ML19331D742

Cycle 6 Reload Rept.

ML19331D742
Person / Time
Site:	Oconee
Issue date:	08/31/1980
From:	BABCOCK & WILCOX CO.
To:
Shared Package
ML16134A664	List: ML16134A663 ML16134A665 ML19331D742
References
BAW-1634, NUDOCS 8009030519
Download: ML19331D742 (38)

Text

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I I BNJ-1636 August 1980 I

I I OCONEE UNIT 3, CYCLE 6

- Reload Report -

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I lI BABCOCK & WILCOX Power Generation Group Nuclear Power Generation Division P. O. Box 1260 Lynchburg, Virginia 24505 l 80090303 6 Babcod & Mcox

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CONTENTS Page

1. INTRODUCTION AND SUMMAR7 . 1-1 I

2. OPERATING HISTORY . . . . . . .. .. . . ... .. .. ..... 2-1

3. GENERAL DESCRIPTION . . . . . .. .... .. . .. . . ..... 3-1

4. FUEL SYSTEM DESIGN . . . . . . .. . . . . ... . . . .... .. 4-1 4.1. Fuel Assembly Mechanical Design .. ... ... . ..... 4-1 4.2. Fuel Rod Design . . . . ... . . ... . . ... . ..... 4-1 4.2.1. Cladding Collapse .. . .. .. . .. . ..... . 4-1 g 4.2.2. Cladding Stress . . . .. .... ... . ..... 4-1 5 4.2.3. Cladding Strain . .. . . .... . ... ..... 4-2 4.3. Thermal Design . . . . . ... . . . .. . .. . . ..... 4-2 4.4. Material Design . . . . ... ... .. . . . . . ..... 4-2 4.5. Operating Experience . . .. .. . . . . . . .. .... .. 4-2

5. NUCLEAR DESIGN . . . . . . . . .. . . . . .. . ... . ..... 5-1 5.1. Physics Characteristics .. . . .. .. . .. . .. .... 5-1 5.2. Analytical Input . . . . . ... . . .. .. .. . ..... 5-2 5.3. Changes in Nuclear Design .. ... . . . .. . . ..... 5-2

6. THERMAL-HYDRAULIC DESIGN . . . .. .. . . . .. .... ..... 6-1 I 7. ACCIDENT AND TRANSIENT ANALYSIS 7.1.

7.2.

General Safety Analysis Accident Evaluation . .

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8. PROPOSED MODIFICATIONS TO TECHNICAL SPECIFICATIONS . . . ..... 8-1 REFERENCES . . . . . . . . . . .. . . . . . . . . . . . ..... A-1 I

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I List of Tables Table Page 4-4 =

4-1. Fuel Design Parameters and Dimensions ............................

4-2. Fuel Thermal Analysis Parameters . . . . . . . . . . . . ..... 4-5 5-1. Oconee 3 Physics Parameters . . . . . . . . . . . . . . ..... S-3 g 5-2. Shutdown Margin Calculation for Oconee 3, Cycle 6 . . . ..... 5-5 3 6-1. Thermal-Hydraulic Design Conditions . . . . . . . . . . ..... 6-2 7-1. Comparison of Key Parameters for Accident Analysis . . . ..... 7-3 7-2. Bounding Values for Allowable LOCA Peak Linear Heat Rates .... 7-3 I

List of Figures Figure 3-1. Core Loading Diagram for Oconee 3, Cycle 6 . . . . . . . ..... 3-2 l 3-2. Enrichment and Burnup Distribution for Oconee 3. Cycle 6 . .... 3-3 5 3-3. Control Rod Locations for Oconee 3, Cycle 6 . . . . . . ..... 3-4 3-4. BPRA Enrichment and Distribution for Oconee 3, Cycle 6 . . .... 3-5 g 5-1. BOC Cycle 6 Two-Dimensional Relative Power Distribution - g Full Power, Equilibrium Xenon, Normal Rod Positions . . ..... 5-6 8-1. Oconee Unit 3 Core Protection Safety Limits . . . . . . ..... 8-2 8-2. Oconee Unit 3 Protective System Maximum Allowable Setpoints ... 8-3 l

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8-3. Oconee 3 Cycle 6 Rod Position Limits - Four-Pump Operation, 0-200 ! 10 EFPD . . . . . . . . . . . . . . . . . . . . ..... 8-4 8-4. Oconee 3 Cycle 6 Rod Position Limits - Four-Pump Operation l After 200 10 EFPD . . . . . . . . . . . . . . . . . . ..... 8-5 3 8-5. Oconce 3 Cycle 6 Rod Position Limits - Two- and Three-Pump Operation, 0-200 10 EFPD . . . . . . . . . . . . . . . ..... 8-6 g 8-6. Oconee 3 Cycle 6 Rod Position Limits - Two- and Three-Pump g Operation After 200 1 10 EFPD . . . . . . . . . . . . . ..... 8-7 8-7. Oconee 3 Cyc]c 6 Operational Power Imbalance Limits . . ..... 8-8 8-8. Oconee 3 Cycle 6 APSR Position Limits, 0-200 1 10 EFPD . ..... 8-9 8-9. Oconee 3 Cycle 6 APSR Position Limits After 200 10 EFPD .... 8-10 I

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1. INTRODUCTION AND

SUMMARY

I This report justifies the operation of the sixth cycle of Oconee Nuclear Sta-tion, Unit 3, at the rated core power of 2568 MWt. Included are the required analyses as outlined in the USNRC document " Guidance for Proposed License Amendments Relating to Refueling," June 1975.

To support cycle 6 operation of Oconee Unit 3, this report employs analytical techniques and design bases established in reports that were previously sub-mitted and accepted by the USNRC and its predecessor (see references).

A brief summary of cycle 5 and 6 reactor parameters related to power capability is included in section 5 of this report. All of the accidents analyzed in the FSAR I have been reviewed for cycle 6 operation. In those cases where cycl- 6 characteristics were conservative compared to those analyzed for previous cy-cles, no new accident analyses were performed.

The Technical Specifications have been reviewed, and the modifications required for cycle 6 operation are justified in this report.

Based on the analyses performed, which take into account the postulated effects of fuel densification and the Final Acceptance Criteria for Emergency Core Cool-ing Systems, it has been concluded that Oconee Unit 3 can be operated safely for cycle 6 at the rated power level of 2568 MWt.

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2. OPERATING HISTORY I The referenced fuel cycle for the nuclear and thermal-hydraulic analyses of Oconee Unit 3, cycle 6, is the currently operating cycle 5. Cycle 4 was ter-t minated after 263 EFPD of operation. Cycle 5 achieved 2Titial criticality on October 28, 1979, and power escalation commenced on October 30, 1979. ne fuel I cycle design length for cycle 6 - 376 EFPD - is based on cycle 5 length of 299 EFPD. No operating anomalies occurred during previous cycle operations that would adversely affect fuel performance in cycle 6.

Cycle 6 will operate in a feed-and-bleed mode for its entire design length, I as did cycle 5.

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3. GENERAL DESCRIPTION I The Oconee Unit 3 reactor core and fuel design basis are described in detail in Chapter 3 of t.he FSAR. I The cycle 6 core consists of 177 fuel assemblies, I each of which is a 15 by 15 array containing 208 fuel rods, 16 control rod guide tubes, and one incore instrument guide tube. The fuel consists of dished-end, cylindrical pellets of uranium dioxide clad in cold-worked Zir -

loy-4. The fuel assemblies in all batches have an average nominal fuel load-ing of 463.6 eg t.ranium. The endensified nominal active fuel lengths, theoret-ical densities, fral and fuel rod dimensions, and other related fuel parameters are given in Tables 4-1 and 4-2.

Figure 3-1 is the core loading diagram for Oconee 3, cycle 6. Thirty-nine of the batch 5 assemblies will be discharged at the end of cycle 5 along with batches 1D, 4B, and 4C, The remaining 17 batch 5 assemblies, designated "5B,"

I batch 6, and the fresh batch 8 FAs - with initial enrichments of 3.02, 2.97, and 3.07 wt % 235 U, respectively - will be loaded into the central ,,ortion of the core. Batch 7, with an initial enrichmsnt of 2.80 wt % 235 U, will occupy primarily the core periphery as in cycle 5. Figure 3-2 is a'n eighth-core map showing the assembly burnup and enrichment distribution at the beginning of cycle 6.

Cycle 6 will operate in a rods-out, feed-and-bleed mode. Core reactivity con-trol is supplied mainly by soluble boron and supplemented by 61 full-length Ag-in-Cd control rods and 60 burnable poison rod assemblies (BPRAs). In addi-tion to the full-length control rods, eight partial-length axial power shaping rods (APSRs) are ;-rov!ded for additional control of axial power distribution.

The cycle 6 locations of the 69 control rods and the group designations are indicated in Figure 3-3. The core locations of the CRAs and APSRAs for cycle 6 are identical to those of the reference cycle; however, the group designa-I tions differ between cycle 6 and the refe ence cycle to minimize power peaking.

The cycle 6 locations and enrichments of the BPRAs are shown in Figure 3-4.

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I Figure 3-1. Core Loading Diagram for Oconee 3 Cycle 6 R10 R9 f

R8 R7 R6 I

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7 7 7 7 7 P12 012 PS 04 P4 B

7 8 7 8 7 8 7 8 7 I 013 D6 K2 K14 D10 03 7 8 6 8 SB 8 5B 8 6 8 7 N14 ES F14 M8 F2 Ell N2 7 8 SB 8 7 8 6 8 7 8 SB 8 7 F4 C7 K6 K10 G13 F12 E

8 6 8 6 8 6 8 6 8 6 8 6 8 Ll', N13 P6 G3 L5 Nll Lil C9 P10 N3 L1 7

7 7 8 7 8 6 6 SB 6 6 8 7 8 7 7 K15 B9 F9 E10 P11 M2 E6 F7 B7 K1 7 8 SB 8 6 6 7 8 7 6 6 8 5B 8 7 H W__ H15 M14 Hll E12 M12 M4 HS 32 Hi 7 5 6 8 5B 8 8 5B 8 A -- Y

? SB 8 7 7 G15 P9 L9 M10 E14 B5 M6 L7 P7 G1 K

7 8 5B 8 6 6 7 8 7 6 6 8 SB 8 7 FIS D13 B6 07 F5 D5 Fil K13 B10 D3 F1 7 7 8 7 8 6 6 53 6 6 8 7 8 7 7 L4 K3 G6 G10 1 09 L12 N "

8 6 8 6 8 6 8 6 6 8 6 8 N

D14 M5 Ll4 E8 L2 Mll D2 7 8 5B 8 7 8 6 8 7 8 SB 8 7 0 Cl3 N6 G2 G14 N10 C3 g 7 8 B12 6 8 C12 SB 8 Bil SB 9 C4 6 8 B4 7

5 P

7 8 7 8 7 8 7 8 7 A10 A9 A8 A7 A6 7 7 7 7 7 Z

l 2 3 4 5 6 7 8 9 10 11 12 13 14 15 XX I

Cycle 5 Location X Batsh No.

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Figure 3-2. Enrichment and Burnup Distribution for Oconee 3, Cycle 6 8 9 10 11 12 13 14 15 3.02 3.07 3.02 3.07 2.97 3.07 2.80 2.80 11 0 18,811 8,410 I

23,433 23,419 0 0 7,675 2.80 2.97 2.97 3.07 3.02 3.07 ".80 I K 8,407 20,308 18,173 0 23,215 0 7,726 I L 2 97 17,748 3.07 0

2.80 10,747 3.07 0

2.80 9,513 2.80 6,438 2.97 3.07 2.97 3.07 M

17,740 0 17,250 0 3.02 3.07 2.80 I

N 21,123 0 5,978 I 0 2.80 6,772 I P I

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X.XX Initial Enrichment, wt % 235U XXXXX BOC Burnup, Mk'd/mtU I

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I Figure 3-3. Control Rod Locations for Oconec 3, Cycle 6 X

A B 3 7 3 c 1 6 6 1 D

7 8 5 8 7 E

1 5 2 2 5 1 y 3 8 4 7 4 8 3 G 6 2 4 4 2 6 11 F- 7 5 7 3 7 5 7 -Y K 6 2 4 4 2 6 L

3 8 4 7 4 8 3 M 1 5 2 2 5 1 N 7 8 5 8 7 0 1 6 6 1 P

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

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I Group No. of rods Function 1 8 Safety 2 8 Safety 3 9 Safety 4 8 Safety 5 8 control 6 8 Control 7 I2 control 8 ,J APSRs Total 69 I

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I I Figure 3-4. BPRA Enrichment and Distribution for Oconee 3, Cycle 6 8 9 10 11 12 13 14 15 I H 1.0 0.5 1.0 K 1.0 0.5 I s 1.0 0.,

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I I X.X BPRA Concentration, wt

• Bq C in Al O 2 3 l

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4. FUEL SYSTEM DESIGN I 4.1. Fuel Assembly Mechanical Design The types of fuel assemblies and pertinent fuel design parameters for Oconee 3, cycle 6, are listed in Table 4-1. All fuel assemblies are identical in concept and are mechanically interchangeable.

I Retainer assemblies will be used on the 60 fuel assemblies containing BPRAs to provide positive retention during reactor operation. Similar retainer assemblies will be used on the two FAs containing the regenerative neutron sources. The justificatica for the design and use af the BPRA retainers is described in reference 3, which is also applicable to the RNS retainers of Oconee 3, cycle 6.

Other results presented in the FSARI fuel assembly mechanical discussions and in previous reload reports are applicable to the reload fuel assemblies.

4.2. Fuel Rod Design The mechanical evaluation of the fuel rod is discussed below.

4.2.1. Cladding Collapse The fuel of batches 6 and 53 is more limiting than other batches due to its longer previous incore exposure time. The batch 6 abd 5B assembly power his-tories were analyzed, and the most limiting assembly was uc,ed to perform the creep collapse analysis using the CROV computer code and procedures described I in topical report BAW-10084, Rev. 2.2 The collapse time for the most limiting assembly was conservatively determined to be more than 30,000 EFPH, which is I greater than the maximua projected residence time of cycle 6 fuel (Table 4-1) .

4.2.2. Cladding Stress The Oconee 3, cycle 6 stress parameters are enveloped by a conservative fuel rod stress analysis. For design evaluation, the primary membrane stress must I be less than 2/3 of the minimum specified unirradiated yield strength. In all cases, the margin is in excess of 30%. The following conservatisms with re-spect to Oconee fuel were used in the analysis:

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1. A lower post-densification fuel rod internal pressure.

2, A lower initial pellet density.

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3. A higher system pressure. 5

4. A higher thermal gradient across the cladding.

4.2.3. Cladding Strain The fuel design criteria specify a limit of 1.0% on cladding circumferential plastic strain. The pellet design is such that the plastic cladding strain is less than 1% at 55,000 mwd /mtU. The following cladding strain conservatisms are applicable with respect to the Oconee 3 fuel:

1. The maximum Specification value for the fuel pellet diar ater was used.

2. The maximum Specification value for tim fuel pellet density was used. p

3. The cladding ID used was the lowest permitted Specification tolerance.

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

4.3. Thermal Design All fuel in the cycle 6 core is thermally similar. The fresh batch 8 fuel inserted for cycle 6 operation introduces no significant differences in fuel thermal performance relative to the other fuel remaining in the core. The design minimum linear heat rate (LHR) capability and the average fuel temptra-ture for each batch in cycle 6 are shown in Table 4-2. LHR capabilities art based on centerline fuel melt and were establiched using the TAFY-34 code with consideration for fuel densification. The maximum fuel rod burnup at EOC 6 is predicted to be 3/,139 mwd /mtU. Fuel rod internal pressure has been evaluated with TAFY-3 for the rod of highest burnup and is predicted to be less than the nominal RC system pressure of 2200 psia.

4.4. Material Design The batch 8 fuel assemblies are not new in concept, nor do they utilire dif-g ferent component materials. Therefore, the chemical compatibility of all pos- 5 sible fuel-cladding-coolant-assembly interactions for the batch 8 fnel assem-l blies is identical to those of the present fuel.

4.5. Operating Experience Babcock & Wilcox operating experience with the Mark B 15 x 15 fuel assembly I has verified the adequacy of its design. As of April 30, 1980, the following 4-2 Babcock a.Wilcox t

I I experience has been accumulated for the eight operating B&W 177-fuel assembly plants using the Mark B fuel assembly:

Maximum assembly Cumulative net I

" D' Current electrical output, Reactor cycle Incore Discharged MWh Oconee 1 19,600 40,000 29,857,021 I Oconee 2 6

5 23,400 33,700 26,232,944 Oconee 3 5 26,300 29,400 25,980,508 TMI-l 4 32,400 32,200 28,840,053 ANO-1 4 25,100 33,222 23,478,392 Rancho Seco 3 37,729 29,378 20,317,332 Crystal River 3 2 23,194 23,194 11,400,975 Davis Besse 1 1 14,600 --

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I Table 4-1. Fuel Design Parameters and Dimensions Batch No.

SB 6 7 8 FA type Mark B4 Mark B4 Mark B4 Mark B4 No. of FAs 17 36 56 68 Fuel rod OD, in. 0.430 0.430 0.430 0.430 Fuel rod ID, in. 0.377 0.377 0.377 0.377 Flex spacers, type Spring Spring Spring Spring Rigid spacers, type Zr-4 Zr-4 Zr-4 Zr-4 m Undensif active fuel 142.23 142.25 142.23 141.8 length, in.

Fuel pellet OD (mean 0.3695 0.3695 0.3695 0.3686 spec), in.

g Fuel pellet initial 94.0 94.0 94.0 95.0 3 density (mean spec),

%TD Initisi fuel enrich- 3.02 2.97 2.80 3.07 m ment, wt % 235U Est residence 26,338 22,522 26,304 29,232 time, EFPH Cladding collapse >30,000 >30,000 >30,000 >30,000 time, EFPH I

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I I Table 4-2. Fuel Thermal Analysis Parameters Batch No.

SB _ 6 7 8 No. of assemblies 17 36 56 68 Initial density, % TD 94.0 94.0 94.0 95.0 Pellet diameter, in. 0.3695 0.3695 0.3695 0.3686 Stack height, in. 142.2 142.2 142.2 141.8 Densified Fuel Parameters Pellet diameter, in. 0.3646 0.3646 0.3646 0.3649 Fuel stack height, in. 140.5 140.5 140.5 140.74 Nominal linear heat 5.80 5.80 5.80 -5.79 rate @ 2568 MWt, kW/ft I Average fuel temp @

nominal LHR, F 1320 1320 1320 1310

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Linear heat rate capa- 20.15 20.15 20.15 20.15 bility (centerline fuel melt), kW/ft Core avg linear heat rate = 5.80 kW/ft.

I (" Densification to 96.5% TD assumed.

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5. NUCLEAR DESIGN I 5.1. Physics Characteristics Table 5-1 compares the core physics parameters of design cycles 5 and 6; the values for both cycles were generated using PDQ07.5-7 Since the core has not I

yet reached an equilibrium cycle, differences in core physics parameters are to be expected between the cycles. The longer cycle 6 will produce a higher cycle burnup than that for the design cycle 5. Figure 5-1 illustrates a rep-resentative relative power distribution for the beginning of the sixth cycle at full power with equilibrium xenon and normal rod positions.

The initial BPRA loading, longer design life, and different shuffle pattern for cycle 6 make it difficult to compare the physics parameters with those of cycle 5. The critical boron concentrations for cycle 6 are higher because the additional reactivity necessary for the longer cycle is not completely offset by burnable poison. The control rod worths differ between cycles due to changes in radial flux and burnup distributions, which also accounts for the smaller BOC stuck and ejected rod worths in cycle 6 compared to cycle 5 values.

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 I position limits presented in section 8. All safety criteria associated with The adequacy of the shutdown margin with cycle 6 these rod worths are met.

st tek rod 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.

3. Flux redistribution penalty.

Flux redistribution was accounted for since the shutdown analysis was calcu-lated using a two-dimensional model. The reference fuel cycle shutdown margin is presented in the Oconee 3, cycle 5 reload report.'

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I The cycle 6 power deficits, differential boron worths, and effective delayed I! ,

neutron fractions differ from those of cycle 5 because of the presence of burn-  !

able poison and the longer cycle length.

5.2. Analytical Input The cycle 6 incore measurement calculation constants to be used to compute core power distributions were obtained in the same manner for cycle 6 as for the reference cycle. 5 5.3. Changes in Nuclear Design There is only one significant core design change between the reference and reload cycles. This change is the increase in cycle lifetime to 376 EFPD and g the subsequent incorporation of BPRAs to aid ic reactivity control. The cal- EU culational methods and design information used to obtain the important nuclear design parameters for this cycle were the same as those used for the reference cycle.

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I I Table 5-1. Oconee 3 Physics Parameters (*}

Cycle 6(b) Cycle 5 Cycle length, EFPD 376 292 Cycle burnup, mwd /mtU 11,766 9137 Average core burnup, EOC, mwd /mtU 20,231 18,711 Initial core loading, mtU 82.1 82.1 Critical boron - BOC (no xenon), ppm HZP, group 8 inserted 1471 1351 HFP, group 8 inserted 1282 1161 I Critical boron - E0C (equil xenon), ppm 385 339 I

HZP, group 8 inserted HFP, group 8 inserted 78 61 Control rod worths - HFP, BOC, % Ak/k Group 6 0.98 1.00 Group 7 1.36 1.70 Group 8 0.50 0.49 Control rod worths - HFP, EOC( }, % Ak/k Group 7 1.48 1.64 I Group 8 0.54 0.51 Max ejected rod worth - HZP, % Ak/k BOC, (N12) groups 5-8 inserted 0.38 0.46 EOC, (N12) groups 5-8 inserted 0.51 0.50 Max stuck rod worth - HZP, % Ak/k BOC (M13) 1.39 1.81 EOC (M13) 1.52 1.75 I Power deficit, HZP to HFP, % Ak/k BOC 1.39 1.34 EOC 2.22 2.11 Doppler coeff - BOC,10-5(Ak/k *F) 100% power (no xenon) -1.49 -1.51 Doppler coeff - EOC, 10-5(Ak/k *F) 100% power (equil xenon) -1.62 -1.57 I

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I Table 5-1. (Cont'd)

Moderator coeff - HFP, 10-4 (Ak/k *F)

BOC (0 xenon, 1282 ppm, group 8 ins.) -0.65 -0.66 g EOC (equil xenon, 17 ppm, group 8 ins.) -2.82 -2.69 5 Boron worth - HFP, ppm /% ok/k 30C (1300 ppm) 116 109 EOC (17 ppm) 102 95 Xenon worth - HFP, % Ak/k Bec (4 days) 2.61 2.65 EOC (equilibrium) 2.74 2.75 Eff delayed neutron fraction - HFP BOC 0.00628 0.00585 EOC 0.00526 0.00519

(*) Cycle 6 data are for the conditions stated in this report. The cycle 5 core conditions are identified in reference 5.

(b) Based on a 299-EFPD cycle 5.

(# Based on 270-EFPD cycle 4.

(d)292 EFPD in cycle 5, 376 EFPD in cycle 6.

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I I Table 5-2. Shutdown Margin Calculation for Oconee 3, Cycle 6 BOC, EOC,

% Ak/k  % Ak/k Available Rod Worth Total rod worth, HZP 8.49 S.92 I Worth red'n due to poir,n burnup Maximum stuck rod, HZ'

-0.29

-1.39

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-1.52 Net worth 6.81 7.10 Less 10% uncertainty -0.68 -0.71 Total available worth 6.13 6.39 Required Rod Worth Power deficit, HFP to HZP 1.39 2.22 I Max inserted rod worth Flux redistribution 0.42 0.57 0.54 1.18 Total required worth 2.38 3.94 I Shutdown Margin Total avail worth - total req'd 3.75 2.45 I worth Note: Required shutdown margin is 1.00% Ak/k.

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I' Figure 5-1. BOC (4 EFPD) Cycle 6 Two-Dimensional Relative Power Distribution - Full Power, Equilibrium Xenon, Normal Rod Positions 8 9 10 11 12 13 14 15 H 1.084 1.258 0.976 1.266 1.122 1.232 1.103 0.650 I

K 1.196 0.991 1.079 1.234 1.032 1.123 0.624 I

L 1.084 1.040 1.208 1.242 0.966 0.486 M 1.105 1.185 1.032 0.921 N 0.970 0.991 0.537 I

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Inserted Rod Group No.

I 0.00NO Relative Power Density I

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I I 6. THERMAL-HYDRAULIC DESIGN I The incoming batch 8 fuel is hydraulically c.nd geometrically similar to the fuel remaining in the core from previous cycles. The thermal-hydraulic de-sign evaluation supporting cycle 6 operation employed the methods and models described in references 1, 5, and 9.

The maximum core bypast, flow for cycle 5 was 10.4% of the total system flow.

Retainers will be placed I

For cycle 6 operation, 60 BPRAs will be inserted.

on these assemblies as described in reference 3. Two assemblies contain re-generative neutron sources and retainers. The number of open assemblies is 46, and the maximum core bypass flow is reduced to 8.1%. The cycle 5 and 6 maximum design conditions are summarized in Table 6-1.

A rod bow DNBR penalty has been caluelated for cycle 6 operation according to procedures approved by reference 10. The burnup used to calculate the penalty is the highest batch 7 burnup, 23,411 INd/mtU. The net rod bow penalty is 1.1% af ter taking credit for the flow area reduction hot channel factor used in all DNBR calculations. However, all plant operating limits based on DNBR criteria include a minimum of 10% DNBR margin from the B&W-2 correlation de-sign limit of 1.30.

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I Table 6-1. Thermal-Hydraulic Design Conditions Cycle 5 Cycle 6 Design power level, MWt 2568 2568 System pressure, psia 2200 2200 Reactor coolant flow, % design flow 106.5 106.5 Core bypass flow, % to?al flow 10.4 8.1 Vessel inlet / outlet coolant temp at 555.6/602.4 555.6/602.4 100% power, F Ref design radial-local power 1.71 1.71 peaking fact or Ref design axial flux shape 1.5 cosine 1.5 cosine llot channel factors: Enthalpy rise 1.011 1.011 a b8 b Active fuel length, in. (a) (a)

Avg hest flux at 100% power, 10 3 176( ) 176(

Btu /h-ft2 (a)

CliF correlation BAW-2 BAW-2 Min DNBR with densification penalty 1.98 2.05

(" See Table 4-2.

Heat flux based on densified length of 140.3 in., which is a con-servative minimuai value.

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7. ACCIDENT AND TRANSIENT ANALYSIS I 7.1. General Safety Analysis Each FSARI accident analysis has been examined with respect to changes in cycle 5 parameters to determine the effect of the cycle 6 reload and to ensure that thermal performance during hypothetical transients is not degraded. The effects I of fuel densification on the FSAR accident results have been evaluated and are reported in reference 9. Since batch 8 reload fuel assemblies contain fuel I rods with a theoretical density higher than those considered in reference 9, 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 based on maximum peaking and burnup for all core cycles; therefore, the dose considerations are independent of the reload batch.

7.2. Accident Evaluation I The key parameters that have the greatest effect on determining the outcome of a transient can typically be classified in three major areas: core thermal I parameters, thermal-hydraul'c parameters, and kinetics parameters, including the reactivity feedback coefficients and control rod worths.

Fuel thermal analysis parameters for each batch in cycle 6 are given in Table 4-2. Table 6-1 compares the cycle 5 and 6 thermal-hydraulic maximum design conditions. Table 7-1 compares the key kinetics parameters from the FSAR and cycle 6. Generic LOCA analyses have been performed for the B&W 177-FA lowered-loop NSS using the Final Acceptance Criteria ECCS evaluation model reported in reference 11. These analyses are generic in nature since the limiting values of the key parameters for all plants in this category were used. Furthermore, the combination of the average fuel temperature as a function of linear heat rate and the lifetime pin pressure data used in the LOCA limits analyses 11,12 I are conservative compared to those calculated for this reload. Thus, the anal-yses and the LOCA limits reported in references 11 and 12 provide conservative I

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results for the operation of Oconee 3, cycle 6 fuel. A tabulation showing the I'i bounding values for allowable LOCA peak LHRs for Oconee 3, cycle 6 fuel is pro- I vided in Table 7-2.

From the examination of cycle 6 core thermal properties and kinetics properties ,

with respect to acceptable previous cycle values, it is concluded that this core reload will not adversely affect the safe operation of the Oconee 3 plant during cycle 6. Considering the previously accepted design basis used in the ,

FSAR and subsequent cycles, the transient evaluation of cycle 6 is considered to be bounded by previously accepted analyses. The ir it.ial conditions of the transients in cycle 6 are bounded by the FSAR and/or the fuel densification report.9 E

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I Table 1 -1. C,.mparison of Key Parameters for \ccident Analysis I Parameter FSARI value Predicted (#}

cycle 6 value I BOC Dqpler coeff, 10-5, Ak/k/*F EOC Doppler coeff, 10-5 Ak/k/*F

-1.17

-1.33

)

-1.49

-1.62 BOC moderator coeff, 10-4, Ak/k'F +0.5(b) -0.65 EOC moderator coeff, 10-4 Ak/k/*F -3.0 -2.82 All rod bank worth, HZP, % Ak/k 10.0 8.49 Boron reactivity worth, 70*F.

ppm /l% Ak/k 75 82 Max. ejected rod worth, HFP, % Ak/k I Dropped rod worth, HFP, % Ak/k 0.65 0.46 0.31 0.20 Initial boron conc, HFP, ppm 1400 1282 I

(" -1.2 x 10-5 Ak/k/F was used for steam-line failure analysis.

-1.3 x 10-5 Ak/k/F was used for cold water accident (pump start-up) .

(b)+0.94 / 10-4 Ak/k/F was used for the moderator dilution accident.

c)Using reference 6.

I I Table 7-2. Bourding Valoes for Allowable LOCA Peak Linear Heat Rates I Core elevation, ft Allowable peak linear heat rate, kW/ft 2 15.5 4 16.6 6 18.0 8 17.0 10 16.0 I

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8. PROF 0 SED MODIFICATIONS TO TECHNICAL SPECIFICATIONS I The Technical Specifications have been revised for cycle 6 operation in accord-ance with the methods of references 13-15 to account for minor changes in power peaking and control rod worths inherent with a transition to 18-month, lumped burnable poison cycles.

Based on the Technical Specifications derived from the analyses presented in this report, The Final Acceptance Criteria ECCS limite will not be exceeded, nor will the thermal design criteria be violated. Figures 8-1 through 8-9 are I

revisions to previous Technical Specification limits.

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I Figure 8-1. Oconee Unit 3 Core Protection Safety Limits I

THERMAL POWER LEVEL, %

120

( 27.44,112) (31.36,112) l ACCEPTABLE

- !!0 l4 PUMP l 5 Mi = 0.79 f, OPERATION l Mg = -2.54 3

- 100 1 I

( 55,90) l -- 90  !

B7.17 , I (40,90) l l ACCEPTABLE l l

g g 3 & 4 PUMP 80 g g g l OPERATION , 3 l -

- 70 I l  !

(-55.65.17) l l l 59.42 I l (40,65.17) 60 l ACCEPTABLE g l l 2,3 & 4 PUMP l l OPERATION- 50 l l l 1

(-55,37.42) '

1 l

~ ~

l (40,37.42) g I

I l

- 30 1 l l -

g i i l-I -- 20 m  ! -

UNAC';EPTABLEl l "

l l UNACCEPTABLE OPERATION i

j -

- 10  ! l " '"

t I g l gi g LI , i il , i i , il I i i B 50 -40 -30 -20 -10 0 10 20 30 40 50 60 Reactor Power imoalance, %

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I Figure 8-2. Oconee Unit 3 Protective System Maximum Allowable Setpoints I THERMAL POWER LEVEL, %

I (-14,108) , 108 - - 110 -(17,108)

ACCEPTABLE l

- - 100 g2 = -1.76 Mj = 0.88 loP Tion l I l - - 20 l

(-40,85) ,80 80.67 l l

80 l /luan,a- - ,0 i i l0PERATION l l -

- 60 l I

,l (-40,57.67) g (30,57.67) l 5L92 I - 50 3

j accEPTastE-2.a . ,

l l l , PU* l

- 40 l OPERATION l

(-40.29.92) I l

_ 30 (30,29.92)

I I I l

- - 20 ll I

UNACCEPTABLE l l UNACCEPTABL OPERATION l' '

- - 10 Cl 8! OPERATION in ni n, u i l= ,=l, i = 1, =l

,g i , i i i 53 -40 -30 -20 -10 0 10 20 30 40 50 l Power imbalance, %

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4 Figure 8-3. Oconee 3 Cycle 6 Rod Position Limits - Four-Pump Operation, 0-200 2 10 EFPD (15b D2) (267.102) 90 -

T (261.97)

CUT 0FF =

100% FP) 80 -

SHUTDOWN MARGIN + (251,80 RESTRICTED 70 -

REGION E

OPERATION IN THIS + Kt/FT LIMIT 60 -

REGION NOT ALL0nED i

50 - (83,50) cn e I- c

" PERMISSIBLE OPERATING REGION g 40 -

a.

5 i

30 -

20 I

(0,14), (13,15) 10 -

0 ' ' ' ' ' ' ' ' ' ' ' ' ' '

as 0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 "r

c n ,0 2,5 5,0 7,5 100 Rod inden, $ Witparawn O

Bank 5 pr 0 25 50 75 100 4 y i i t i i

=: 50 g 0, ,25 , ,5 7

10,0 Bank 7 m W W W W W M W W W W M M M M M M W W

1

, Na, 1 me 1 1

I Figure 8-4 Oconee 3 Cycle 6 Rod Position Limits - Four-Pump Operation After 200 10 EFPD I

I 100 _

(208.102) (267,102)

POWER LEVEL 90 ~

CUT 0FF = 1005FP

( I 1 I 80 -

SHUT 00NN MARGIN + (251.80) i LIMli RESTRICTED 10N

- 70 -

OPERATION IN THIS E REGION NOT ALLORED a

I 60 -

~

o

1. 50 -

(I35.50) PERMIS$1BLE OFERATING REGION g 40 -

30 (84.27) 20 -

I (0,10) (58.15) 104 0 I 0 0

20 40 60 80

+ -

100 120 140 150 i

180 i

200 220 240 e

260 r .

280 300 0 25 50 75 100

, , , , Roa inces, % witnaraan 1 Bank 6 0 2,5 50 75 100

'8"" I O 25 50 75 100 i i , , i Bank 7 I

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i Oconee 3 Cycle 6 Rod Position Limits - Two- and 4

I Figure 8-5.

Three-Pump Operation, 0-200 1 10 EFFD i

i A

RESTRICTED 3 RCP 100 -

B - RESTRICTED 2,3-RCP 90 -

80 -

(158,77) (242'77)

] (300.77) 70 -

3 DCP SHUTDOWN MARbn. .lMIT

8

" PERMSSIBLE OPERATING 60 REGION 3-RCP i

3 N (158,52)

- 50 -

(300,52) f a

2-RCP (164,52)

SM LIMI J 40 -

A l j (83.38) gg FT LIMIT a.

30 -

PERMISSIBLE OPERATING REGION (94.30) 2,3 RCP (83.26) 113.12) A j (0,11) j 10

, (0,8) l 0 ' ' ' ' ' ' ' ' ' ' i i

@ 0 20 40 60 80 100 120 140 160 180 200 220 240 230 280 300 25 I

i

[

o f ,

50 75 100 Rod index *, Witndrawn

! o Bank 5

, O 25 50 75 100 t

Bank 6 0

x  ! 2,5 50 7,5 10p Bank 7 i

m M W m m m m M M M M M M M M M M M Figure 8-6. Oconee 3 Cycle 6 Rod Position Limits - Two- and Three-Pump Operation After 200 1 10 EFPD 100 -

A - RESTRICT 3-RCP B - RESTRICT 2,3 RCP 90 -

80 -

(242,77) (300,77)

(208,77) ,

i 70 y*9 E OPERATION IN THIS REGl0N g h W ALLOWE0 g[

g 60 -

- (208,52)

• 50 -

(1 8,50) @ E) oo b d 1 j 40 -

3 RCP (135.30) 2-RCP SHUT 00RN MARGIN + [g SHUT 00RN MARGIN 30 -

Ll4lT \g@g tigiy

'Y PERMISSIBLE OPERATING 20 REGION 2.3 RCP

, (ES.12) 10 -(0.8) _

a # 58.8.5) *

(0.6) '

0 ' ' ' ' ' ' ' ' ' ' ' ' ' '

m 0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 g

0 l 25, 50 75 I

100 f

Rod index, % Witndrawn E Bank 5

• 5,0

p. O, 2,5 75, 10,0

E Bant 6 5

n 0

i 25 i

50 i 75, 10,0 Bank 7

l i

Figure B-7. Ocor.ee 3 Cycle 6 Operational Power I= balance Limits RESTRICTED g REGION g

< 23.102)

(-30.102) eg

- 100

(-23.92) (25.52)

(-38.80) -- 50 (2L E31 I'

I ~

( l m

N -- 60

~

o i PERNISSIELE CPERATING .- l i REGION E 1 2 -- 40 I

-- 20 I, i

I,

. . . 0 i *

-40 -30 -20 -10 0 + 10 +20 +30 4,,,,,=..,m=,,,.=.s I

(

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- , , , , , ..g.

-w.-w--w-,w -- 7 -,e,pw m.--ye-- ee--- - ' - ww- 'e+--wN*-= wen---'-- ' - " " - - - ' - " - " "* "'

I I Figure 8-8. Oconee 3 Cycle 6 APSR Position Lim'ts, 0-200 t 10 EFPD (6. 5,102) jgg _

r (36.102) RESTRICTED REGION (6.5,92) (34.5.92)

I 80 < (0.80)

~

(55,80)

I h 60 -

m M

I (100.50)

O 40 -

PERMISSIBLE OPERATING REGION I

I 20 -

I I O 0 20

' I 40 I

60 I

80 100 Bank 8 Position, % Ittnaraan

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I Figure 8-9. Oconee 3 Cycle 6 APSR Position Limits After 200 1 20 EFPD (13,102) 100

- 7 (36,102)

(11.5,92) (34.5,92) 80 <

/(0,80) (55.80) s I

E 60 -

<(100,50)

! 40 -

PERMISSIBLE OPERATING REGION E

l I

20 -

0 3

0 20 40 60 80 100 Bank 8 Position, % Witnarawn l

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I REFERENCES 1 Oconee Nuclear Station, Units 1, 2, and 3 Final Safety Analysis Report, Docket Nos. 50-269, 50-270, and 50-287.

2 A. F. J. Eckert, H. W. Wilson, and K. E. Yoon, Program to Determine In-reactor Performance of B&W Fuels - Cladding Creep Collapse, BAW-10084A, Rev 2, Babcock & Wilcox, January 1979.

3 BPRA Retainer Design Report, BAW-1496, Babcock & Wilcox, May 1978.

4 C. D. Morgan and H. S. Kao, TAFY - Fuel Pin Temperature and Gas Pressure Analysis, Babcock & Wilcox, BAW-10044, May 1972.

5 Oconee Unit 3, Cycle 5 - Reload Report, BAW-1522, Babcock & Wilecx, March 1979.

6 B&W Version of PDQ07 Code, BAW-10117A, Babcock & Wilcox, January 1977.

7 Core Calculational Techniques and Procedures, BAW-10ll8, Babcock & Wilcox, October 1977.

8 Assembly Calculations and Fitted Nuclear Data, BAW-10ll6A, Babcock & Wilcox, May 1977.

9 Oconee 3 Fuel Densification Report, BAW-1399, Babcock & Wilcox, November 1973.

10 L. S. Rubenstein (NRC) to J. H. Taylor (B&W) Letter, " Evaluation of Interim Procedure for Calculating DNBR Reductions Due to Rod Bow," October 18, 1979.

11 ECCS Analysis of B&W's 177-FA Lowered-Loop NSS, BAW-10103, Rev. 2, Babcock

& Wilcox, September 1975.

12 J. H. Taylor (B&W) to S. A. Varga (NLC), Letter, July 18, 1978.

13 Power Peaking Nuclear Reliability Factors, BAW-10119, Babcock & Wilcox, January 1977.

A-1 Babcock 8.WilCOX

I 14 Normal Operating Controls, BAW-10122, Labcock & Wilcox, August 1978.

15 Verification of the Three-Dimensional FLAME Code, BAW-10125A, Babcock & g Wilcox, August 1976. 5 I

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