSR. ', fitndrawn 3-17 Babcock 8. Wilcox e-m

Figure 6-Id.

Oconce 1. JW le

.\\PSR Pesition I.isits for Operatien Mter 250 t:10) EFPD (31.0.102)

(32,0,102)

RESTRICTED REGION 90 (B.6.90)

,(34.4,90)

(6.2.80)

,(34.4.80) 80

- i 70 (6.1.70)

(34.4,70)

(85.4,60)

E 60 <w (6.1.60)

(0.60)

(100,60) 50 PERMISSIBLE f.

40 OPERATING E'

REGION 30 20 10 i

)

O I

I I

I I

I I

I I

O 10 20 30 40 50 60 10 80 90 100 APSR, *. Witturawn Babcock & Wilcox 3 gg 8

Figure 8-17.

IECA-Limited.txir.um Allowable Linear lieat Rate for Oconee Unit 1 20 i

6 6

4 i

a s

a a

i 16 a

3

d 16 3

.s 14 2

Generic FAC BAW-lOlO314 g

.E x2 12 -

^

] (,

f I

f f

f f

f a

f f

f G

6 8

10 1.'

Axial Location of Peak Power From Bottom of Core, ft 8-19 Babcock s. Wilcox

i l

1 1

l 6

)

9.

STARTUP PROGRAw.

The planned startup testing associated with core performance is described be-low.

These tests verify that core performance is wit hin ti.e assu=pt ions of the safety analysis and provide the necessary data for continued safe plant operatlon.

Pre-Crit f eal, fests 1.

Cont rol rod drive trip t ine testing.

Zero Power rents 1.

Crit ical boron concent rat ion.

2.

Temperature reactivity coefficient.

3.

Control rod group worth.

4.

Ejectcd red worth, i

l Power Tests 1.

Core power distribution verification at approximately 40, 75..ind 100'. FP norcul cont rol rod group configuration, l

Core power distribution verification at approximately 40 FP with worst-I case dropped rod fully inserted.

1.

Incore/out-of-core detector imbalance correlation verification at approxi-na tel:. 7 5% FP.

4.

Power Doppler reactivity coefficient at approximately 100: FP.

5.

Teraperature reactivity coefficient at approximately 100% FP.

4 9-1 Babcock & Wilcox

REFERE.NCFS ueonee Nuclear Station, t* nit 1 - Final Safety Analysis Report, Docket 50.'69 Ocence t' nit 1. Cycle 3 Reload Report, BAW-1427, Babcock & Wilcox, Decem-ber 1975.

Friel Assembly Stress and Deflection Analysis for loss-of-Coolant Accident and Selsr>Ic Excitation, ILW-10008, Part 2, Rev. 1, Babcock & Wilcox, June 1974.

Program to Determine In-;tesctor Performance of B&W Fuels - Cladding Creep t:o l l.y s e, P.W-10084P, Rev.

1, Babcock & Wilcox, October 1976.

ocenes 1 Fuel Densification Report, SAW-1388, Rev. 1 Babcock & Wilcox,

.lulv I W 1.

C.

!). " irgan and H. S. Kao, TAFY - Fuel Pin Temperature and Gas Pressure Analysis, B.W-10044, Babcock & Wilcox, May 1972.

B. J.

Bae.cher and J. V. Pegram, Babcock & Wilcox Model for Predict ing In-Reactor IVnaification, BAW-10083P, Rev. 1, Babcock & Wilcox, November 1976 (Proprietary).

Cerrelation of Critical Heat Flux in a Bundle Cooled by Pressurized Water, P.AW-10000A, B.iheock & Wilcox, June 1976.

}

K. W. 11111, e,t al., Effects on Critical Heat Flux of local Heat Flux Spike or Local Flew Blockage in PWR Rod Bundles, 74-WA/HT-54, ASME Winter Annual Meeting, L w York, November 1974.

30 CHF - Critical Heat Flux Correlation for CE FA With Standard Spacer Grid -

Part 2. " Nonuniform Axial Power Distribution," CENPD-207. Combustion En-gineering, Juae 1976.

72.2. Z Ji~. 3 T Z.1 2 ~ ~ C ~.Z Z.!. T11 H " ~ -

sabcock a witcox

r II " Core Physics Methods Data Used as Input to LOCA Analysis," XN-75-43, August 1975 and letter D. A.

Bixel, Consumers Power, to R. A.

Purple.

April 5, 1976.

' Geonee 1, Cycle 2 Reload Report, BAW-1404, Rev. 1 Sabceck & Wilcox, October 1974 a

' Oconee 2 Fuel Densification Report, BAW-1395, Babcock & Wilcox,. lune 1973.

I' ECCS Analysis of B&W's 177-FA lewered-leep NSS, SAW-10103, Rev. 1. Babcock

& Wilcox, September 1975.

II A.

Schweneer to K.

E. Suhrke, Letter, Notember 1975.

I*

K.

E. Suhrke to A. Schweneer, Letter: "S&W Operating Experience of Reactor Internals Vent Valves," August i., 1975.

3#

FIA.'!E - Three-Dimensional Neded Code f or Calculating React ivity and Power Distributiens, BAW-1012t. A. Babcock & Wilcox, August 1976 16 Verificat ion of Three-Dimensional FIRIE Code, BAW-101.!5, Babcock & Wilcox, August 1976.

W.

O. l'a r ke r ( Duke Powe r Co. ) t o B. C. Rusche (USNRC),i.etter. February 27, 1976.

.< + ~

i a

l i

I l

A-2 Babcock s. Wilcox I

- n _ _ _m _ _ __._

.. _ _