ML19322C147

Cycle 4 Reload Rept.

ML19322C147
Person / Time
Site:	Oconee
Issue date:	08/31/1978
From:	BABCOCK & WILCOX CO.
To:
References
BAW-1491, NUDOCS 8001090560
Download: ML19322C147 (50)

Text

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_BAV-1491 3067 August 1978

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' Nuclear Power Generation Division ,

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CONTENTS f

Page 4 1. INTRODUCTION AND SUMMnRY . . . . ... .. ...... ... .... 1-1

2. OPERATING HISTORY

. . . . . . . .... . . . ....... .... 2-1

3. CENERAL DESCRIPTION

. . . . . . .... . . . ... . .... ... 3-1

, 4. FUEL SYSTEM DESIGN .

1 . . . . . . ...... . ........ ... 4-1 1 4.1. Fuel Assembly Mechanical Design . . . . ... .... .... 4-1

[} 4.2. Fuel Rod Desi's . . . . . ..... . . ........ ... 4-1 4.2.1. Cladding Collapse l] 4.2.2. Cladding Stress

..... . . ........ ... 4-1 4.2.3. Cladding Strain

. .... .. . . ....... ... 4-1 )8

4. 3. Thernal Design .

. ....... . ..... ..... 4-1

. . . . . . ... .. . .... ... .... 4-2 4.4 Material Design . . . . . .... .. . ... . .... ... 4-2 4.5. Grerating Experience . . . .. .. ... .... ... .... 4-2

5. NUCLEAR DESIGN .

. . . . . . . . . ... .. . . .. ... ..... 5-1 5.1. Physics Characteristics 5.2. . .. .. . . . ... . ... .... 5-1 Analytical Input . . .

5.3. . . . ... . . . ... . ... .... 5-2 Changes in Nuclear Design .... .. . .... ... .... 5-2 6 TliERMAL-HYDRAULIC DESIGN . 1

. . . .... . . . ........ ... 6-1 -

7. ACCIDENT AND TRANSIENT ANALYSIS ...... . .. .. ... .... 7-1 7.1. Cencral Safety Analysis . ..... . . ... . . . . ..... 7-1 7.2. Accident Evaluation . . . . .. ... . ........ ... 7-1 B. f PROPOSED MODIFICATIONS TO TECHNICAL SPECIFICATIONS . . ... .... 8-1 l l 9. STARTUP PROGRAM - PHYSICS TESTING .. ... . .. . . .... ... 9-1 9.1. Precrit'. cal Tests . . . . . ... .. . ....... .... 9-1 9.1.1. Control Rod Trip Test .. . . . ....... .... 9-1 9.1.2. Reactor Coolant Flow . .. ... ........... 9-1 g 9.1.3. RC Flow Coastdown . ...... . . ......... 9-1 9.2. Zero Power Physics Tests . .... ... . ..... ..... 9-2

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9.2.1. Critical Boron Concentration . . ....... .... 9-2 9.2.2. Temperature Reactivity Coefficient . ....... .. 9-2 9.2.3. Control Rod Group Reactivity Worth . . . . . . . . . . 9-2 9.2.4. Ej ec t ed con t rol Rod Reac t iv i ty Wo r t h . . . . . . . . . 9-3 l

Babcock s,Wilcox

I CONTENTS (Cont'd) l Page 9.3. Power Escalation Tests . .. . .. ... . ... . .... ..

9.3.1. 9-3 Core Power Distribution Verificacios at .40, 75, and i 9.3.2.

100% FP With Nominal Control Rod Grcur Configuration . 9-3 Incore Vs Excore Detector Imbalance Correlation Verif. ation at % O% FP .. ... ... . ...... *-5 9.3.3. Temperature Reactivity Coefficient at %200* FP . ...

I 9.3.4 Power Doppler Re. activity Coefficient at ICD % FP ...

9-5 9-5 s,

9.4 Procedure for Use When Acceptance Criteria Are .W t Met .... 9-5 REFERENCES .

. . . . . . . . . .. . . .. .. .... ... .... 10-le I

List of Tables i Table 4-1. Fuel Design Parameters and Dimensions 4-2. Fuel Thermal Analysis Parameters . ... . ... .. ..... 4-3 4-3. ... .. . ... . ...... 4-4 5-1. Operating Experience Ur.ing Mark B Fuel Assembly ........ 4-4 Oconee 2. Cycle 3 and 4 Physics Parameters . .... . ...... .

l 5-2. 5-3 Shutdown Margin Calculation for oconee 2 Cycle 4 . . ...... 5-4 h-l. Thermal-Hydraulic Design Conditions 7-1. . ... ... .... .... 6-3 7-2. Compar' in of Key Parameters for Accident Analysis .

Bound g Values for Allowable LOCA Peak Linear Heat . ...... 7-3 Kate= .... 7-3 List of Figures Figure 3-1. Core Loading Diagram - Oconce 2. Cycle 4 ...

3-2. ...... ...

Oconce 2, Cycle 4 Enrichment and Burnup Distribution . ..... 3-3 3-3. Control Rod Locations for Oconee 2, Cycle 4 .. ...... ... 3-4 5-1. 50C (4 EFPD) Cycle 4 Two-Dimensional Relative Power 3-5 J

Distribution - Full Power, Equilibriun Xenon, Normal Red Positions, Croups 7 and 8 Inserted ............

3-1. Core Protection Safety Limits . ... 5-5 8-2. .. ... . . .... .. .... 8-2 8-3.

Protective System Maximum Allowable Setpoints . ..... .... 8- 3 Rod Position Limits for Four-Pump Operation From 0 to 250 1 10 EFPD - Oconec 2, Cycle 4 . .

6-4. . ... . .. ...... . 8-4 Rod Position Limits for Four-Pu=p Operation After 250 2 10 EFPD - Oconee 2. Cycle 4 S-5. . .... ... .. .... .. .... 8-5 Rod Position Limits for Two-250 1 10 EFPD - Oconee 2 Cycle an:i Three-Pump Oper;ttion From 0 to 4 .... .

.......... 8-6

_ 111 ., Babcock s Wilcox

Figures (Cont'd)

Figure Page 8-6.

Rod Position Limits for Two- ar.d Three-Pump Operation After 250 10 EFPD - Oconee 2, Cycle 4 8-7. . . . . . . . . . . . . . . . S-7 Operational Power Imbalance Envelope for Operation Fron 0 to 250 2 10 EFPD - Oconee 2, Cycle 4 . . . . . . . . . . . . . . . 8-8 8-8.

Operational Power Imbalance Envelope fc.r Operation Af ter 250 1 IG EFPD - Oconee 2, Cycle 4 . . . . . . . . . . . . . . . . . .

8-9. 8-9 g APSR Position Limits for Operation From 0 to 250 1 10 EFPD -

Oconee 2, Cycle 4

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8-10. APSR Position Limits for Operation Af ter 250 ! 10 EFPD -

Oconee 2 Cycle 4

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k 1. INTROC.CCTION AND S3 MARY I

This report justifies the operation of the fourth cycle of Oconee Nuclear Sta-tion, U.11t 2, at tbe rated core pcwer of 2568 MWt. Ir.cluded are the required analyses as out11:.ed in the USNRC docunent " Guidance for Proposed License Amendmerts Relat Lng to Refueling," June 1975.

To support cycle 4 cperatien of Oconee Unit 2, this report e= ploys analytical techniques and design bases established in reports that were previously sub-nitted and accepted by the USNRC and its predecessor (see references). '

A brief summary of cycle 3 and 4 reactor parameters related to power capability is included in section 5 of this report. A12 of the accidents analyzed in the FSARI have been reviewed for cycle 4 operation. In those cases where cycle 4 characteristics were conservative conpared to those analyzed for previous cy-eles, no new accidest analyses were performed.

l The Technical Specifications have been reviewed, and the modifications required for cycle 4 operation are justified in this report.

2

( Based on the analyses performed, which take into account the postulated effects of fuel densification and the Final Accepta3ce Criteria for Emergency Core Cool-Ing Systems, it has been concluded that Oconee Unit 2 can be operated safely.for cycle 4 at the rated power level of 2568 MWt.

Because of performasce anomalies observed at other B&W plants, orifice rod as-sed lies will not be used in Oconee 2, cycle 4 This change from normal prac-tice has been accounted for in the analyses performed for cycle 4. In addition, retainer assemblies will be installed on two fresh batch 6 fuel assemblies con-taining regenerative neutron sources.

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1-1 Babcock s.Wilcox

2. OPERATING HISTORY The reference fuel cycle for the nuclear and thermal-hydraulic analyses of the ,

fourth cycle of Oconee Nuclear Station Unic 2 is the currently operating cycle

3. Cycle 3 achieved initial criticality on August 26, 1977, and was escalated to 40: power on August 28, 1977. The 100% power level of 25'38 MWt was reached on January 18, 1978. No operating anomalies occurred during cycle 3 operation that would adversely affect the fuel performance in cycle 4 during the design length of 292 EFPD. No control rod interchanges are planned for cycle 4 Con-trol rod group 7 will be withdrawn at 250 (210) EFPD of operation.

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3. CENF.RAL DESCRIPTION The Oconee 2 reactor core is described in detail in Chapter 3 of the F~AR.I The core consists of 177 fuel assemblies, 173 of which have a 15 by 15 array containing 208 fuel rods, 16 control rod guide tubes, and one incore instru-ment guide tube. The fuel consists of dished-end, cylindrical pellets of ura-nium dioxide clad in cold-worked Zircaloy 4. The other four FAs in cycle 4 are demonstration 17 by 17 EAs - two Mark C and two Mark CR. All fuel assen-blies in cycle 4 except the 17 = 17 demonstratics assemblics maintain a con-stant nominal fuel loading of 463.6 kg of uranium. The undensified nominal active fuel lengths, theoretical densities, fuel and fuel rod dimensions, ar.d other related fuel paramerets are included in Tables 4-1 and 4-2 of this re-port.

Figure 3-1 is the core loading diagram for Oconee 2, cycle 4. Nine once-burned batch 1 assemblics with an initial entichment of 2.06 wt : 235g ytti be reloaded into the core. Batches 4, 5, and SA with initial enrichments of 2.64, 3.03, and 2.53 wt % 235 U, respectively, will be shuffled to new locattens.

Batch 6, with an initial enrichment of 2.91 wt : 235U, will occupy the core pe-riphery and eicht interior locatior.y. Figure 3-2 is an eighth-core map show-ing the assemb i burnup and enrichment distribution at the beginning of cycle 4.

Reactivity control is supplied by 61 full-length. Ag-in-Cd control to s and soluble baron shim. In addition to the full-length control rods, eight partial-length axial power shaping rods (APSRs) are provided for additional control of axial power distribution. The cycle 4 locations of the 69 control rods and the group designations are indicated in Figure 3-3. The core locations of the total pattern (69 control rods) for cycle 4 are identical to those of the initial cy .

cle described in Chapter 3 of the FSAR.I However, the group designations dif-fer between cycle 4 and the first cycle to minimize power peaking.

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The no .inal systeca pressure is 2200 psia, and the densified nominal heat rate is 5.80 kW/ft at the rated core power of 2568 Mut, r

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Figure 3-1. Core Loading Diagra:s - Oconee 2. Cycle 4 F Fwl Treme'er Cem.1

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Figure 3-2. Oconee 2, Cycir.,r. Enricic. cat and Eurcup 31stributio 8- 9 10 11 12 13 14 15 2.06 2.64 2.53 2.C 6 2.91 l 2.64 2.61 2.91 11 12.965 12,152 12,188 12.965 0 18.296 18.057 0 2.91 2 64 3.03 2.64 3.03 2.64 K 2.91 0 21.460 11,21A 18,505 11,222 17, f.f.8 0 2.64 3.03 3.03 2.64 2.91 2.fM L

p' 16.482 9.769 7,325 17,514 0 0' 3.03 2.64 3.03 2.91 M

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Figure 3-3. Co.. trol Rod Locat is r. f or Oconee 2, Cycle I.

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CROUP NO.CF ROOS F;,NCTI Cti l 12 SAFETf 2 8 SAFETT 3 8 SAFET?

4 8 SAFETT 5 9 CONTR3t 6 8 CONTR2t 7 8 C0%1RCL 8 8 APSRs TOTAL 69 3-3 Babcock & Wilc0x 4

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4. FUEL SYSTDi DESIGN

.e v 4.1. Fuel Assembly Mechanical Design The types of fuel assemblies and pertinent fuel design parameters and dimensions for oconec 2, cycle 4 are listed in Table 4-1. All Mark B fuel assemblies are identical in concept and ar. mechanically interchangeable. The Mark CR demon-stration asnmblies of batch 5 are mechanically identical in function to the Mark C demonstration assemblies of batch 4 All results, references, and iden-f tified conserv..tivas presented in section 4.1 of the previous Oconee 2 reload 2

report are applicable to the cycle 4 reload core.

4.7. Fuel Rod Design

_4_. 2.1. Cladding Collapse 4

I Creep collapse analyses were performed for three-cycle assembly power hi stories.

, The batch 4 fuel is more limiting than the other batches becauses of it' pre-k vious incore exposure time. The batch 4 asse.bly power histories were analyzed and the most limitir.g Mark B and Mark C assc=blies determined. The power his-tories of these assemblies were used to calculate the fast neutron flux level for the energy range above 1 McV. Both Mark B and Mark C creep collapse times

, were determinul to be more than 30,000 EFPH (effective full-power hours), which is longer than their maxirum projected incore residence time (see Table 4-1).

The creep collapse analyses were performed based on the conditions set forth in referer.ces 2 and 3.

4.2.2. Cladding Stress

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The batch 1 reinserted fuel is the most limiting batch from a cladding stress viewpoint owing to its lower propressurization and density. The batch 1 fuel has been analyzed and documented in the Oconee 2 Fuci Dessification Report."

} _4.2.3. Cladding Strain The-fuel design criteria specify a limit of 1.0% on cladding plastic circum-ferential strain. The pellet design is established for cladding plastic strain 4-1 Babcock 4.Wilcox

s of less than 1% at maximum values of design local pellet burnup and heat gen-erstion rate, which are considerably higher than the values the Oconee 2 fuel is expected to see. The strain analysis is also based en the maximus Specifi-cation value for the fuel pellet diameter and density and the Icwest permitted Specification tolerance for the cladding LD.

_4.3. Thernal Design All fuel assemblies in this core are thermally similar. The fresh bati fuel inserted for cycle 4 operation Latroduces no eignificant differ 2n fuel thernal performance relative to the batch 3 fuel discharged at t..e end of cycle 3.

The design minimum linear heat rate (LHR) capacity and the average fuel temperature for each batch in cycle 4 are shown in Table 4-2. LHR limit-attons were established using the TC2Y codes with fuel densification to 96.5%

of d ecretical density.

4.4. Material Design The batch 6 fuel assemblies are not new in concept, nor do they utilize differ-ent component materials. Therefore, the chemical conpatibility of all possible fuel-cladding-coolant-assembly interactions for the batch 6 fuel assemblies are identical to those of the present fuel.

4.5. Operating Uxperience Babcock & Wilcox ojerating experience with the Mark B 15 x 15 fuel assembly has verified the adequacy of its design. As of April 30, 1978, the experience de-scribed by Tabic 4-3 hat been accumulated for the eight operating B&W 177-fuel assembly plants using the Mark B fuel assembly, l

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Table 4-1. Fuel Design Paras 3ters and Dimensions Twice-burned FA4 Once-burned FAs Batch 4 Mark C deno F F%

Batch 5/SA Batch IB Mark CR demo hatch 6 Fuel assembly type Mark B-4 17=17 nrray Kirk B-4 Mark H-2 17=17 .irray Mark H-4 No. of nosemblics 34 2 s0/4 9 2 56 Fuel rod OD, In. 0.430 0.379 0.430 0.430 0.379 0.430 Fuel rod ID, in. 0.377 0.332 0.377 0.377 0.332 0.377 Ficxible spacers, type Spring . Spring Spring Corrugated Spring Spring Rigid spacers, type Zr-4 Zr-4 Zr-4 Zr02 Zr-4 Zr-4 Undens'fied active fuel 142.6 143.0 142.25 144.0 143.0 142.25 length, in.

Fuel pellet OD (mean 0.370 0.324 0.3695 0.370 0.324 P.3695 y specified),in.

Fuel pellet initial 93.5 94.0 94.0 92.5 94.0 93.0 density, % TD Initini fuel enrichment, 2.64 2.64 3.03/2.53 2.06 3.03 2.91 wt % 235g BOC burnup, mwd /mtU 17,778 12,152 8,941/12,188 13,302 9,769 0 CinddIng rollapse >30,000 >30,000 >30,0C0 >30,000 >30,000 >10,000

. time, I:FPil 1:ntimated rentdence 21,576 21,576 24,576 17,808 24.5/6 27,000 time (max), EFPit E

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1 Table 4-2. Fuel nemal Analysis Parameters 1

Batch IB Batch 4 Batch 5 Batch SA Batch 6 No. of a.=semblics 9 56(*} 52 4 56 Initial density, % TD 92.5 93.5 94.0 94.0 94.0 Pellet diameter, in. 0.370 0.370 0.3695 0.3695 0.3695 Stack height, in. 144 142.6 142.25 142.25 142.25 r

Densified Fuel Larameters(C}

Pellet diameter, in. 0.3632 0.3645 0.3646 0.3646 0.3646 ,

Fuel stack height, in. 141.1 140.5 140.5 140.5 140.5 Nominal UIR at 2568 5.77 5.80 5.80 5.80 5.80 '

MWt, kW/ft i Avg fuel temp at 1335 1320 1320 1320 1320 nominal UIR, F C.HR capability (cen- 19.8 20.15 20.15 20.15 20.15 terline fuel melt),

kW/ft

(* Includes two Mark C (17x17) demonstration assemblies.

U'} Includes two Mark CR (17x17) demonstration assemblies.

I")Densification to 96.5% TD assumed.

Table 4-3. Operating Experience Using Mark B Fuel Assembly L

Max assembly burnup, yg Current * **"

Reactor net electrical cycle _ Incore Discharged output %h Oconee 1 4 28,500 25.300

' 21.413,802 Oconec 2 3 28,000 26,800 16,081,241 Oconee 3 3 27.600 27,200 17,415,480 TM1-1 4 24,700 32,200 18,710,670 ANO-1 3 24,400 28,300 14,705.349 Rancho Seco 2 23.700 17,170 11,088,121 Crystal River 3 1 10,700 -

4,978,690 .

Davis Besse 1 1 4,000 -

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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 both cycles were generated using PDQ07. (The similarity between the two is to be expected since the core has nearly reached an equilibrium cycle.)

The accumulated average core burnup,will be lower in cycle 4 than in cycle 3 bec.tuse of the shorter cycle lifetime. Figure 5-1 illustrates a representa-tive relative power distribution for the beginning cf the fourth cycle at full power with equilibrium xenon and normal rod position.

The critical boron concentrations for the beginning of cycle 4 are very close to those of cycle 3. End-of-cycle (EOC) conditions vary between cycles 3 and 4, causing higher critical boron concentrations. Cs indicated in Table 5-2, the control rod worths are sufficient to maintain the required shutdown nargin.

However, due to changes in isotopics and the radial flux distribution, the hot, f u l l- powe r entrol rod worths will be less than those for cycle 3. The yele 4 elected rod worths for the same number of regulating banks inserted are lower than those in cycle 3. It is difficult to cenpare values between cycles or be-tween rod patterns since neither the rod patterns f rom which the CRA is assu=ed to be ejected nor the isetopic distributions are identical. Calculated ejected l red worths and their adherence to criteria are considered at all times in life I and at all power levels ic *be development of the rod insertion limits presented in section~8 The uaximum .cuck rod worths for cycle 4 are similar to those in I cyc!c 3. The adequacy of the shutdown margin with cycle 4 stuck red worths is demonstrated in Table 5-2. The following conservatisms were applied for the shutdown calculations:

1. Poison material depletion allowance.

2. 10' uncertainty on net rod worth.

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3. ' Flux redistribution penalty. '

5-1 Babcock s.Wilcox I

Flux redistribution was accounted for since the shutdown analysis was calcu-lated using a two-dimensional model. The shutdo.n calculation at the end of cycle 4 is analyzed at appecximately 250 EFFD (:10). This is the latest ti=e

(*10 days) in core life at which the transient bank is nearly full inserted.

After 250 EFPD, the transient bank will be aleost fully withdrawn, thus in-creasing the available shutdown margin. The reference fuel cycle sbutdown margin is presentd in BAW-1452.2 The cycle 4 power deficits from hot zero power to hot full power are similar to but slightly lower than those for cycle 3. Doppler coefficients noderator coef ficients, dif ferential boron worth, and xe,non worths are similar for the two cycles. The effective delayed neutron fractions for both cycles show a decrease with burnup.

_S.2. Analytical Input The cycle 4 incore measurement calculation constants used for computing core power distributions were prepared in the same manner as for the reference cy-cle.

5.3. Changes in Nuclear 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 ob-tain the important nuclear design para =eters for cycles 3 and 4. The opera-tional limits (Technical Specification changes) for the relcad cycle are shown in section 8.

The FLAME code was used in setting the Technical Specification limits.

The nuclear characteristics of the two ba ch 5 Mark CR (17 x 17) fuel assem-blies are nearly identical to the Mark B (15 = 15) assemblies that make up the balance of batch 5.

The two batch 4 Mark C demonstration assemblies are also comparrole to batch 4 Mark B assemblies. Therefore, the presence of the 17 r 17 demonstration FAs (two Mark C and two Mark CR) .111 not discernibly affect over-all core reactivity coefficients or performante. However, since the two Mark C and two Mark CR fuel assemblies are desonstration assemblies, standard practice dictates their placement in nonlimiting core locations during cycle 4 l

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Table 5- 1. Oconee 2, Cycle 3 ar :' Physics Para =cters c,ese 3(*) Cy.:, d6)

Cule length. ETP3 W 272

' C,cle burnwp. W d/stu 9392 913S Avera te core buresp - EOC. WJ/ett 19.554 14.1-1 Intt ant core la stag, atU S2.0 82.0 Cratical beroe - S M (ne In). Pro hP. group 8 37.5% wdI 'I 1135 ll-J

%*P. gewwps 7 a*I 8 inserted 1253 1271 hTP. grcupo 7 ad 8 inse rt.J 1370 1375 Orttical boren. EOC (eg le), p;e nap) 2+0 32 4 py[ . Steve 8 37.51 wd eq Ie 3g gg C.strel rod wrth - ttTP, W. I ah /h Cro%p 6 1.12 0.97 semp 7 G.68 9.7 r.r , 8 37. 5g eat 0.33 0.47 Cent rol rod worth - hTP (250 ETPD). 2 Ak/k crowy 7 1.12 1.M Crs.p 8 17. 51 t.4 0.-0 0.*$

mm ejes ted r.=t wrah - K2P. % f.k/k(

>< 0. 4 e 0. 15 250 LT?D J.47 0..L

• tas eteck rod wrth - EZP.

• ak/k W * .94 2.22 253 f f P3 2.04 2.C4 P.wer deficit - ICP to NFP.1/k/k W -3.63 -1.52 253 LTPD -2.17 -1.e+

av pler coef f - Doc.13*S ak /h *r 141 pswr (grcup 8 In no Ie. -1.47 *t.47 e r a t ical beroe)

DrpPler coeff - E M. 10*5 ikTa *F 10JZ psenet faroup 8 te, eq 1e. -1.5. -1.59 17 r;e borca)

%4es ster c wf f - unN 10** aath 'r FC (group 8 te. no Xe. cr&tical boros) -0.49 -0.41 EN (group 8 to. eq Ie. 17 Spm beres.) -2.58 -2.58 Berm wrtif = EFP p;en/1(Ak/h)

SM (3000 FPe Mrne) 1M 137 IM (17 p:e borce)  % 9e he wrth - HFP.1 Ak/k

%JC (4 ESPD) 2.65 2.e5 EOC (egallabstum) 2.75 2.75 fif active delayed acuti an f raction - hFP BN 0.'0596 0. Cant!

i ECC 0.03522 0.00517 (al Assed on cycle 2 length of 277 EFFD.

Cycle 4 data are for the conditian stated md are based ou a cye2e 3 eterat ang length of 3u2 EFFD; cycle 3 dat.a ma, act be at the see c M ittene as cycle 4.

(c)NZPt het sero power. HFPt hot full power.

El ected rod valw for groupe 5. 6. 7, and 9 insec ted.

5-3 Babcock a. Wilcox E _.

Table 5-2. Shutdown Margin Calculation for Oconee 2 Cvele 4(#

BOC, I ak/k EOL, % ak/k I Available Rod Worth Total rod worth, HIP (C) 8.E6 8.86 Vorth reduction due to burnup -0.26 -0.31 of poison material

~

Maximum stuck rod. HZP -2.22 -2.09 Net worth 6.38 6.46 Less 10% uncertainty -0.64 -0.65 Total available worth 5. 74 5.81 Required Rod Worth Power deficit, HFP to HZP 1.52 1.99 Max allowable inserted rod 0.88 1.48 worth Flux redistribution 0.47, 0.89 Total required worth 2.87 4.36 Shutdown Margin i

Tctal available worth minus 2.87 1.45 1 total required worth

(" Required shutdown margin is 1.00% Ak/k.

I For shutdown margin calculations, this is defined as ap- '

proximately 250 EFPD - the latest time in core life at which the transf ent bank is nearly fully in.

(* HZP: hot zero power, HFP: hot full power.

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Figure 5-1. BOC (4 EFPD) Cycle I. Two-Di::metsional Relative Power Distribution - Full Power, E.trilibriu:n Xenon.

Normal Rod Positions, Groups 7 a:id 8 Inserted S 9 10 11 12 13 it. 15 7

H 1.04 1.16 1.00 0.93 1.37 C 5I. 0.43 0.62 K 1.38 0.86 1.11 0.98 '

.06 0.78 0.71 7 8 '

L 0.59 1.15 1.13  :.c4 1.26 0.70 M 1.29 1.05 1.33 1.16 f

N 0.98 1.16 0.80 0 C. 5:.

P R

0 Inserted Rod Group .%.

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4 TIIERMA1.-HYDRAULIC DESIGN

<- i, l The incoming batch 6 f uel '{s hydraulically and gecnetrically similar to the ,

fuel remaining in the cor:?,from previous cycles. The thermal-hydraulic design l evaluation supporting cyc1594 operation used the methods and models Cescribed in references 2, 4, and 6 pith two exceptio (. The co eptions are the core by-

}

1 passflowandreferencedehignradialx local peaking used in the analysis.

Fuel asse=blies .,,t containing coC.rol rods or neutron sources usually have orifice rod assemblies (ORAsk t installed in the guide tubes Co minimize core bypass flow, including thd two neutron sources, th?re are a total of 108 pos-sible locations for ORAs. ;During cycle 3 operation, 70 OkAs we: c InstallG, leaving 36 vacant fuel assemblies. Themaxi(ecorebypassusedfoOcycle3 analysis was 8.34% based .. an assumed 44 ORAs removed. For efcle 4 operation, all ORAs will be removed, leaving 106 vacant fuel assemblies and a maximum core bypass flow of 10.42.

To offset the effect of the increased core bypass flow in cy O 4 on thermal-hydraulic design, the reference da:.ign radi;3 = local peaking factor (F,,g) Ms been reducco trom1.78to(~,1. This reduction in FAH 18 ullY Supported by the cycle 4 nuclear design; for which tu.c maximum predicted radial r local peaking factor is 1.57. Reactor core safety limits have been re-evaluated based on the reduced F and C.e increased core bypass flw. W cycle 3 and gg 4 maximum design conditions and significant parameters are shown in Table 6-1.

Thepotentialeffectoffud,1rodbowonDNBRwasconsideredbyincorporating l suitable margins into DSB-limited core cafety limits and RPS setpoints. The 1 maximum rod bow penalty was calculated from the following equation:

f=0.065+0.001449.TtI o (1) aC = rod bow magnitude, nils, Co = initial gap (138 mils).

BU = maximum assembly burnup, HWd/stU.

6-1 Babcock s.Wilcox L_

An 11.2% rod bow penalty based on an assumed burnup of 33,0004M'd/=tc is ap-plied to all analyses that define plant operating limits and to design tran-sients. No f uel assembly will achieve a burnup as high as 33,000 .9%d/mtU dur-

~

'ing cycle 4 operation. A thermal margin credit equivalent to 1" CNBR to offset the rod bow penalty has been used as a result of the flow area (pitch) reduc-tion factor included in all thermal-hydraulic analyses. The 1% DN3R is the only credit applied to offset the rod bow penalty.

The flux / flow trip setpoint was determined by analyzing an assumed two-pump coastdown starting f rom an initial power level (indicaced) of 102%. A fidl flow trip setpont of 1.055 is established for cycle 4 based on a minimum DNBR of 1.30 plus a suitable margin to offset the 11.2% rod bow penalty.

The DNBR analysis has been based on a core configuration consisting of 177 Mark B (15 = IT) fuel assemblier. Comparative analyses have been performed to show that -1 theinsertionoftheMarkC{f.7x 17) demonstration assemblies will Increase the HIJBR in the hot assembly. The demonstration a[yemblies have been 1

placed in low-power-producir.j core loc { pions to ensure that they till not be I limiting and to provide minimum impact on tne hot assembly performance. There-fore, the presence of the Mark C demonstrati .n assemb1A?s in cycle 4 will not

• discernibly abkect s.

the thermal-hydraulic character of t'e reactor.

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t Table 6-1. Thermal-Hydraulic Design Condi_ Ions Cycle 32 Cycle 4 Power level, M'*t

. 2568 2368 System pressure, psi. 2200 2200 .

Reactor coolant flow, % of 106.5 106.5 design flow V ssel inlet coolant temp 555.6 553.6 at 100% power, F Vessel outlet coolanc temp 602.4 602.4 at 100% power, F Reference design radial-local 1.78 1.71 power peaking factor Reference design axial flux 1.5 cos 1. 5 cos shape Active fuel length, in. 140.5 140.5 Avg heat flux, 100% pow (), 176 "} 176 8

103 Btu /h-ft2 Cl!F correlation BAW-2 EAW-2 Ilot channel factors Enthalpy rise 1.011 1.011 Heat flux 1.014 1.014 Flow ares 0.98 0.98 -

MDNBR with densification 1.91 1.98 penalty (3) Based on densified length of 140.3 inches.

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h 7. ~ ACCIDENT AND TRANSIENT ANALYSIS "

l 7.1. Ceneral Safety Analysis Each FSARI accident analysis has been examined with respect to changes in .

cycle a parameters to determine the effect of the cycle 4 reload adC to ensure that thernal performance during hypotheticC1 transients is not degraded. The ef fects of fuel densification on the FSAR accident results have been (valuated and are reported in reference 4. Since batch 6 reload fuel assemblies contain C'

fuel rods with a theoretical density higher than those considered in reference

4. the conclusions in that reference are still valid.

No new dose Calculations were performed for the reload report. The dose con- '

siderations in the FSAR were 'oased on maximum peaking and burnup for a&j. core O cycles; tierefore, the dose considerations are 12 dependent of the reload batch.

7.2. Accident Evaluation The key paCkseters that have the greatest et[ic: on the outcome of a trac [ lent can typically be clasaified in three major areas: core thermal parameters, _

thernal-hydrasslic parameters, and kinetics parameters including the reactivity feedback coefficients and control rod worths.

l Fuel theJ. mal analysis parameters for each batch in cycle 4 are ampared in .

Table 4-2. A comparison of the cycle 4 the'r'mal-bydraulic w;ximum design con-ditions to the previous cycle 3 values is presented in Table 6-1. The key ki-netics parameters froC the FSAR and cycle 4 are compared &s Table 7-1.

A generic LOCA analysis h s been performed for the B&'.i 177-FA lowered-loop NSS using the Final Acceptance Criteria ECCS evaluation model reported in reference

7. This analysis is generic in nature since the limiting values of the key pa-

• rameters for all plants in this category were used. Furthermore, the combina-tion of the average fuel temperature as a function of linear heat rate and the s l l

. 7-1 Babcock & VVilcox 4

J lifetise pin pressure data used in the LOCA limits analysis7 is conservati#e c<mpared to those calculated for this reload, n.u a , the analvsis and the LOCA limits reported in reference 5 provided conservative results for the operatica

~

of Oconee 2, cycle 4 fuel. Table 7-2 shous the 'oounding values for allowable LOCA peak linear heat rates for Oconee 2, cycle 4 fuel.

From crasinations of cycle 4 core thermal properties and kinetics properties with respect acceptable previous cycle values, it is concluded that this core relaat I not adversely af fect the ability to operate the Oconee 2 plant safely curing cycle 4 Considering the previously accepted design basis

j. used in the FSAR and subsequent cycles, the transient evaluation of cycle 4 is considered to be bounded by p(hviously accepted analyses. The initLal condi-tions of the transL'[;ts in cycle 4 are bounded by the FSAR, the fuel densifica-tion report" and/or subsequent cycle analyq7s.

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Table 7-1. Camarison of Kev bramters for A cident Analysis FSAR. densif Predicted Par.neter value cycle 4 value BOL Doppler coeff,10-5 Ak/k *F -1.17('} -1.47 E01. c,ngpler coef f,105 Ak/k 'F -1.33 -1.59 BOL moderator coeff, 10-4 Ak/k *F +0. 5 M -0.63 EOL moderator eeef t,10-" Ak/x *F -3.0 -2.5S All rod Ack wrth (H2P), % aklk 10.0 8.86 Boron reac:ivity w rth 0 70F, 75 75 ppm /1% (.1kik)

Max ejceted rOJ wrth (llFP), *. akik 0.65 0.41 Dropped red worth OiFP), 2 Ak/k O.a6 0.20 Initla1 tvrco cece OIFP), ppa 1400 1078

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-1.2 = 10'5 2kik *F was used for steam line failure analysis;

-1.3 = 10~5 ak/h *F was used for cold water analysis.

+0.94 = 10" ak/k *F was used for the moderator dilution accident.

Table 7-2. Bounding Talues for Allevable LOCA Peak L1- e:r Mcat Rates Allowable peak IllR, Core elevation uif k'.'!f t 2 15.5 4 16.d 6 LS.O 8 -17.0

, 10 16.0 j

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k 8. PROPOSED MODIFICATIONS TO TECHNICAL SPECIFICATIONS The Technical S. :cifications have been revised for cycle 4 operation to account for minor changes in pawer peaking and centrol rod worths inherent with non-

'I equilibrium cycles. In addition, changes were the result of the following:

1. The Technical Specification limits based on DNBR and LHR criteria include appropriate allowances for projected fuel rod bow penalties, i.e., poten -

tial reduction in DNBR and increase in power peaks. A statistical combi-nation of the nuclear uncertainty factor, engineering hot channel factor, aad rod bow peaking penalty was used in evaluating LHR criteria, as ap-proved in reference 8.

2. Per reference 9, the power spike penalty due to fuel densification was not used in setting the DNER* .ad ECCS-dependent Technical Specification limits.

3. The allo-able quadrant tilt limit for cycle 4 is 5.0%. A power peaking penalty (1.075) appropriate for this quadrant tilt limit was used to set j imbalance and rod insertion limits. '

l 1

Based on the Technical Specifications derived from the analyses presented in l this report, t'.e Final Acceptance Criteria ECCS lizits will not be exceeded, i l

nor will the thermal design criteria be violated. Figures 8-1 through 8-10 l I

illustrate revisions to previous Technical Specification limits. 'l' m

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I'igure 8-1. Core Protection Safety 1.inits

',OF RATED THERWAL POWER ONER LIMIT -- 120

( -33.60,112) 112.0

@ (10.64.Il2.0)

- 110 I KB FT Lluli k g ACCEPTABLE 4 PUNP tiggy e OPERATION I I " 100 g (31.0.I00.01

-52.0.95.0) l g ,

~~

! 85.31

. f I ACCEPTABLE I g 3 & 4 PUNP

" 00 g }

l 73.31 l OPERATION 68.31 -- 70 I g I

g l

, 58.20 - - 60 k ,

I g

g gACCEPTABLE I I l 2.3 14 PUNP

" 50 l I

' I 41.20 - - 83 l g I I i l 1 I

" 30 g i ,

l l 1 I I

, g

- 2a g g  ;

I - - 10 l l l l l

) - J , , ,

-60 -50 -40 -30 -20 Ib 0 IC 20 30 4h 50 60 Axial Poser leaatac:e 5 CURVE REACTOR CUOLANT FLOW (G/u) -

1 374.880 '

l 2 280.035 l

3 183.690 .

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Figure 8-2. Protective Systes .Maximus 2.41cwable Setpoints THERNAL PCsER LEVEL. 5 UNACCEPTISLE OPERATION 2.463 105.5

( -19.0.105.5) 3.

l ACCEPTASLE-4 PUEF 100 8 l0PERATICst I (20.0.95.0)

'g" l

- -W I

( 4 2.0.80) 78.81 30 ACCEPTAitE I

l3&4PLu? "3 ~

gOPERATICM (20.0.68.31)

I l _ _.

( 42.0.53.31) SI 7e I ACCEPTASLE"" 50 l 2.3. & 4 l l PUNP . .g (20.0.41.20) l g0PERATICM [  ;

I l l -

-3

( -42.0.26.20)  :

I 3

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-10 I I

:  :  : /  :

I

-60 -50 -40 -30 -20 10 0 10 - 20 30 40 50 60 Power Issalance. 5 33 Babcock & Wilcox 1

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Figure S-3. Rod Position Limits for Four-Pu p Operation Frc= 0 to 250 : 10 EFPD - Oconee 2 Cycle .'.

110

( - 3 100 (171.102) (206.102)

OPERATION IN THl$

S0 REGION l$ NOT (373 92)

I ALLORED .206.92) kPORERLEVEL CUTOFF 80 -

(151.4.80) (225.5.80)

E 70 RESYRICTED SHUTOOWN RESTRICTED REGICN

" 0 ulRGIN REGION

~~

L ilfli

.$ 50 t37.50) (125.5.50 (251.4.50) 40 _

PERutSSIBLE

[ OPERATING (300.4 30 (0.28) 20 -

10 -

0 0,. 0 ) , , ,

1 0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 3C0 Roa in::et. 'r 50 I I I I f f f 1 l t t 0 25 50 75 100 25 50 1 15 100 Group 5 Group 7 i i 1 e a 0 25 50 75 100 Grcup 6 l

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F1;;ure 8-4. ilod Position Limits fcr Tour-Pu=p Operation A;ter 250 : 10 ETPD - Ocenee 2 Cycle 4 110 CPEFATION lis THf$ gg33 ggy; I:C PEGION is h0i ALL0eED (259.102) - (3:3 IC2)

(253 7.92) 33 ,

(235.80) P0eER E0 -

Il SM'JT00sN MARGl4 TOFF E 70 -

LIWII RESTRICIED

= REGION 3 60 -

f .

50 - (52.50 (165,50s PERulS$1BLE OPERATING PEGION ll; 40 -

3 30

- (0. 2 7 )

20 -

10 -

0 m - i . . , , , , , , , , ,

0 22 40 60 60 100 120 140 160 183 200 220 240 260 280 300 Ras incer, S 33

' ' ' ' . ,i , , , ,

0 25 50 75 100 0 25 50 15 100 Group 5 Group 7 I I t , ,

0 25 50 75 100 Graup 6 l

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Figure B-5.

Rod Position Li=its for Two- and Three-Pump Operatien From 0 to 250 r 10 EFPD - Oconee 2. Cycle 4 CPERAilch IN THIS 110 , REGION IS NQi ALL0sEO slTH 2 OR (130.102) (151.4.102) (225.5.302) 10b . 3 PU"II (108.102) (246.5.102) 90 - (1 .5.97) y (251.4.97) e RESTRIOTED 80 #

o REGION FOR 3

2 p g Y 213 PUWP g

e"

. SHUf00fW \g S c OPER.

3

=N h $

ss (i25.5. n a5i .4.5x g (300.74)

,Y 50 .f37.50) 8 (300.51) 40 -

x f

30 M 8 PERulSSIBLE (0.28) OPERATlhG REGION 20 -

10 -

0 ' ' ' ' ' ' ' ' ' - ' - - e e 6 20 40 60 80 100 120 140 ISO 180 200 220 24D 250 280 300 320 Roa leces. ', 30 8 e i I e i e i i  ;

O 25 50 75 100 0 25 50 75 100 Group 5 Group 7 i e i t i 0 25 50 75 100 Group 6 l

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> Figare 8-6.

V Rod Position Li=f ts for Two- and D.ree-Pu=p 0;>cration Af ter 250 : 10 EFPD - Cecnec 2, Cycle 4 110 3 P Et& TION th Tels s t33* 102) (180 102) (235 102)

ICO #EGION Is NOT ALLCeED BliH i300.102)

$3 2 CE 3 PUNPS (168,92) 5, g8 80

• S TC -

' g g d

_E_ S

co , g#' # (168.63)

E 50 - (52,50) p 40 -

O g PERulS5s% E 30 g yyy CPERAil4G REGICM 20 -

10 -

O e e e i e i t t t t t t i f 0 23 40 60 80 100 120 140 ISO 120 200 220 243 263 230 3CD Ras Iraen. ',ID E 8 8 I t t t t i f 0 25 50 75 100 0 25 52 75 ICD Group 5 Group 7 9 f f f f 0 25 50 15 100 Group S i

)

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Figure 8-7. Operat ional Power Imbalance Envelope for Operation From 0 to 250 : 10 EFPD - Oconee 2. Cycle 4 Power '. of 2568 MWt RESTRICTED REGION . 110 -

(-13.2.102) ,_

(8.102)

(-I42.92) , ,, g (8.6.92).

(-202,80) -- 80 70 PERMIS$1BLE - 60 OPERATING REGl04 -

- 50

- - 40 1

-- 30

- 20 l

. . 10 0 , , ,

-30 -20 -10 0 1d 2'O [0 Core imbalance. 7.

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Figare 8-8. Operational Power Imbalance Envelope for Operation Meer 250 : 10 EFPD - Ococee 2, Cycle !.

Power ', of 2568 Met

_ _ 110 RESTRI:.TED REGICH

(-22.s.co2) I#O#I

(-2o.i.92) (ir.a.92)

- 80 PERMISSIBLE OPERATING REG 10H -

- 70

- 60

- 50

- 40

- 30

- 20 .

10

, , , 0 , i 30 20 -10 0 10 20 30 Imoalance. %

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Figure 8-9. APSR Position Limits for Operation From 0 to 250 2 10 EFPD - Oconec 2. Cycle 4 g _

RESTRICTED REGION (6.1 102) 100

}(25.5.102) gg (3.2.92) (27.5.92) 80 (0.80) (29.3.80) j 70 -

RESTRICTED m REGION m

fl 60 - -

E 50 - (64.4.50)

(100,40; d' 40 -

PERWISSISt.E OPERATING 30 -

REGION  !

l 20 -

10 -

0 I s t I t t I I i 0 10 20 30 40 50 80 70 80 9C 100 APSR, S WO l I

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Figure 8 APSR Position Limits for Operation After

.30 10 EFPD - Ocor.ee 2. Cycae 1

( _

RESTRICTED REGION 100 - (32.102) 90 (2,3,973 (32.92) 80 (0.80)

(33.7.80)

10 -

w RESTRICTED REGION 5

60 PERNISSIBLE

. OPERATING Q

50 - REGION (64 4.50; l> 40 -

(IOD.40) 30 -

20 -

10 _

0 i , , , , ,

10 20 30 40 50 50 70 80 90 100 APSR $. WO i

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! 8-11 Babcock s.Wilcox

9. ST/.RTI!P PROGRMI - PHYSICS TESTING The planned startup test program associated with core -performance is outlined belev. These tests verify that core performance is within the assunptions of the safety analysis and provide conf.1rmation for continued safe operation of the unit.

9.1. Precritical Test s 9.1.1. Control Rod Trip Test Precritical control rod drop times are recorded for all control rods at hot f ull flow conditions before zero power physics teating begins.' Acceptance criteria state that the rod drop time f rom fully withdrawn to 3/4-inserted

-l shall be less than 1.66 seconds at the conditions above.

It abould be noted that safety analysis calculations are based on a rod drop time of 1.40 seconds f rom fully withdrawn to 2/3-inserted. Since the most ac-curate position indication is obtained from the zone reference switch at the 3/4-anserted position, this position is used insterf of the 2/3-inserted posi-tion for data gathering. The acceptance criterion of 1.40 seconds corrected to a 3/4-inserted position (by rod insertion versus time correlation) is 1.66 seconds.

9.1.2. Reactor Coolant Flow RC flow with four RC pumps running will be measured at hot zero power, steady-state conditioas. Acceptance criteria require that the measured flow be within allowable limits.

i 9.1. 3. RC Flow Coastdown l

The coaz.tdown of RC flow from the tripping of the most limiting RC pump c6m-

.bination from four RC pumps running will be measured at hot zero power condi-tions. The coastdown of RC flow versus time will then be compared to the re-quired value-to determine whether acceptance is met. )

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l 9.2. Zero Power Physics Tests Mr 9.2.1. Critical Baron Concentration Criticality is obtained by deboration at a constant dilution rate. Once cri-ticality is achieved, equilibrium boron is obtained and the critical boron con-centration determined. The critical boron concentration is calculated by cor-recting for any rod withdrawal required in achieving equilibrium boron. The acceptance criterion placed on critical boron concentration is that the actual beron concentratiun must be within 2100 ppm of predicteo.

9._2. 2. Tenperature Reactivity Coefficient The isothersal temperature coef ficient is measured at approximately the all-rod s-ou t configuration and at the hot zero power rod ins rtion limit. The average coolant temperature is varied by first dacressing then increasing t emperature by SF. During the change in temperature reactivity feedback is compensated by discrete change in rod motion; the change in reactivity is then calculated by the summation of reactivity (obtained frna reactivity calcula-t ion on strip chart recorder) associated with the tesperature change.

Acceptance criteria state that the measured value shall not differ frem the predicted value by more than 20.4 = 10~" (ak/k)/*F (the predicted value is ob-tained f ros Physics Tast Manual curves).

The moderator coef ficient of reactivity is calculated in conjunction with the temperature coefficient measurement. Af ter the temperature coaf ficient has 1.cen ocasured, a predicted value of fuel Doppler coefficient of reactivity is i added to obtain moderator coef ficient. This value must not be in excess of the acceptance criteria limit of +0.5 = 10-" (ak/k)/*F.  !

l 9.2.3. Control Rod Group Reactivity Worth )

l Contrul b.nk group reactivity worths (groups 5, 6, and 7) are measured at ho' zero power conditions using the boron / rod swap method. This method ccasists I 1

of establishing a deberation rate in the RC system and compensating fer the ]

reactivity changes of this deborat!on by inserting control' rod groupe 7, 6, and 5 in incresental steps. The reactivity changes that occur during these measure-ments are calculated based on Reactimeter data, and differential rod wort.is are obtained f rom the measured reactivity worth versus the change in rod groep posi- I tion. The dif ferential rod worths of the contro11 Log groups are then semmed to obtain ir.tegral rod group worths.

9-2 Babcock s.Wilcox 1

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1 The acceptance criteria for the control bank group worths are as follows:

1. Individual bank 5, 6, 7 worth:

predicted value - measured value =

measured ,

100l < 15.

2. Sutr of groups 5, 6, and 7:

predicted - measured a 100 < 10.

measured  ; ,

9.2.4 Ejected Control Red Reactivity *. 'o rt h After the CRA groups have been positiened at the rod insertion limit, the ejected rod is borated to 103 withdrawn and the worth obtained by adding the incremental changes in reactivity by boration.

Af ter the ejected rod has been borated to 1,V0% withdrawn and equilibriu:n boron established, the ejected rod is then sva;,peId in versus the controlling rod group and the worth deter =Ined by the change in controlling rod group position.

The boron swap and rod swap values are used to determine ejected rod worth.

Acceptance criteria for the ejected rM worth test are as follows:

1. fP'* *** "* "" ~ " **"#* ""

• measured value = 103 1 20.

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2. Measured value (error adjusted) s 1.01 ak/k.

The predicted ejected rod wrth is given in the Physics Test Manual.

9.3. Power Escalation Tests 9.3.1. Cure Pcwer Distribution Verification at 40, 75, and i

1007 FP With Nominal Control Ecd Group Configuration l Core power distribution tests are perforned at 40, 75, and 100% FP. The test at 40: FP is essentially a check on pwer distribution in the core to bring attention to any abnormalities before escalating to the 75% FP plateau. Rod index is established at a nominal full power configuration, at which the core power distribution calculations are performed. APSR position is established to provide a core power imbalance correaipcmding to the imbalance at which the core power distribution calculations are perforned.

The following acceptance criteria are placed on the 407 FP test:

1. The worst-case maximun linear heat rate must be less than the LOCA limit.

• -3 Babcock a.Wilcox

2. The minimun DNBR must be gre. iter than 1.30.

3.

The value obtained from the ex,trapolation of the minimum DNBR to the next power plateau overpower trip setpoint must be greater than 1.30 or fall outside the RPS power / imbalance trip envelope.

4 The value obtained from the ntrapolation of the worst-case maximum linear '

heat rate to the next power plateau overpower trip setpoint must be less -

than the fuel ::elt limit or fall outside the RPS power / imbalance trip h

envelope. g 5.

The quadrant power tilt shall not exceed the limits specified in the Tech-niral Specifications.

6 The highest measured radial peak and the highest predicted radial peak shall be within the following limits:

Ipredicted - measured measured s W < 8 1 7.

The highest measured total peak and the highest predicted total peak shall be within the,following limits:

I predicted - measured , g caasured

, g '{

Items 1, 2, 5, 6, and 7 above are established to verify core nuclear and ther-mal calculation models, thereby verifying the acceptability of data from these models for input to safety evaluaticos.

Items 3 and 4 estab.11sh the criteria shereby escalation to the next power plateau may be acceeplished without exceeding any safety limits specified by the safety analysis with regard to DN3R and linear heat rate.

The power distribution tests perfor=ed at 75 and 100:

40 FP are identical to the FP test except that core equilibrium x mon is estchlished prior to the 75 and 100: FP tests. Accordingly, the 75 and 100: FP measured peak accept-tance criteria are as follows' 1.

The highest measured radial peak and the highest predicted radial peak shall be within the following limits:

predicted - neasured , , --

measured g3 I

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2.

The highest measured total peak and the highest predicted total peak shall be within the following limits:

gredicted - teasured measured 100l < 7.'S 9.3.2 Incore Vs Excore Detector Imbalance Correlation Verification at %40% FP Imbalances are set up in the core by control rod positioning and are read simultaneously on the incore detectors and excore power range detectors. The imbalances from the excore detectors cust exceed those on the incore detectors by a factor of 1.25.

If t he ratio of excore to incore detector imbalance is less than 1.25, gain amplifiers in the excore detector signal processing equip-ment are adjusted to give the needed gain.

9.3.3. Tenperature 5,eactivity Coefficient at sIOO% FP The average RC temperature is decreased and then increased by about 5F at con-stant reactor power.

The reactivity associated with each temperature change is obtained from the change in the controlling rod group position. Control-ling rod group worth is ceasured by the fast . insert / withdrawal r.ethod.

The temperature reactivity coef ficient is calculated from the measured reactivity

.ind tecperature changes.

Acceptance criteria are that the moderator temperature coefficient shall be negative.

9.3.4 Power Doppler Reactivity Coefficient at 100% FP Reactor power is decreased and then increased by about 5% FP. The reactivity change is obtained from the change in controlling rod group position. Control rod group worth is measured using the fast insert / withdrawal method. Reactivity corrections are made for changes in xenon and RC temperature that occur during the ceasurement.

The power Doppler reactivity coefficient is calculated from the measured reactivity change, adjusted as stated above, and the ceasured power change.

The predicted value of the power Doppler reactivity coefficient is given in the Physics Test Manual. Acceptance criteria state that the mea-sured value shall be more negative than -0.55 x 10-4 (ak/k)/ FP.

9.4. Procedure for t?se When Acceptance Criteria Are Not Met An evaluation is performed before the test program is continued if acceptance-criteria for any. test are not met. This evaluation is performed by sito test 9-5 CC C" m

persornel, with participation by Babcock & Wilcox tecnuacal persc=nel as re-quired. Further specific actions depend on evaluation resielts. 3 ese actions can include retesting with more detailed attention to test prercqz; sites, added tests to search for anomalies, or detailed analysis (by design personnel) of potential safety problems because of parameter deviation. The plant is not escalated in power until evaluation shows that plant safety will cut be compro-nised by such escalatten.

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k REFERENCES I Geonee Nucleay Station, Units 1, 2 and 3, Final Safety Analysis Repcrt, Docket Nos. 50-269, 50-270, 50-287, Duke Power CO.

J Oconec 2 Cycle 3 Reload Report, BAW-1452, Babcock & Wilcox, Lynchburg, Va.,

April 1977.

3 PrOgran to Determine In-Reactor Performance of 5&W Fue(7 - Cladding Creef

?

Collapse, BAW-10084, Rev. 1 Babcock.& Wilcox, Lynchburg, Va., Novenber 1 147f.. l

Geon=e 2 Fuel Densification Report, BAW-1395, Babcock & Wilcox, Lynchburg, Va. , .'une 1973.

5 C. D. Morgan and H. S. Kao TAFY -- Fuel Pin Temper;ture and Cas Pressure Analysis, BAU-10044, Babcock & Wilcox, Lynchburg, Va., May 1972.

Aarndoent to.oconee 2. Cycle 3 Reload Report (BA'e-1&52), Babcock & Wilcox, Lynchburg, V9 ., June 15, 1977.

7 R. C. Jones, J. R. Biller, and B, M. Dunn ECCS Asalysis of B&W's 177-FA Levered Loop NSS, BAW-10103A. Rev. 3, Babcock & Wilcox, Lynchburg, Va. ,

October 1977.

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8 Letter, S. A. Varga (USNRC) to J. H. Taylor (Br4), " Comments on' B&W's Sub- l aittal on Combination of Peaking Factors," May 13, 1977.

9 Let ter, S. A. Varga (USNRC) to J. H. Taylor (B&~4), " Update of BAW-10055 --

Fuel Densification Report," December 5, 1977.

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