ML20032D677

Cycle 6 - Reload Rept

ML20032D677
Person / Time
Site:	Oconee
Issue date:	08/31/1981
From:	BABCOCK & WILCOX CO.
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References
BAW-1691, NUDOCS 8111170381
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Text

e-I BAW-1691 August 1981 I

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OCONEE UNIT 2, CYCLE 6

- Reload Report -

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BABCOCK & WILCOX I

Nuclear Power Group Nuclear Power Generation Division P. O. Box 1260 Lynchburg, Virginia 2^505 abcock M cox I

8111170381 811113 PDR ADOCK 05000270 P

PDR

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I CONTENTS Page 1.

INTRODUCTION AND

SUMMARY

1-1 2.

OPERATING HISTORY 2-1 3.

GENERAL DESCRIPTION 3-1 4.

FUEL SYSTEM DESIGN.

4-1 5

4.1.

Fuel Assembly Mechanical Design 4-1 4-1 m

4.2.

Fuel Rod Design 4-1 4.2.1.

Cladding Collapse 4-2 3

4.2.2.

Cladding Stress 4-2 g

4.2.3.

Cladding Strain 4-2 4.3.

Thermal Design.

4-2 4.4.

Material Design 4-3 4.5.

Operating Experience.

5.

NUCLEAR DESIGN.

5-1 5.1.

Physics Characteristics 5-1 5-2 5.2.

Analytical Input......................

5.3.

Changes in Nuclear Design 5-2 6.

THERMAL-HYDRAULIC DESIGN.

6-1 7.

ACCIDENT AND TRANSIEST ANALYSIS 7-1 7-1 7.1.

General Safety Analysis 7.2.

Accident Evaluation 7-1 8.

PROPOSED MODIFICATIONS TO TECHNICAL SPECIFICATIONS.

8-1 REFERENCES.

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I List of Tables Table Page I

4-1.

Fuel Design Parameters and Dimensions 4-4 4-5 4-2.

Fuel Thermal Analysis Parameters - Oconee 2 Cycle 6 5-1.

Oconee 2, Cycle 5 and 6 Physics Parameters.

5-3 I

5-2.

Shutdown Margin Calculation - Oconec 2 Cycle 6 5-4 6-2 6-1.

Thermal-Hydraulic Design Conditions 7-1.

Comparison of Key Parameters for Accident Analysis.

7-3 I

7-2.

LOCA Limits, Oconee 2 Cycle 6, After 50 EFPD 7-4 7-3.

LOCA Limits, Cconee 2 Cycle 6, 0 to 50 EFPD 7-4 7-4.

Comparison of FSAR and Cycle 6 Accident Doses 7-5 I

l List of Figures Figure 3-1.

Core Loading Diagram - Oconee 2 Cycle 6.

3-2 3-2.

Oconee 2 BOC 6 Enrichment and Burnup Distribution I

After a 390-EFFD Cycle 5 3-3 3-3.

Oconee 2 Cycle 6 Control Rod Locations and Designations.

3-4 3-4.

Oconee 2 Cycle 6 BPRA Enrichment and Distribution.

3-5 5-1.

BOC 6 Two-Dimensional Relative Power Distribution -

Full Power, Equilibrium Xenon, Normal Rod Positions, 5-5 Group 8 Inserted 8-1.

Core Protection Safety Limits for Oconee Unit 2, Cycle 6 8-2 I

Core Protection Safety Limits for Ocor.ee Unit 2, Cycle 6 8-3 8-2.

8-3.

Maximum Allowable Setpoints for Oconee Unit 2, Cycle 6 8-4 8-4.

Protective System Maximum Allowable Setpoints, I

8-5 Oconee 2, Cycle 6.

8-5.

Oconee 2 Cycle 6 Rod Position Limits - Four-Pump Operation, O to 50 ! 10 EFPD 8-6 I

8-6.

Oconee 2 Cycle 6 Rod Position Limits - Four-Pump Operation, 50 10 to 225 10 EFPD.

8-7 8-7.

Oconce 2 Cycle 6 Rod Position Limits - Four-Pump Operation After 225 10 EFPD.

8-8 I

8-8.

Oconee 2 Cycle 6 Rod Position Limits - Three-Pump Operation, O to 50 10 EFPD 8-9 8-9.

Oconee 2 Cycle 6 Rod Position Limits - Three-Pump I

Operation From 50 10 to 225 10 EFPD.

8-10 8-10.

Oconee 2 Cycle 6 Rod Position Limits - Three-Pump 8-11 Operation Af ter 225 ! 10 EFFD.

I 8-11.

Oconee 2 Cycle 6 Rod Position Limits - Two-Pump 8-12 Operation, O to 50 ! 10 EFPD 8-12.

Oconee 2 Cycle 6 Rod Position Limits - Two-Pump Operation From 50 ! 10 to 225 10 EFPD.

8-13 I

8-13.

Oconee 2 Cycle 6 Rod Position Limits - Two-Pump Operation After 225 10 EFPD.

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Babcock & Wilcox

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I Figures (Cont'd)

Flgure Page H-14.

Oconee 2 Cycle 6 Operational Power Imbalance Limits, O to 50 1 10 EFPD 8-15 H-15.

Oconee 2 Cycle 6 Operat ional Power Imbalance Limi t s,

50 t 10 to 225 ! 10 EFPD 8-16 H-16 Oconee 2 Cycle 6 Operational Power Imbalance Limits After 225 ! 10 EFPD 8-17 1

ll H-17.

Oconee 2 Cycle 6 APSR Position Limits, O to 50 2 10 EFPD 8-18 8-18.

Oconee 2 Cycle 6 APSR Position Limits, 50 ! 10 m,

to 225 t 10 EFPD 8-19 f

8-19.

Oconee 2 Cycle 6 APSR Position Limits After 225 2 10 EFPD 8-20 "l

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I 1.

INTRODUCTION AND

SUMMARY

This report justifies the operation of the sixth cycle of Oconee Nuclear Sta-tion, Unit 2, 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 2, this report employs analytical techniques and design bases established in reports that were previously sub-mitted and ac-epted 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 I

FSAR have been te'riewed for cycle 6 operation.

In those cases where cycle 6 characteristics,ere conservative compared to those analyzed for previous cy-cles, no new accident analyses were performed.

One Mark BZ demonstration fuel assembly containing Zircaloy intermediate grids is in the core as part of batch 7; reference 2 describes the demonstration as-sembly.

In addition, four burnable poison rod assemblies (BPRAs) will remain I

in the core for a second cycle to gather burnup data on burnable poison.

Neither the Mark BZ demonstration assembly nor the reinserted BPRAs will ad-versely affect cycle 6 operation.

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 I

of fuel densification and the Final Acceptance Criteria for Emergency Core Cool-ing Systems, it has been concluded that Oconee Unit 2 can be operated safely for cycle 6 at the rated power level of 2568 MWt.

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2.

OPERATING HISTORY The reference fuel cycle for the nuclear and thermal-hydraulic analyses per-formed for cycle 6 operation is the currently operating cycle 5.

Cycle 5 achieved initial crit icality on June 21, 1980 and power escalation began on June 24, 1980.

The 100% power level of 2568 MWt was reached on July 9, 1980.

No operating anomalies occurred during cycle 5 operation that would adversely affect fuel performance during cycle 6.

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3.

GENERAL DESCRIPTION The Oconee 2 reactor core is described in detail in Chapter 3 of the FSAR.I The cycle 6 core consists of'177 fuel assemblies, all of which have 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 Zircaloy-4. All fuel assemblies in cycle 6 have a constant nominal fuel loading of 463.6 kg of uranium. The un-densified nominal active fuel lengths, theoretical densities, fuel and fuel rod dimensions, and other related fuel parameters are included in Tables 4-1 and 4-2 of this report.

Figure 3-1 is the core loading diagram for Oconee 2, cycle 6.

Batches 6B and 23s, respectively, will be 7, with initial enrichments of 2.91 and 3.07 wt %

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Batch 8, with an initial enrichment of 3.17 wt %

shuffled to new locations.

U, will be loaded in a checkerboard pattern.

Figure 3-2 is an eighth-core 235 map showing the assembly burnup and enrichment distribution at the beginning of cycle 6.

Reactivity control is supplied by 61 full-length Ag-In-Cd control rods, 64 fresh BPRAs, and soluble boron shim.

In addition to the full-length control rods, eight partial-length axial power shaping rods (APSRs) are provided ofr 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 total pattern (69 control rods) for cycle 6 are identi-cal to those of the reference cycle described in the Oconee 2, cycle 5 reload report.3 However, the group designations dif fer between cycle 6 and the ref-erence cycle to minimize power peaking.

The cycle 6 locations and enrichments I

of the BPRA assemblies are shown in Figure 3-4.

The four BPRAs in core loca-tions C8, H13, 08, and H3 during cycle 5 remained in their fuel assemblies, which were shuffled to locations R8, H1, A8, and H15, repsectively, for cycle 6.

These BPRAs will undergo a second burn during cycle 6.

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I Figure 3-1.

Core Loading Diagram - Oconee 2 Cycle 6 I-A 7

7 7*

7 7

K4 L5 08 Lil K12 7

8 7

8 7

8 7

6 7

lB M4 L3 K2 L13 M12 7

8 7

8 7

8 7

8 7

8 7

C HIS N3 K6 K10 N13 R8 7

8 7

8 6B 8

6B 8

6B 8

7 8

7 D

D11 H11 A7 N2 A9 MS DS 8

7 8

6B 8

6B 8

6B 8

6B 8

7 8

E C12 013 L1 L15 03 C4 7

7 8

6B 8

6B 8

6B 8

6B 8

6B 8

7 7

F D9 C10 C1 M14 P4 P5 GIS C6 D7 7

8 7

8 6B 8

6B 7

6B 8,

6B 8

7 8

7 G

E10 F9 A10 P11 P7 M2 A6 F7 E6 7*

7 8

6B 8

6B 7

6B 7**

6B 8

6B 8

7 7*

Y-H13 P9 P12 N14 K14 L14 G2 D2 B4 B7 H3 g

7 8

7 8

6B 8

6B 7

6B 8

6B 8

7 8

7 K

M10 L9 R10 E14 B9 B5 R6 L7 M6 7

7 8

6B 8

6B 8

6B 8

6B 8

6B 8

7 7

L N4 010 K1 B11

~

B12 E2 K15 06 N7 8

7 8

6B 8

6B 8

6B 8

6B 8

7 8

M 012 C13 F1 FIS C3 04 7

8 7

8 6B 8

6B 8

6B 8

7 8

7

.N11 E8 R7 D14 R9 H5 N5_

7 8

7 8

7 8

7 8

7 8

7 0

A8 D3 G6 G10 D13 HI 7

8 7

8 7

8 7

8 7

P E4 F3 G14 F13 E12 7

7 7*

7 7

R G4 F5 C8 Fil G12 I

z 1

2 3

4 5

6 7

8 9

10 11 12 13 14 15 I

Batch

• H1, A8, H15, and R8 contain reinserted BPRAs.

Previous Core Location

• • H9 location contains Mark BZ demon-l stration assembly (once-burned).

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I Figure 3-2.

Oconee 2 BOC 6 Enrichment and'Burnup Distribution After a 390-EFPD Cycle 5 6

9 10 11 12 13 14 15 2.91 3.07*

2.91 3.17 2.91 3.17 3.07 3.07**

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20091 13823 18798 0

18795 0

13827 15498 I

2.91 3.17 2.91 3.17 3.07 3.17 3.07 K

16795 0

13861 0

16385 0

16284 I

2.91 3.17 2.91 3.17 3.07 3.07 L

16796 0

19317 0

14500 15115 I

2.91 3.17 3.07 3.17 I

• 1 16111 0

12826 0

3.07 3.17 3.07 N

15474 0

15392 I

3.07 O

9608 i

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• Mark BZ demonstration assembly (once-burned).

• • Reinserted BPRA.

235 XXX Initial Enrichment wt %

U XXXXX BOC Burnup, mwd /mtU I

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Oconee 2 Cycle 6 Control Rod Locations and Designations a.

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4 7

4 C

2 6

6 2

D 7

8 5

8 7

E 2

5 1

1 5

2 F

4 8

3 7

3 8

4 G

6 1

3 3

1 6

-Y W-7 5

7 4

7 5

7 H

K e

1 3

3 1

6 L

4 8

3 7

3 8

4 pt 2

5 1

1 5

2 N

7 8

5 8

7 0

2 6

6 2

y 4

7 4

R l

Z l

1 2

3 4

5 6

7 8

9 10 11 12 13 14 15 Croup No. of rods Function I

3 X

Group Number 8

3 8

Safety 4

9 Safety 5

8 Control 6

8 Control 7

12 Control 8

8 APSRs Total 69 3-4 Babcock & Wilcox

I Figure 3-4.

Oconee 2 Cycle 6 BPRA Enrichment and Distribution 8

9 10 11 12 13 14 15 I

Re ins.

H 1.20 1.20 BPRA*

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1.20 1.20 0.20 I

L 1.20 1.20 0.80 I

M 1.20 1.20 1.20 I

N 1.20 1.20 0.20 0

1.20 0.80 0.20 I

P 0.20 I

Reins.

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BPRA*

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1 X.XX LBP Concentration, wt % bgg in Al203 lI

• 0riginally 0.80 wt % bgg in Al203 (at BOC 5).

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4.

FUEL SYSTEM DESIGN 4.1.

Fuel Assembly Mechanical Design The types of fuel assemblics (FAs) and pertinent fuel parameters for Oconee 2, cycle 6 are listed in Table 4-1.

All Mark B FAs are identical in concept and are mechanically interchangeabic.

One reinserted Mark BZ demonstration fuel assembly is included in batch 7.

The Mark BZ is a 15 x 15 fuel assembly similar to the Mark B assembly described in reference 1 except that six intermediate spacer grids are of Zircaloy ma-terial, and an Inconel 718 spring replaced the Inconel X750 holddown spring.

The Mark BZ assembly is described in reference 2, which also states that re-actor safety and performance are not adversely affected by the presence of the one demonstration assembly.

In addition, batch 7 includes three Mark B4 as-I semblies containing the Inconel 718 holddown spring.

Retainers will be used on two batch 7 fuel assemblies that contain regenera-tive neutron sources (RNS), on 64 batch 8 assemblies containing BPRAs, and on the four batch 7 assem1bies with once-burned BPRAs that will be inserted for I

cycle 6.

The justification for the design and use of the retainer is described in references 4 and 5.

4.2.

Fuel Rod Design The fuel rod design and mechanical evaluation are discussed below.

4.2.1.

Cladding Collapse The fuel of batch 6B is more limiting than batches 7 and 8 due to its previous l

incore exposure time. The batch 6B assembly power histories were analyzed to determine the most limiting three-cycle power history for creep collapse. This power history was then compared to a generic analysis to ensure that creep-i ovalization will not affect fuci pwerformance during Oconee 2 cycle 6.

The generic analysis was based on reference 6 and is applicable to the batch 6B design.

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I The creep-collapse analysis predicts a collapse time longer than 35,000 EFPH, which is longer than the expected residence time of 27,432 EFPH (Table 4-1).

,4. 2. 2.

Cladding Stress The Oconee 2, cycle 6 stress parameters are enveloped by a conservative fuel I

rod stress analysis.

No new method was used for analysis of cycle 6 that had not been used on the previous cycle.

4.2.3.

Cladding Strain The fuel design criteria specify a limit of 1.0% on cladding plastic tensile circumferencial strain. The pellet is designed to ensure that plastic clad-ding strain is less than 1% at design local pellet burnup and heat generation The design values are higher than the worst-case values the Oconee 2 rate.

cycle 6 fuel is expected to see.

The strain analysis is also based on the upper tolerance values for the fuel pellet diameter and density and the lower tolerance for the cladding inside diameter (ID).

4.3.

Thermal Design All fuel assemblies in this core are 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 tem-perature for each batch in cycle 6 are shown in Table 4-2.

The maximum fuel rod burnup at EOC-6 is predicted to be 37,046 mwd /mtU. Fuel rod internal pres-sure has been evaluated using TAFY-3 for the highest-burnup fuel rod and was 7

predicted to be less than 2200 psia.

_4. 4.

Material Design The batch 8 FAs are not new in concept, nor do they utilize different compo-nent materials.

One Mark BZ demonstration assembly, described in section 4.1, will be reinserted during cycle 6.

This assembly uses Zircaloy grids and an Inconel 718 spring. Therefore, the chemical compatibility of all possible fuel-cladding-coolant-assembly interactions for the batch 8 FAs are identical to those of the present fuel.

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4.5.

Operating Experience Babcock & Wilcox operating experience with the Mark B fuel assembly has veri-fled the adequacy of its design. As of August 31, 1981, the following experi-ence has been accumulated for the eight B&W 177-FA plants using the Mark B I

fuel assembly:

Maximum assembly #

umulative net (b) burnup, mwd /mtU Current electrical output, Reactor cycle Incore Discharged mWh Oconee 1 7

40,000 40,000 36,855,958 Oconee 2 5

34,778 33,780 31,580,263 Oconee 3 6

29,134 32,061 31,113,594 TMI-1 5

32,400 32,300 23,840,053 I

ANO-1 5

25,000 33,222 27,801,798 Rancho Seco 5

29,493 3~/,730 25,235,809 Crystal River 3 3

28,892 22,389 15,674,038 Davis Besse 1 2

22,904 13,252 9,907,249 I

1. As of August 31, 1981.

As of May 31, 1981.

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I Table 4-1.

Fuel Design Parameters and Dimensions Twiced-burned Once-burned Fresh FAs, g

FAs, batch 6B FAs, batch 7 batch 8 5

FA type Mark B4 Mark B4( )

Mark B4 No. of FAs 37 64/4 72 m

Fuel rod OD, in.

0.430 0.430 0.430 Fuel rod ID, in.

0.377 0.377 0.377 Flex spacers, type Spring Spring Spring Rigid spacers, type Zr-4 Zr-4 Zr-4 Undensif. active fuel 142.25 142.25/

141.80 length (nom), in.

141.38 g

Fuel pellet initial 94.0 94.0/

95.0 5

density (nom), % TD 95.0 Fuel pellet OD (mean 0.3695 0.3695/

0.3686 specification), in.

0.3686 Initial fuel enrichment, 2.91 3.07 3.17 23s wt %

U BOC burnup (avg), mwd /mtU 17,155 14,662 0

Cladding collapse 35,000 35,000 35,000 time, EFPH Estimated residence 27,432 28,650 28,800 time (max), EFPH l

(a) Batch 7 includes one Mark BZ demonstration assembly containing six Zirculoy intermediate spacer grids and an Inconel 718 holddown spring.

Batch 7 al-so includes three Mark B4 assemblies containing the Inconel 718 holddown spring.

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Table 4-2.

Fuel Thermal Analysis Parameters -

Oconee 2 Cycle 6 I

Batch 6B Batch 7 Batch 8 No. of assemblies 37 64(")/4(b) 72 Nominal pellet density, % TD 94.0 94.0/95.0 95.0 Pellet diameter, in.

0.3695 0.3695/0.3686 0.3686 I

Stack height, in.

142.25 142.25/141.38 141.80 Densified Fuel Parameters ("}

Pellet diameter, in.

0.3646 0.364/0.3649 G.3649 Fuel stack height, in.

140.47 140.47/140.32 140.7 Nominal LHR at 2568 MWt, kW/ft 5.80 5.80 5.79 Avg fuel temp at nominal LHR, F 1320 1320/1310 1310 I

LHR to {, fuel melt, kW/ft 20.15 20.15 20.15 Core average densified LHR is 5.80 kW/ft.

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(*) Includes four reinserted LBP assemblies.

} Batch 7 includes one Mark BZ demonstration assembly containing six Zircaloy intermediate spacer grids and an inconel 718 holddown spring. batch 7 also includes three Mark B4 assemblies containing the Inconel 718 holddown I

spring

(' 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 PDQ07.8-io The average cycle burnup will be higher in cycle 6 than in the design cycle 5 because of the longer cycle 6 length.

Figure 5-1 illustrates a representative relative power distribution for the beginning of cycle 6 (BOC-6) at full power with equilib-rium xenon and normal rod positions.

Both cycle 5 and cycle 6 are feed-and-bleed cycles. The differences between the physics parameters of the two cycles are the result of the longer cycle 6 design life, the different LBP loading, and the different shuffle patterns.

The critical boron concentrations for cycle 6 are higher than those for cycle I

5 because the additional reactivity necessary for the longer cycle is not com-pletely offset by the burnable poison. The control rod worths differ between cycles because of changes in the radial flux and burnup distributions. This also accounts for differences in ejected and stuck rod worths. 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.

All safety criteria associated with these worths are i

met.

The adequacy of the shutdrwn margin with cycle 6 stuck rod worths is demonstrated in Table 5-2.

The following conservatisms were applied for the

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shutdown calculations:

1.

10% uncertainty on net rod worth.

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Flux redistribution penalty.

3.

Poison material depletion allowance.

I Flux redistribution was accounted for since the shutdown analysis was calcu-lated using a two-dimensional model. The reference fuel cycle shutdown margin I

is presented in the Oconee 2, cycle 5 reload report.3 I

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I 5.2.

Analytical Input The cycle 6 incore measurement calculation constants to be used for computing core power distributions were prepared in the same manner as those for the a

reference cycle.

5.3.

Changes in Nuclear Design There are no significant core design changes between the reference and reload cycles.- The same calculational methods and design information were used to obtain the important nuclear design parameters for this cycle.

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Oconee 2, Cycle 5 and 6 Physics Parameters ("

I Table 5-1.

Cycle S mid(*)

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Cycle length, EFPD 360 400 Cycle burnup, mwd /mtU 11,266 12.518 Average core burnup EOC, mwd /mtU 20,957 21,736 Initial core loadi s, atU 82.1 82.1 Critical boron, BOC (no Ze), ppm I

HZP, group 8 inserted 1459 1552 HFP, group 8 inserted 1263 1344 Critical boron EOC (eq Xe), ppm 87 4

group 8 inserted, eq Xe Control rod worth, HFP, BOC, % Ak/k Group 7 1.60 1.47 Grono 8 0.41 0.37 Control rod worth, HFP, EOC. % Ak/k Group 7

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1.59 1.53 I.

Group 8 0.49 0.49 Max ejected rod worth, HZP, % Ak/k(*}

BOC 0.71 0.63 EOC 0.62 0.57 I

Max stuck rod worth. HZP, % Ak/k BOC 1.93 1.78 EOC 1.73 1.80 I

Power deficit, HZP to HFP, % Ak/k BOC

-1.60

-1.56 EOC

-2.27

-2.38 I

Doppler coeff, 10-5 (Ak/k *F)

BOC,100% power, no Xe

-1.57

-1.52 EOC, 100% power, eq Xe

-1.69

-1.77 Moderator coeff, ITP, 10-" (ak/k *F)

I BOC (group 8 f., no Xe, 1263 ppm boron)

-0.61

-0.63 EOC (group 8 in, eq Xe, 17 ppm boron)

-2.87

-2.98 Boron worth, HFP, ppm /%(Ak/k)

I BCC (950 ppm boron) 116 124 EOC (17 ppm boron) 106 107 Xenon worth, HFP, % Ak/k I

BOC (4 EFPD) 2.62 2.56 EOC (equilibrium) 2.73 2.70 Eff delayed neutron fraction, HFP BOC 0.00613 0.00628 EOC 0.00526 0.00522

• } Cycle 6 data are for the conditions stated in this report. The cycle 5 core conditions are identified in reference 3.

} Cycle 5 data are based on a cycle 4 length of 353 EFPD.

(c) Cycle 6 data are based on a cycle 5 length of 390 EFPD.

(d)HZP denotes hot zero power (532F T**E), MFP denotes hotfull power (577,F T,

).

I") Ejected rod worth for groups 5 through 8 inserted.

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I Table 5-2.

Shutdown Margin Calculation -

Oconee 2 Cycle 6(a) 1I BOC, % Ak]k EOC(#}, % Ak /k Available Worth Total rod worth, HZP(

9.04 9.42 Worth reduction due to burnup of poison material

-0.42

-0.42 Maximum stuck rod, HZP

-1.78

-1.80 Net worth 6.84 7.20 Less 10% uncertainty

-0.68

-0.72 Total available worth 6.16 6.48 Required Rod Worth Power deficit, HFP to HZP 1,56 2.38 Max allowable inserted rod worth 0.25 0.50 Flux redistribution 0.61 1.20 Total required worth 2.42 4.08 Shutdown Margin Total available worth minus total required worth 3.74 2.40

(^) Based on a cycle 5 length of 390 EFPD.

(b 'T HZP: hot zero power, HFP: hot full power.

(C 400 EFPD.

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Note: Requirer shutdown margin is 1.00% Ak/k.

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Figure 5-1.

BOC 6 (4 EFPD) Two-Dimensional i Tlative Power Distribution - Full Power, Equil briun Xenon.

Normal Rod Positions, Group 8 Inserted 8

9 10 11 12 13 14 15

,I 0.945 1.107 1.055 1.280 1.097 1.253 1.037 0.500 I

K 1.046 1.250 1.163 1.236 1.174 1.123 0.508 X

L 1.113 1.224 0.926 1.196 0.906 0.398 I

M 1.082 1.201 1.109 0.899 I

N 1.149 1.047 0.487 I

0 0.605 i

?

I R

lI I

• Mark BZ demonstration assembly (once-burned).

Inserted rod group No.

XXX Relative Power Density 1

(I 5-5 Babcock & Wilcox i.

I I

I I

6.

THERMAL-HYDRAUI.IC DESIGN The incoming batch 8 fuel is hydraulically and geometrically simitar to the fuel remaining in the core from previous cycles. The thermal-hydraulic de-sign evaluation supporting cycle 6 operation utilized the methods and models described in references 1, 3, and 11 except for the core bypass flow.

The maximum core bypass flow due to the insertion of 56 BPRAs in cycle 5 was 8.1%.

For cycle 6 operation, 64 fresh and 4 once-burned BPRAs will be in-serted, leaving 38 FAs with unplugged guide tubes, resulting in a decrease in maximum core bypass flow to 7.6%.

This decreased bypass flow and the conse-quent increase in core flow establish the cycle 5 analysis as conservative for I

and applicable to cycle 6 operation. The reinserted BPRAs do not impact the thermal-hydraulic analysis. The cycle 5 and 6 maximum design conditions and I

significant parameters are shcwn in Table 6-1.

The Mark BZ low-absorption demonstration assembly will be limited to a 1.54 I

design peak to ensure that this assembly is not limiting.

The 1.71 design radial-local peak remains valid for all other assemblies.

The fuel rod bow penalty fer cycle 5 has been calculated with the procedure approved in reference

1. 2.

Using this procedure, the penalty is based on the maximum assembly burnup for the fuel batch containing the limiting (highest power) assembly. The calculated penalty is then reduced by 1% to reflect a generic credit, which is available because of the use of a flow area (pitch) reductior. factor in the thermal-hydraulic analysis.

For cycle 6 operation the act rod bow penalty that results from this calculation is 0.4% DNBR based on the maximum assembly burnup (19,.30 mwd /mtU) of batch 8 fuel.

Additional analyses were performed for the limiting assemblies in fuel batches 6B and 7 I

(based on steady-state power distributions) to demonstrate that the increase in DNBR associated with the lower peaking of e assemblies (relative to I

the limiting batch 8 assembly) more. than of f sats the increased rod bow DNBR pe. cities that would be calculated on the basis of maximum assembly burnup values for these batches.

6-1 Babcock & Wilcox I

i I

t l

Table 6-1.

Thermal-Hydraulic Design Conditions W

Cycles 5 and 6 Power level, MWt 2568 Systcm pressure, psia 2200 Reactor coolant flow, % design flow 106.5 Vesse: inlet coolant temp., 100%

555.6 g

power, F m

Vessel outlet coolant temp., 100%

602.4 power, F Ref design axial flux shape 1.5 cos Ref design radia]-local power 1.71 g

peaking powe B

Active fuel length, in.

(a) i Avera e heat flux, 100% power, 176(b) 2 10 Btu /h-ft CHF correlw.lon BAW-2 Hot channel factors e

Enthalpy rise 1.011 Heat flux 1.01%

g Flow area OM g

Minimum DNBR with densification 2.' a Penalty l

(")See Table 4-2.

( ) Based on densified length of 140.3 inches.

I I

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I I

6:2 Babcock a,Wilcox

I I

I 7.

ACCIDENT AND TRALSIENT ANALYSIS 7.1.

General Safety Analysis I

Each FSAR accident analysis has been examined with respect to chenges in cycle 6 parameters to determine the effect of the cycle 6 reload and to ensure that thermal performance during hypochetical transients is not degraded. The effects of fuel densification on the FSAR accident results have been evaluated and are reported in reference 11.

Since batch 8 reload fuel assemblies con-tain fuel rods with a theoretical density hi her than those considered in ref-F erence 11, the conclusions in that reference are still valid.

7.2.

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

Fuel thermal analysis parameters for each batch in cycle 6 are compared in Table 4-2.

A comparison of the cycle 6 thermal-hydraulic maximum design con-ditions to tha previous cycle 5 values is presented in Tabic 6-1.

The key kinetics parameters from the FSAR and cycle 6 are compared in Table 7-1.

A generic LOCA analysis has been performed for the B&W 177-FA, lowered-loop NSS using the Final Acceptance Criteria ECCS Evaluation Model; this study is reported in BAW-10103, Rev. 3.13 The anal sis in BAW-10103 is generic since

/

the litriting values of key parameters for all plants in this catege y were I

used.

Furthermore, the combination of average fuel temperature as a function of LHR and the lifetime pin pressure data used in the BAW-10103 LOCA limits analysis is conservative comparc3 to those calculated for this reload. Thus, the analysis and the LOCA limits reported in BAW-10103 provide conservative results fec the operation of Oconee 2 cycle 6 fuel, except as noted in the follo ing parageaph.

7-1 Babcock & Wilcox

1 I

Table 7-2 shows the b unding vaMs for allowable LOCA peak LHRs for Oconee 2 cycle 6 fuel af ter 50 U7D The IDCA kW/f t limits have been reduced for the first 50 EFPD. The reduction will ensure that conservative limits are main-tained while a transition is being made in the fuel performance codes that provide input to the ECCS analysie in order to account for mechanistic fuel l

densification. The limits for the first 50 EFFD are shown in Table 7-3.

From the examinations of cycle 6 core thermal properties and kinetics proper-ties with respect to acceptable previous cycle values, it is concluded that this core reload will not adversely affect the ability to operate the Oconee 2 plant safely during cycle 6.

Considering the previously accepted design basis used 1. the FSAR and subsequent cycles, the tre:.sient evaluation of cycle 6 is considered.o be bounded by previously accepted analyses. The in-p 1

itial conditions of the transients in cycle 6 are bounded by the FSAR, the fuel densification reportil, and/or subsequent cycle analyses.

The radiological dose consequences of the accidents presented in Chapter 14 of the FSAR were re-evaluated for this reload. The reason for the re-evalua-tion is that, even though the FSAR dose analyses used a conservative basis for the amount of jlutonium fissioning in the core, improvements in fuel manage-ment technigtes have increased the amount of energy produced by fissioning plutonium.

Since plutonium-239 has different fission yields than uranium-235, t'

mixture cf fission product nuclides in the core changes slightly as the plu;. onium-239/uraniu=-235 fission ratio charce.S.

i.e., plutonium fissions pro-g duce more of so=c nuclides and less of others. The general trend is that more plutoniu-fissions tend to produce slightly higher thyroid doses and slightly lower whole body doses.

But, since the radiological doses associated with 5

each accident are impacted to a different extent by each nuclide and by vari-ous mitigating f actors and plant design feat tres, the radiological consequences of the FSAR accidents were recalculated using the specific cycle 6 parameters in erder to obtain an accurate evaluation of the effects of the increase in the amount of plutonium fissioning. The bases used in the dose calculations are identical to those presented in the FSAR except for the following two no-table differences:

1.

The fission y'. elds and half-lives used in the new calculaticns are based on more current data.

E g,

7-2 Babcock & Wilcox

F I

I 2.

The steam generator tube rupture accident evaluation considers the amount of steam released to the environment via the main steam re-lief valves because of the slower depressurization due to the re-I duced heat transfer rate caused by tripping of the reactor coolant pumps upon actuation of the high-pressure injection (a post-TMI-2 modification).

Table 7-4 shows the radiological doses presented in the FSAK nd thore cal-culated specifically for cycle 6.

As can be seen from the table, some doses I

are slightly higher and some are slightly lower than the FSAP values; however, all doses are well below the 10 CFR 100 limits of 300 Rem to the thyroid and 25 Rem to the whole bcJy. The small increases in some doses are essentially offset by a reduction in others. Thus, the radiological impact of accidents during cycle 6 would not be significant y differen than those described in Chapte r 14 of the FSAR.

I I

Table 7-1.

Comparison of Key Paramete a for Accident Analysis FSAR and Pudicted densification cycle 6 Parameter report value value BOL Doppler cceff, 10-5 Ak/k *F

-1.17(#}

-1.52 EOL Doppler coeff, 10-5 Ak/k *F

-1.33

-1.77 fD)

BOL moderator coeff, 10-" Ak/k *F

+0.5

-0.63 EOL moderator coef f,10-" Ak/k *F

-3.0

-2.98 G?

All rod bank worth, HZP, % Ak/k 10.0 9.04 g

Boron reactivity worth (cold),

75 87 pp:/ l*;(J k/k)

I Max ejected rod worth, HFP, % Ak/k 0.65 0.32 Dropped rod worth, HFP, % Ak/k 0.46 0.20 Initial boron conc, HFP, ppm 1400 1344 I

(*) 1.2 x 10-s Ak/k-F was used for steam line failure analysis, and -1.3 x 10-5 Ak/k *F was used for cold water analysis.

(b)+0. 94 x 10-" Ak/k *F was used for the moderator dilution accident.

I 7-3 Babcock & Wilcox

4 i

Table 7-2.

W CA Limits, Oconee 2 Cycle 6, After 50 EFPD Allowable Core peak LHR, elevation, fc kW/ft 2

15.5 g

4 16.6 6

18.0 8

17.0 10 16.0 5

Table 7-3.

WCA Limits, Oconee 2 Cy cle 6, O to 50 EFPD Allowabic Core peak LliR, elevation, ft kW/ft 2

14.5 4

16.1 6

17.5 8

17.0 10 16.0 l

I E

I

.i I

E E

Babcock & Wilcox 7-4

I Table 7-4.

Co.nparison of FSAR and Cycle 6 Accident Doces FSAR doses, Cycle 6 doses, I

Accident Rems Rems Fuel handling 8

Thyroid at EAB 0.43 0.50 Whole body at EAB 0.027 0.028 I

Steam generator tube failure Thyroid at EAB 0.00034 0.310 Whole body at EAB 0.023 0.058 Wtste gas tank Thyroid at EAB 0.13 0.27 Whole body at EAB 0.19 0.17 Rod ejection accfdent Thyroid at EAB 0.19 0.21 Wheele body at EAB 0.001 0.0005 Thyroid at LPZ 0.22 0.23 Whole body at LPZ (a) 0.0007 Steam line 'rreak Thyroid at EAB 0.19 0.20 Whole body at EAI 0.002 0.002 LOCA Thyroid at EAB 4.6 5.0 I

Whole body at EAB 0.01 0.01 l

Thyroid at LPZ 5.0 5.5 Whole body at LPZ 0.014 0.014 I

MHA Thyroid at EAB 186.

193.

Eole bcdy a*.

EAB 1.4 1.4 I

Thyroid at LPZ 144.

180.

Whole body at LPZ 0.65 0.62 l

(" Not reported in tne FSAR.

I I

l

(

7-5 Ba'acock s. Wilcox

! g

W

- _ - _.---_ _._..-.-_j

i I

l I

I 8.

PROPOSED MODIFICATIONS TO TECHNICAL SPECIFICATIONS The Technical Specifications for cycle 6 operation have been revised in ac-cordance with the mothods of references 15 through 17 to account for power peaking and control rod worths inherent in an extended, lumped burnable poison cycle.

In addition:

1.

A high flux trip setpoint of 104.9% of rated power, and r high RC outlet temperature trip of 618F have been established tot cycle 6 operat1on.

2.

The 0-50 EFPD operating limits on rod index, APSR position, and axial power imbalance were established based on the interim LOCA linear heat rate linits, which account for mechanistic fuel densification ".

l After 50 EFTD, the FAC LOCA LHR limits were used.13 Based on the Technical Specifications derived from the analysis precented in this report, the Final Acceptance Criteria ECCS limits will not be exceeded, nor will the thermal design criteria be violated.

Figures 8-1 through 8-19 are revisions to previous Technical Specification limits.

I I

I I

I I

8-1 Babcock & Wilcax I

I, Fi gure 8-1.

Core Protection Safety Liciits for Oconee Unit 2, Cycle 6 THERMAL P0nER LEVEL, t FP UNACCEPTABLE E

(

. 120 CPERATICN 5

( 41.0.I12.0)

(42.0.112.0)

(-42.0.lu...[8 ACCEPTABLE g

l ll 4 PUMP I

I OPERATION 100

(

ll l

(59.0.93.3) g i

II t-41.0.90.606)

(47.0.90.006) 3

(-42.0,E5.906)lj

[g ACCEPTABLE l

I l

l 3 & 4 PUWP 80 I

l 5

CPERATION UNACCEPTAELE gg II I

(59.0.71.90E)

I!

(-41 0 63 150)

(42.0.E3.150) l g

i

\\

l i 60 ACCEPTA3LE

( 42 0.55.45) l 2.3 & 4 PUMP

\\

g l

5 I!

CFERATION (59.0,44.45) l i

I ll ll l

I I

I!

-- 20 l

I I

l i II.

l, g

-E3

-40

-20 0

+20

+40

+60 Rea:ter Poner Ir.calance, t l

I l

I E

E 8-2 Babcock & Wilcox

~

I I

Figure 8-2.

Core Protection Safety Limits for Oconee Unit 2, Cycle 6 2400 ACCEPTABLE OPERATION 2200 E

N 2000 S

U

/

I g

g2 3

I e

a 1800 I

1600 560 580 600 620 640 Reactar Coolant Outlet Temperature F I

l CURVE COOLANT FLOR. GPN PDAER, t P!!MPS OPERATING TYPE OF Li(IT 1

374,880 (1005) 112 4

ONBR 2

280,035 (74.7%)

90.606 3

DNPR 3

103,690 (49.0s) 63.150 1 PER LOOP 00AllTY l

iI I

I I

8-3 Babcock & Wilcox l

l

l I

Figure 8-3.

Maximum Allowable Setpoints for Oconee Unit 2, Cycle 6 THERMAL POWER LEVEL, 5 FP l

l1 (17,108.0)

(-17,108.0)

ACCEPTABLE 2

-10 M

Mi = +1.0 I4 PUNP

--100 g

UNACCEPTABLE g

OPERATION

(-28,97.0) 1 OPERATION l

l I

l l

l UNACCEPTABLE l

l( 17,80.6) 80 (17.80.6)

(42,83.0)

OPERATION l

" ACCEPTABLE l

l I 3&4 PUMP l

l

(-28,69.6) 10PERATION 1

g l

I y

- 60 I

l I

l 1

l(-i7.s2.9) i(i7.s2.9)

(42.ss.8)

ACCEPTABLE l

g I

2, 3 f.4 l

5

(-28,41.9) 10 RATif t:

40 I

I I

I l

I 1

(42,27.9) lcli I

20 I

l i

1 i~.

I

=

I in In n

n i

g a

~;

i n i e l

,I l f

'I i

-40

-20 0

+20

+40 Reactor Power imoalance, %

I E

. E 8-4 Babcock & Wilcox

e I

Figure 8-4.

Protective System Maximum Allowable Setpoints, Oconee 2 Cycle 6 2400 I

2300 T = 61B*F P = 2300 PSIG I

y 2200 ACCEPTABLE a

OPERATION E

g I

2100

~

I R

R I

2000 b

as UNACCEPTABLE A*

OF_ RAT 10N 1

~

~-

//

1900 4

P = 1B00 PSIG 1800 T = 584F 1

I I

540 560 580 5

Reactor Outlet Temperature, F

I 8-5 Babcock 8.Wilcox 8

Figure 8-5.

Oconee 2 Cycle 6 Rod Position Limits - Four-Pump operatlon, O to 50 t to F.FPD SHUT 00WN MARGIN,

LIMIT 100 (II I

)

POWER l

OPERATION IN THIS LEVEL (277.5.92)

REGION NOT AL101E0 CUT 0FF = 1000 FP 80 hfh'fh (264 5.80) r

-g

,KW/FT LIMIT m

60

?

(58,50)

(200,50)

S

[

40 j

PERMISSIBLE OPERATING 20 REGION f

(0,15.3) i (0.2.5)k O

0 20 40 00 80 100 120 140 100 180 200 220 240 260 280 300 4

0 25 50 75 100 Rau Index, t Witndrawn g

i i

i i

g Group 5 N

0 2,5 5,0 7,5 100 se P

E Group 6 0

25 50 75 100 g

e i

i i

j Group 7 I

a

m W

M M

M M

M M

M W

M M

M M

M M

M M

M Figure 8-6.

Oconee 2 Cycle 6 Rod Position Limits - Four-Pump operation, 50 i 10 to 225 10 EFPD i

100 (184,102)

(274.102).

POWER LEVEL (271'02)

SHUTOOWN CUT 0FF = 1000 FP l

MARGIN LIMIT 80 (258,80) x 8

OPERATION IN THIS RESTRICTED KW/FT N

REGION NOT ALLOWED OPERATION LIMIT 60 e

(118,50)

(200,50)

?

I 40 PERMISSIBLE OPERATING REGION 20 (0,8.1 (57,15) 1 0

i i

r i

e j

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

,0 2,5 5,0 75 100 Rod index, t Witnarawn cx2 g

l 8

Group 5 E-0 25 50 75 100 e.

i M_

Group G 8

0 25 50 75 100 x

Group 7

Figure 8-7.

Oconec 2 Cycle 6 Rott Position Limits - Four-Pump Operation After 225 10 EFPD 120 (274,102) 100 (243,102)

(264.5,92)

CUT 0FF = 1000 FP SilVTDOWN w

MARGlil LIMI 80 OPERATION IN THIS

~

(

}

REGION NOT ALLOWED W/FT LIMIT i

E 60

(

%I (176,50)

(200.50)

J

?

=

40 l

PERNISSIBLE OPERATING REGION 20 (101.15)

(

' }

\\.

0 0

20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 l

0 25 50 75 100 Rad inaex, % Witnurann l

l Group 5 o

Q.

0 25 5,0 75 10,0

E Grcup 6

??

0 25 50 75 100 l

g i

i i

i i

Group 7

I Figure 8-8.

Oconee 2 Cycle 6 Rod Position Limits - Three-Pump Operation, O to 50 1 10 EFPD 100 I

80 (117,77)

(258,77)

SHUTDOWN "I

RESTRICTED OPERATION

=

KW/FT 60

- OPERATION iN LIMIT e

I THIS REGION NOT E

ALL0nED (200,53) a 40 (58,38) g E

20 FERMISSIBLE OPERATING REGION (0,11.98)

(90,15)

I (0,2.

0 i

f i

0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 Rod Index, *, Witndrawn I

O 25 50 75 100 t

i f

f Group 5 I

0 25 50 75 100 i

f f

f f

Group 6 I

0 25 50 75 100 f

I f

f f

Group 7 I

I I

I 8-9 Babcock & Wilcox

l Figure 8-9.

Oconee 2 Cycle 6 Rod Position Limits - Three-Pump Operation From 50 10 to 225 ! 10 EFPD 100 I

80 (184,77)

(252,77)

I b

RESTRICTED T

SHUTOOWN

=

OPERATION 60 MARGIN

=

g OPERATION IN LIMIT THIS REGION (200,50)

NOT ALLOWF0 40 (118,38)

PERMISSIBLE OPERATING 20 REGION E

(57,11.75)

,6.6) i-0 0

20 40 61 80 100 120 140 160 180 200 220 240 260 280 300 l

0 25 50 75 100 Group 5 g

0 25 50 75 100 g

f f

f f

f Group 6 3

0 25 50 75 100 g

f f

I f

Group 7 I

I I

I g

8-10 Babcock & Wilcox

I Figure 8-10.

Oconee 2 Cycle 6 Rod Position Limits - Three l'um p Operation After 225 t 10 EFPD 100 80 (243,77)

. ! (250,77)

SHUT 00NN MARGIN I

j LIMIT KW/FT ii_:TRICTED LIN i 60 OPERATION IN THIS I,

y OPERATION REGION NOT ALLOWE0 (200,50)

I a

40 g

(176,38) 1 20 PERMISSIBLE OPERATING REGION (0,4.5)

(101,11.75) 0 0

20 40 60 80 100 120 140 160 180 200 220 240 260 280 30d I

O 25 50 75 100 t

I t

i I

Group 5 5

0 25 50 75 100 t

t I

f f

\\

Group 6 I

O 25 50 75 100 i

f I

f

__.]

Group 7 I

I I

l1 8-11 Babcock & Wilcox u

I Figure 8-11.

Oconee 2 Cycle 6 Rod Position Limits - Two-Pump Opetation, O to 50 1 10 EFPD 100 I

I I

80 SHUTOOWN MARGIN

=

LIMIT-E 60 (204,52) gE N

OPERATION ils

~

(11,52)

E FEGION NO.

a ALLOWED (200,50) 40 RESTRICTED KW/FT LIMIT o

OPERATION a.

10,J.S s (58,26)

PERMISSIBLE DPERATING 20 0,15) 0p0,2.

.g 1

+

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

n 25 50

,75 100 30d Index, to Witndrawn l,

l Group 5 0

25 50 75 100 gg t

I t

Group 6 0

25 50 75 100 gg a

f f

f I

GIOup 7 E

I I

I

.E 8-12 Babcock & Wilcox

r-1 I

Figure 8-12.

Oconec 2 Cycle 6 Rod Position Limits - Two-Pump Operation From 50 1 10 to 225 ! 10 EFPD I

100 I

I S

E RESTRICTED 60 - CPERATION IN THIS (204,52)

REGION NOT (184,52)

ALLOWED g

40 MARGIN LIMIT LIMIT SHUTDOWN 200,59)

KW/FT r

E i

(

20 I

PERMISSIBLE OPERATING REGION

~

(0,5.1)

(57.8.5) 0 0

20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 R00 Index, t Witnarawn 0

25 50 75 100 i

m' I

i GIOUP 5 0

25 50 75 100 I

i t

f a

l GIOUp 6 0

25 50 75 100 f

f f

a 1

GIOUp 7 I

I I

I 8-13 Babcock & Wilcox

I Figure 8-13.

Oconee 2 Cycle 6 Rod Position Limits - Two-Pump Operation After 225 10 EFPD 100 I

80 I

E 5

=

=

60 o

4 (243,52) g SHUTOOWN MARGIN 3

6 40 LIMli g

n.

OPERATION IN THIS REGION (176,26) 20 NOT ALLOWED PERMISSIBLE OPERATING REGION (0,3.7)

(101,8.5)

I 0

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

25 50 75 100 Gr00p 5 0

25 50 75 100 Gr0up 6 5

0 25 50 75 100 g

i I

f Group 7 I

I I

I 8-14 Babcock & Wilcox

I Figure 8-14.

Oconee 2 Cycle 6 Operational Power Imbalance Limits, O to 50 10 EFPD

(-I7.5. 102)

(47.5.102)

C q

RESTRICTE0 100 I

REGION

(-17.5,92)

(+20.92)

I

(- 8,80)

PERMISSIBLE

-- 80

> (+22,80; OPERATING I_

REGION

-- 70 E

E -- 60 o

a f-- 50 m

i

-- 40

-- 30 E

- - 20 I

10

-40

-30

-20

-10 0

10 20 30 40 Axial Poner imoalance, f; I

8-15 Babcock & Wilcox

1 I

Figure 8-15.

Oconee 2 Cycle 6 Operational Power Imbalance Limits, 30 2 10 to 225 ! 10 EFPD

(-20,102)

( +17. 5,10 2 )

RESTRICTED

-- 100 REGION

(-24,92)

(+20,92)

I

('

)

- - 80

(+22,80)

PERMISSIBLE JPERATING l

REGION

-- 70

=

=

$ -- 60 1

2 O

A'

=

. -- 50 1

~~ 40

-- 30 I

i

-- 20 I

- - 10 3

-40

-30

-20

-10 0

10 20 30 40 Axial Power inoalance, ',

g 8-16 Babcock & Wilcox

I Figure 8-16.

Oconee 2 Cycle 6 Operational Power Imbalance Limits After 225 10 EFFD RESTRICTE0

(.20,102)

( +17. 5.10 2 )

REGION

- 100

(-22,92)

( +20,92) l

-- 90 I

(-32,80)

- - 80

(+22,80)

PERMISSIBLE OPERATING REGION

-- 70 E

=

s Eg

-- 60

~

o

'k 1

k d

-- 50 5

-- 40 l

-- 30

-- 20 l I

- - 10 I

l i

e i

r 40

-30

-20

-10 0

10 20 30 40 l

Axial Power Imualance, 5 lE 8-17 Babcock & Wilcox

i l

I Figure 8-17.

Oconee 2 Cycle 6 APSR Position Limits, O to 50 2 10 EFPD (4,102)

(35,102) 100 C

N (9,92)

(36,92; 80 (

(0,80)

(42,80)

RESTRICTE0 REGION E

m M

60

~

(100,50)

J 40 PERMISSIBLE OPERATING REGION 5'

f 5

29 t

0 0

20 40 60 30 100 APSR Position, t nitnaraan B'

E I

I I

l l

I 8-18 Babcock & Wilcox

tI Figure 8-18.

Oconee 2 Cycle 6 APSR Position Limits, 50 ! 10 to 225 ! 10 EFPD (9,102)

(35,102)

(9,92)

(37,92) 80 (

(0,89)

RESTRICTED l

REGION

'l E

60 (100,50) 40 a

PERMISSIBLE OPERATING REGION 20 5

0 l

0 20 40 60 80 100 APSR Position, litndrawn I

I I

I I

I 8-19 Babcock & Wilcox

I Figure 8-19.

Oconee 2 Cycle 6 APSR Position Limits After 225 1 10 EFPD I

l (8.5.102)

(33.5,102) 100

- C l

(8.5,92)

(42,97) 80 (

(0.80)

(45,80)

RESTRICTED REGION l

=

60

=

[

(100,50)

W D

PERWISSIBLE OPERATING REGION o

40 0

B 20 5

0 0

20 40 60 80 100 APSR Position, % Witndrawn I

I l

l I

l I

I I

I 8-20 Babcock & Wilcox

r-

~

I I

REFERENCES _

1 Oconee Nuclear Station, Units 1, 2, and 3 - Final Safety Analysis Reports, Docket Nos. 50-269, 50-270, and 50-287, Duke Power Company.

2 Mark-BZ De:ronstration Assembly - Licensing Report, BAW-1533, Rev 1, Babcock

& Wilcox, Lynchburg, Virginia, March 1980.

8 Oconec Unit 2 Cycle 5 Reload Report, BAW-1565, Rev 1. Babcock 6 Wilcox, Lynchburg, Virginia, March 19a0.

" BPRA Retainer Design Report, BAW-li96, Babcock & Wilcox, Lynchburg, Virginia, May 1978.

5 J. H. Taylor (B&W) to S. A. Varga (NRC), letter, "BPRA Retainer Reinser-t ion, January 14, 1980.

8 Program to Determine In-Reactor Performance of B&W Fuels - Cladding Creep Collapse, BAW-10084A, Rev 2, Babcock & Wilcox, Lynchburg, Virginia, October I

1978.

7 C. D. Morgan and H. S. Kao, TAFY - Fuel Pin Temperature and Gas Pressure I

Analysis, BAW-10044, Babcock 6 Wilcox, Lynchburg, Virginia, May 19/2 B&W version of PDQ07 Code, ' LAW-10ll JA, Babcock 6 Wilcox, Lvnchburg, Virginia, January 1977.

' Core Calculational Techniques and Prcaedures, BAW-10ll8, Babcock & Wilcox, Lynchburg, Virginia, October 1977.

10 Assem'aly Calculations and Fitted Nuclear Data, BAW-10ll6A, Babcock 6 Wilcox, Lynchburg, Virginia, May 1977.

11 Geonee 2 Fuel Densification Report. BAW-1395, Babcock & Wilcox, Lynchburg, Virginia, June 1973.

12 L. S. Rubenstein (NRC) to J. H. Taylor (B&W), " Evaluation of interim Proce-dure for Calculating DNBR Reductioas Due to Rod Bow," October 18, 1979.

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13 ECCS Analysis of B&W's 177-FA Lowered-Loop NSS, BAW-10103A, Rev 3, Babcock

& Wilcox, Lynchburg, Virginia, July 1977.

1" N. H. Taylor (B&W) to L. U, Rubenstein (NRC), Let ter, September 5,1980.

l 15 Power Peaking Nuclear Reliability Factors. BAW-10119, Babcock & Wilcox, Lynchburg, Virginia. January 1977.

l' Normal Operating Controls, BAW-10122, Pabcock & Wilcox, Lynchburg, Virginia, August 1978.

m Verification of the Three-Dimensional FLAME Code, BAW-10125A, Babcuck &

Wilcox, Lynchburg, Virginia, August 1976.

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