ML19352B163

Cycle 7 Reload Rept.

ML19352B163
Person / Time
Site:	Oconee
Issue date:	03/31/1981
From:	BABCOCK & WILCOX CO.
To:
Shared Package
ML15223A731	List: ML15223A730 ML15223A732 ML19352B163
References
BAW-1660, NUDOCS 8106030246
Download: ML19352B163 (49)

Text

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BAW-1660 March 1981 is il i

OCONEE UNIT 1, CYCLE 7

- Reload Report -

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BABCOCK & WILCOX j Nuclear Power Group l l Nuclear Power Generation Di'rision i P. O. Box 1260 Lynchburg, Virginia 24505 l Dabcock & Wilcox

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CONTENTS j Page

1. INTRODUCTION AND

SUMMARY

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2. OPERATING HISTL.lY . . . . . . . . . . . . . . . . . . . . .... 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 . . . . . . . . . . . . . . . . . . ....

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( 4.2.1. CladdinS Collapse . . . . . . . . . . . . . .... 4-1

4.2.2. Cladding Stress . . . . . . . . . . . . . . .... 4-2 4.2.3. Cladding Strain 4-2 E

. . . . . . . . . . . . . . . ... g l 4.3. Thermal Design . . . . . . . . . . . . . . . . . . . .... 4-2 4.4. Material Design . . . . . . . . . . . . . . . . . . .... 4-3 l 4.5. Operating Experience . . . .. . . . . . . . . . . . .... 4-3 t 5. NUCLEAR DESIGN . . . . . . . . . . . . . . . . . . . . . . .... 5-1 5.1. Physics Characteristics . . . . . . . . . . . . . . .... 5-1 1 5.2. Analyti?al Input . . . . . . . . . . . . . . . . . . .... 5-2 l 5.3. Changes in Fuclear Design . . . . . . . . . . . . . .... 5-2 i

6. THERMAL-HYDRAULIC DESIGN . . . . . . . . . . . . . . . . . .... 6-1

7. ACCIDENT AND TRANSIEN1 ANALYSIS . . . . . . . . . . . . . .... 7-1 7.1. General Safety Analysis . . . . . . . . . . . . . . .... 7-1 i 7.2. Accident Evaluation . . . . . . . . . . . . . . . . .... 7-1 l

8. PROPOSED MODIFICATIONS TO TECHNICAL SPECIFICATIONS ... 8-1 l

. . .... 5 REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . .... 9-1 m P

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I List of Tables Table Page I 4-1. Fuel Design Parameters and Dimensions ... ... ....... 4-4 4-2. Fuel Thermal Analysis Parameters - Oconee 1, Cycle 7 . . .... 4-5 I 5-1.

5-2.

6-1.

Oconee 1 Physics Parameters . . . .. . . . . . . . .. ....

Shutdown Margin Calculation for Oconee 1, Cycle 7 Thernal-Hydraulic Design Conditions .. .. . .. ..... ,.

5-3 5-4 6-2 7-1. Comparison of Key Parameters for Accident Analysis . ..... . 7-4 I 7-2.

7-3.

LOCA Limits, Oconee 1, Cycle 7, After 50 EFPD .. . ... ...

LOCA Limits, Oconee 1, Cycle 7, 0-50 EFPD ... ... .. ...

7-4 7-5 7-6 7-4. Comparison of FSAR and Cycle 7 Accident Doses . . ... ....

I List of Figures Figure 3-1. Core Loading Diagram for Oconee 1, Cycle 7 . . . ....... 3-2 3-2. Enrichment and Burnup Dist.ribution for Oconee 1, Cycle 7 3-3 I

3-3. Control Rod Locations for Oconee 1, Cycle 7 . . . . . ..... _3-4 3-4. BPRA Concentration and Distribution for Oconea 1, Cycle 7 . . . 3-5 5-1. Oconee 1 Cycle 7 BCC Two-Dimensional Relative Power Distribution - Full Power, Equilibrium Xenon, Normal 5 Rod Positions . . ... . . . ... . .. .. . . . .. .... 5-5 8-1. Core Protection Safety Limits for Oconee Unit 1.... .... 8-2

. 8-2. Protective System Maximum Allowable Setpoints for Oconee Unit 1... . . . . .. . ... .. ........ 8-3 8-3. Rod Position Limits for Four-Pump Operation, 0-50 EFPD, Oconee 1, Cycle 7 . . . . .. . . . . . .... . .... 8-4 I 8-4.

8-5.

Rod Position Limits for Four-Pump Operation, 50-200 EFPD, Oconee 1, Cycle 7 . . . . . . . . . . . . . .

Rod Position Limits for Four-Pump Operation After

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'I 6-6.

200 EFPD, Oconee 1, Cycle 7 .

Rod Position EFPD, Oconee Limits for 1, Cycle 7 Three-Pump Operation, 0-50 8-6 8-7 8-7.

I Rod Position Limits for Three-Pump Operation, 50-200 EFPD, Oconee 1, Cycle 7 . . . ... . . . . .. . ... .... 8-8 8-8. Rod Position Limits for Three-Pump Operation After 200 EFPD, Oconee 1, Cycle 7 . . .. . . . ... . . . .. .... 8-9

8-9. Rod Position Limits for Two-Pump Operation, 0-50 EFPD, Oconee 1, Cycle 7 . . . . . . . .. . .. . .. .... . 8-10 8-10. Rod Positicn Limits for Two-Pump Operation, 50-200 8-11

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EFPD, Oconee 1, Cycle 7 . . . . .. . . . . . .... .....

Rod Position Limits for Two-Pump Operation After 200 EFPD, Oconce 1, Cycle 7 . . .. . . . .. .. .. . .... 8-12

8-12. Power Imbalance Limits for 0-50 EFPD, Oconee 1, Cycle 7 . ... 8-13 8-13. Power Sbalance Limits for 50-200 EFPD, Oconec 1, Cycle 7 . . . . . ......... . . . .. .. . .. .... 8-14 I

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I Figures (Cont'd) ,

Figure Page 8-14. Power Imbalance Limits After-200 EFPD, Oconee 1, Cycle 7 . . . . . .. , . . . . . . . . . . . .... 8-15 8-15. APSR Position Limits for 0-50 EFPD, Oconee 1 Cycle 7 . .... 8-16 8-16. APSR Position Limits for 50-200 EFPD, Oconee 1, Cycle 7 . . .. 8-17 8-17. APSR Position Limits After 200 EFPD, Oconee 1, Cycle 7 .... 8-18 I

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1. INTRODUCTION AND

SUMMARY

I This report justifies the operation of the seventh cycle of Oconee Nuclear Station, Unit 1, at the rated core power of 2568 MWt. Included are the re-quired analyses as outlined in the USNRC document " Guidance for Proposed License Amendments Relating to Refueling," June 1975.

To support cycle 7 operation of Oconee 1, this report employs analytical tech-niques and design bases established in reports that have been submitted to and accepted by the USNRC and its predecessor (see references).

A brief summary of cycle 6 and 7 reactor parameters related to power capability is included in section 5 of this report. All of the accidents analyzed in the FSAR have been reviewed for cycle 7 operation.1 In those cases where cycle 7 characteristics were cot.servative compared to those analyzed for previous cy-cles, no new accident analyses were performed.

One fuel assembly from batch 4 will be irradiated for a fifth evde ac part of a joint Duke Power /B&W/ Department of Energy program to demonstrate reliable

fuel performance at extended burnup and to obtain post-irradiation data, in addition, four Mark BZ demonstration fuel assemblies containing Zircaloy-4 in-termediate grids will be inserted in the core as part of the fresh batch 9;

I reference 2 describes the demonstration assemblies. The fifth-burn batch 4 assembly and the four Mark BZ assemblies will not adversely affect cycle 7 operat ion.

The Technical Spec . fications have been reviewed, and the modifications required for cycle 7 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 Cooling Systems, it has been concluded that Oconee Unit 1 can be operated safely for cycle 7 at the rated power level cf 2568 MWt.

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2. OPERATING HISTORY

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The reference fuel cycle for the nuclear and thermal-hydraulic analyses of Oconce 1, cycle 7 is the currently operating cycle 6. The cycle 7 design j length of 427 EFPD is based on a planned cycle 6 length of 383 EFPD. No op-erating anomalies have occurred during previous cycle operations that would adversely affect fuel performance in cycle 7.

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I I 3. GENERAL DESCRIPTION The Oconee Unit 1 reactor core and fuel design basis are described in detail in section 3 of the Final Safety Aaalysis Report for Oconee Nuclear Station, Unit 1.1 The cycle 7 core contains 177 fuel assemblies, each of which is a 15 by 15 array of 208 fuel rods,16 control rod guide tubes, and one incore instrument guide tube. The fuel consists of dished-end, cylindrical pellets W of uranium dioxide clad in cold-worked Zircaloy-4. The fuel assemblies in all batches have an average nominal fuel loading of 463.6 kg of uranium. The undensified nominal active fuel lengths, theoretical densities, fuel 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 1, cycle 7. Sixteen of the batch 7 assemblies will be discharged at the end of cycle 6 along with batch 6 72 . The remaining 40 batch 7 assemblies (designated 7B), one batch 4E assem-bly discharged at the end of cycle 5, batch 8A, and the fresh batch 9 - with initial enrichments of 3.02, 3.20, 2.97, and 3.28 wt % 23s U, respectively -

will be loaded into the central portion of the core. The batch 8B fuel, with an initial enrichment of 3.07 wt % 23s U, will occupy mainly the core periphery.

! Figure 3-2 is an eighth-core map showing the assembly burnup and enrichment distribution at the beginning of cycle 7.

Reactivity is controlled by 61 full-length Ag-In-Cd control rods, 60 burnable lI poison rod assemblies (BPRAs), and soluble boran shim. In addition to the full-length control rods, eight axial power shaping wds (APSRs) are provided for additional control of the axial power distribution. The cycle 7 locations  ;

l of the 69 control rods and the group designations are indicated in Figure 3-3.

The core locations are identical to those of the reference cycle. The cycle 1

7 locations and concentrations of the BPRAs a a shown in Figure 3-4.

The nominal system pressure is 2200 psia and the core average densified nomi-l nal heat rate is 5.80 kW/f t at the rated power of 2568 MWt.

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I Figure 3-1. Core Loading Diagram for Oconee 1, Cycle 7 L3 K2 N5 f

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A 88 88 88 8B 88 g

N3 M2 K8 M14 N13 3 0

88 9 88 9 BB 9 88 9 BB N11 L5 L2

• L14 L11 M4 C

88 9 8A 9 78 9 78 9 BA 9 88 C12 H11 R10 B4 R6 M8 C4 0

BB 9 6A 9 78 9 7B 9 78 9 8A 9 BB O H13 K1 KIS 08 08 E6 E

9 BA 9 BB 9 78 9 78 9 88 9 8A 9 C10 Bil L15 N14 K4 R8 K12 P4 L1 B5 C6 F

88 8B 9 78 9 78 8A 78 BA 78 9 78 9 8B 8B 89 B10 A9 09 C3 CI3 07 A7 B6 B7 88 9 78 9 78 8A 78 9 78 BA 78 9 78 9 BB H W- M12 88 H9 8B 9

• N2 7B 9 H15 78 9 gnr 9 78 H1 9

014 78 9

H7 88 E4 88

_y P9 P10 R9 N9 n3 013 N7 R7 P6 P7 K

88 9 78 9 78 BA 76 9 7B 8A 78 9 78 9 88 010 P11 FIS B12 G4 A8 G12 02 F1 PS 06 L 88 8B 9 78 9 7B BA 78 BA 78 9 78 9 88 8B M10 C8 G1 GIS H3 MS M

9 BA 9 .

8B 9 78 9 7B 9 BB 9 8A 9 012 E8 A10 P12 AS H5 04 N 88 9 8A 9 78 9 78 9 78 9 BA 9 88 g E12 F5 F2 F14 Fil "3 W 0

88 9 BA 9 7B 9 78 9 BA 9 88 03 E2 E14 013 P 80 9 G8 88 E

BB 9 BB 9 9 88 5 F3 G2 011 G14 F13 R

88 88 88 8B BB Z

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

• MARK-BZ DEMONSTRATION ASSEMBLIES XX CY6 LOCATION X BATCH I:

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'I ,1 .,e 3 2. z.,1cs ..< . e .... e1.<<1s.<1.. <. ec..c. 1. cx c1 7

,0 is i4 15 g 8 e 11 12 3.20 3.28 3.02 3.28 3.02 3.28 3.07 3.07 I MARK-8Z H

39949 0 16982 0 13106 0 14440 14700 3.02 2.97 3.02 3.28 3.02 3.28 0.07

'I K 14524 14877 16451 0 21020 0 13869 3.02 3.28 3.02 3.28 3.07 3.07 13101 0 13328 ' O 10801 14790 3.07 3.28 2.97 3.28 M

15232 0 14616 0 2.97 3.28 3.07 I N 14536 0 11864 3.07 14714 8

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I X.XX INITlAL ENRICHMENT XXXXX 800 BURNUP, MWO/MTU l

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I Figure 3-3. Control Rod Locations for Oconee 1, Cycle 7 I I

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C 1 6 6 1 0 7 8 4 8 7 E

F 3

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2 7

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6 2 4 4 2 6 l H W 7 4 5 3 5 4 7 Y

! K B 2 4 4 2 6 L 3 8 7 5 7 8 3 l

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l N 7 8 4 8 1 0 1 6 6 1 P l l l 3 7 3 R IIII Z

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

l GROUP NO. OF RODS FUNCTION 1 8 SAFETY X GROUP NUMBER 3 y 9 S 4 8 SAFETY E 5 8 CONTROL E l 6 8 CONTROL 7 12 CONTROL g.

8 8 APSRs W TOTAL 69 3-4 Babcock & Wilcox E

I I Figure 3-4. BPRA Concentration and Distribution for Oconee 1, Cycle 7 e 0 i0 ,, >2 is i4 is y i.00 i.00 i.00 I "

i.20 0.20 g ,

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X.XX BPRA CONCENTRATION, WT. %48 C IN Al 023 I

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[ 4. FUEL SYSTEM DESIGN 4.1. Fuel Assembly Mechanical Design The types of fuel assemblies and pertinent fuel design parameters for Oconee 1 cycle 7 are listed in Table 4-1. All the fuel assemblies are identical in con-cept and are mechanically interchangeable. Four Mark BZ demonstration fuel as-semblies are included in batch 9B. The Mark BZ is a 15 by 15 fuel assembly similar to the Mark B assembly described in reference 1, except that the six intermediate apacer grids will be of Zircaloy and a slightly redesigned hold-down spring is incorporated. The Mark BZ assembly is described in reference 2, which also states that reactor safety and performance are not adversely af-fected by the presence of the four demonstration assemblies.

Batch 4E contains one high burnup Mark B fuel assembly. This assembly wil] be in its fifth burnup cycle during cycle 7.

Retainer assemblies will be used en the two batch 8 fuel assemblies that con-tain regenerative neutron source (RNS) assemblies and on the 60 assenblies that contain BPRAs. Some of the retainers will be reinserted on fresh assemblies for a second cycle. The justification for the design and use of the retainers is described in references 3 and 4.

4.2. Fuel Rud Design The mechnical evaluation of the fuel rod is discussed below.

4.2.1. Cladding Collapse The fuel assembly af batch 4E is more limiting than those of other batches be--

cause of its longer previous incora exposure time. The batch 4E assembly power history was analyzed and used to perform the creep collapse analysis usf r.g the I CROV computer code and procedures described in reference 5. The collapse time for the batch 4E assembly was conservatively determined to be more than 40,000 effective full-power hours (EFPH), which is greater than the maximum projected residence time of cycle 7 fuel (Table 4-1); of the remaining fuel, batch 7 is 4-1 Babcock & Wilcox

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the most limiting. The power history and 'uel design parameters for the most limiting batch 7 fuel assembly were compared with thc,se used in the Mark B generic creep collapse analysis and were found to be enveloped. Thus, the collapse time for the most limiting batch 7 fuel assembly is conservative-ly determined to be more than 35,000 EFPH, which is longer than the maximum projected three-cycle exposure time of 27,202 EFPH (Table 4-1).

4.2.2. Cladding Stress The Oconee 1, cycle 7 stress parameters are enveloped by a conservative fuel rod stress analysis. No new methods were used for the analysis of this cycle other than that used on previous cycles.

4. 2. 3. Cladding Strain The fuel design criteria specify that the cladding average circumferential strain is not to exceed 1.0% inelastic strain. The pellet design is estab-lished for plastic cladding strain of less than 1% at maximum design local pellet burnup (55,000 mwd /mtU) and heat generation rate (20.15 kW/ft) values that are higher than the values the Oconee 1 fuel is expected to see. However, the batch 4E fuel assembly will experience a local pellet burnup greater than 55,0C0 mwd /mtU. An analysis was performed to show that the maximum cladding circumferential uniform strain is less than 1% up to a maximum local pellet

'.,urnup at 74,000 mwd /mtU, which is much higher than the batch 4E fuel assem-E T

bly will attain. This analysis assumes an unrealistically high transient t induce cladding deformation. Thus, fuel rod cladding strain will not affect E

5 cycle 7 fuel performance.

4. 3. Thermal Design All fuel assemblies in this core are thermally similar. The fresh batch 9 fuel inserted for cycle 7 operation introduces no significant dif ferences in fuel thermal performance relative to the other fuel remaining in the core.

The design minimuni linear heat rate (LHR) capacf ty and the average fuel tem-i perature for each batch in cycle 7 are shown in Table 4-2.

f The batch 4E assembly reinserted for a fif th cycle of exposure was analyzed 5

with TACO 2s, which predicted the fuel rod internal pressure to be less than the nominal reactor coolant system pressure of 2200 psfa.

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l l l The maximum fuel rod burnup at EOC7, excluding the high burnup assembly, is predicted to be 34,709 mwd /mtU. Fuel rod internal pressure nas been evaluated 7

with TAFY-3 for the highest burnup fuel rod and is also predicted to be less than 2200 psia.

4.4. Material Design l

The batch 9 fuel assemblies are not new in concept, nor do they utilize dif-ferent component materials, except for the Zircaloy. grids of the four Mark-BZ assemblies described in section 4.1. Therefore, the chemical compatibility of all possible fuel-cladding-coolant-assembly interactions for the batch 9 fuel assemblies are identical to those of the present fuel.

4.5. Operating Experience Babcock 6 Wilcox operating experience with the Mark B 15 by 15 fuel assembly has verified the adequacy of its design. As of July 31, 1980, the following experience has been accumulated for the eight operating B&W 177-f uel assem-bly plants using the Mark B fuel assembly:

Max assembly Cumulative net Current electrical output, Reactor cycle Incore Discharged MWh Oconee 1 6 22,100 40,000 31,675,436 Oconee 2 5 24,532 33,700 26,888,897 Oconee 3 5 31,100 29,400 27,394,492 TMI-1 4 32,400 32,200 28,840,053 I ANO-1 Rancho Seco 4

4 26,900 27,900 33,222 37,730 24,430,676 21,858,687 Crystal River 3 2 23,194 23,194 11,400,975 Davis Besse 1 1 14,884 -

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8 Table 4-1. Fuel Design Parameters and Dimensions ,

Four-times Twice- Once- Fresh B burned FA, burned FAs, burned FAs, FAs, 5 batch 4E batch 7B batches 8A/8B batch 9 FA type Mark B3 Mark B4 Mark B4 Mark B4/ l Mark BZ

• No. of FAs 1 40 20/48 64/4 g Fuel rod OD, in. 0.430 0.430 0.430 0.430 W Fuel rod ID, in. O.377 0.377 0.377 0.377 Flex spacers, type Spring Spring Sp ring Spring Rigid spacers, type Zr-4 Zr-4 Zr-4 Zr-4 ,

Undensified active fuel length (nom), in. 142.0 142.25 142.25/141.38 141.8 =

Fuel pellet initial >94.5 94 94/95 95 density (nom), % TD Fuel pellet OD (mean 0.3685 0.3695 0.3695/0.3686 0.3586 specification), in.

Initial fuel enrich- 3.20 3.02 2.97/3.07 3.28 5 ment , wt % 23sU BOC burnup (avg), 39,949 15,931 i4,698/13,480 0 mwd /mtU Cladding collt.pse >40,000 > 35,000 >35,000 >35,000 .

time, EFPH Estimated residence 38,100 27,202 30,024 30,696 time (max), EFPH ,

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Table 4-2. Fuel Thermal Analysis Parameters - Oconee 1, Cycle 7 Batch No_.

4E 7B 8A 8B 9(*)

No. of assemblies 1 40 20 48 68 Nominal pellet density, % TD 95.5 94.0 94.0 95.0 95.0 Pellet diameter, in. 0.3685 0.3645 0.3695 0.3686 0.3686 I Stack height, in. 141.0(N 142.25 142.25 141.38 141.8 Densified Pellet Parameters "

Pellet diameter, in. 0.3640 0.3646 0.3646 0.3649 0.3649 Fuel stack height, in. 140.30 140.47 140.47 140.32 140.74 Nominal LHR @ 2368 MWt, kW/ft 5.8 5.8 5.8 5.8 5.8 Avg fuel temp. at nominal LHR, F 1320 1320 1320 1320 1310 LHR to (, f uel melt, kW/f t 20.15 20.15 20.05 20.15 20.15 Core uverage densified LHR at 2568 MWt = 5.8 kW/ft I

" Includes tour Mark BZ demonstration assemblies.

Conservative calculational parameter.

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Densification to 96.5% TD assumed.

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5. NUCLEAR DESIGN 5.1. Physics Characteristics Table 5-1 compares the core physics parameters of design cycle 7 with those of the reference cycle 6. The values for both cycles were genereted using PDQ07. 8..l o The average cycle burnup will be higher in cyc'e 7 than in the design cycle 6 because of the longer cycle 7 length. Figure 5-1 illustrates a representative relative power distribution for the beginning of cycle 7 at full pcwer with equilibrium xenon and normal rod positions.

Since the core has not yet reached an equilibrium cycle, differences in core physics parameters are to be expected between the cycles. The critical boron concertrations for cycle 7 are higher because the additional reactivity neces-sary for the longer cycle is not completely offset by the burnable poison. The I control rod worths differ between cycles due to changes in radial flux and burnup distributions. This also accounts for the larger ejected and stuck rod worths in cycle 7 compared to cycle 6 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 position limits presented in section 8. All safety criteria associated with these rod worths are met. The adequacy of the shutdown margin with cycle 7 stuck rod worths is demonstrated l

in Table 5-2. The following conservatisms were applied for the shutdown cal-culations:

1. Poison material depletion allowance.

2. 10% uncertainty on net :-d worth.

3. Flux redistributioa penalty.

Flux redistribution was accounted for since che shutdown analysis was calcu-lated using a two-dimensional model. The reference fuel cycle shutdown margin is presented in the reload report for Oconee 1, cycle 6.11 I

The cycle 7 power deficits, differential boron vocths, and ef fectiva delayed neutron fractions differ from those for cycle 6 because of the longer cycle leagth and higher critical boron concentrations.

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1 Il l 5.2. Analytical Input Ii i The constants used to compute core power distributions from incore detector mensurements were obtained in the same manner for cycle 7 as for the reference 3l cycle 6.

5.3. Changes in Nticlear Design Il I There are no significant changes in core design between the reference and re-load cycles. The calculational 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-Table 5-1. Oconee 1 Physics Parameters

e-Cycle 6 Cycle 7(#>

f Cycle lenrth, EFPD 372 427 l Cycle burnup, mwd /mtU 11,641 13,363 Average core burnup, EOC, mwd /mtU 19,927 22,505 Initial are loading, mtU 82.1 82.1 i

l Critical boron, BOC (no Xe), ppm

!, HZP, group 8 inserted 1448 162P l HFP, group 8 inserted 1253 1464 l Criticel boron, EOC (eq Xe), ppm

! H2P, group 8 inserted 377 380 HFP, proup 8 inserted 83 68 f

{ Control rod worths, HFP, BOC, % Ak/k j Group 6 1.04 0.97 l Croup 7 1.51 1.45

Group 8 0.50 0.47

! Control rod worths, HFP, EOC, % Ak/k Group 7 1.60 1.54 Group 8 0.54 0.53

{I Max ejected rod worth, HZP, % Ak/k(*

EOC (N-12)

, 0.42 0.55 d

EOC (N-12) 0.44 0.62 I Max stuck rod worth, HZP, % Ak/4

, BOC (N-12) 1.24 1.44

{ EOC (N-12) 1.44 1.65 f Power dcficit, HZP to H9 % Ak/k j BOC 1.56 1.35 EOC 2.48 2.25 Doppler, coeff, 10~5 (Ak/k 'F)

BOC, 100% power, no Xe -1.48 -1.52

! EOC, 100% power, eq Xa -1.61 -1.62 f

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

) BOC (0 Xe, crit ppa, gp 8 ins) -0.72 -0.48 ,

i EOC (eq Xe, 17 ppm, gp 8 ins) -2.85 -2.87 l Boron worth, HFP, ppm /% Ak/k i Boc (1150 ppm) 116 123 EOC (17 ppm) 102 105 i

I Xenon worth, HFP, % Ak/k f BOC (4 EFPD) 2.60 2.58 l EOC (equilibrium) 2.73 2.74 Eff delayed neutron fraction. HFP BOC 0.00612 0.00626 EOC 0.00516 0.00520 I (a) Cycle 7 data are for the conditions stated in this report.

The cycle 6 core conditions are identified in reference ed on 320 EFFD at 2568 MWt, cycle 5.

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Cycle 7 data are based on a cycle 6 length of 383 EFPD.

HZP denotes hot zero power (532F Tavg), HFP deno*.es hot I *ull power (579F Tavg)-

• Ejected rod worth for groups 5 throug' 8 inserted.

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Table 5-2. Shutdown Margin Calculation for Oconee 1, Cycle 7 l l

BOC, % Ak/k EOC, % Ak/k Available Rod Worth Total rod worth, HZP 8.68 9.17 Worth reducti m due to burnup of poison material -0,42 -0.42  ;

Maximum stuck rod, HZP -1.44 -1.65 ,

Net worth 6.82 7.10 Less 10% uncertainty 0.68 0.71 Total available worth 6.14 6.39 Recuired Rod Worth Pcwer deticit, HFP to HZP 1.35 2.25 Ma:,: allowable inserted rod worth 0.30 0.55 ,

Flux redistribution 0.60 1.19 Total required worth 2.25 3.99 '

Shutdown margin (total available worth minus tctal rr,uired worth) 3.89 2.40 t

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

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5 Figure 5-1. Oconee 1, Cycle 7 BOC (4 EFPD) Two-Dimensicaal Relative Power Distribution - Full Power, Equilibrium Xenon, Normal Rod Positions

8 9 10 11 12 13 14 e5 0.780 1.182 1.077 1 J5 1.217 1.233 0.968 0. 50 4 MARK-BZ l.082 1.134 1.137 1.266 1.023 1.075 0.499

I L 1.178 1.290 1.0 1.226 0.910 0.387

I M 1.239 1.251 1.086 0.896 il N 1.111 1.007 0.486

!I 0 0.527 I

P il R

I 1

18 INSERTED R00 GROUP NO.

X.XXX RELATIVE POWER DENSITY I

.I 5-5 Babcock & Wilcox

.II

-g9--- ----,,--ap-,- y- y--w, -,c. -- -.y--r-+- -e-,- -w

8 l I

I

6. THERMAL-HYDRAULIC DESIGN The incoming batch 9 fuel is hydraulically and geometrically similar to the fuel remaining in the core from previous cycles. The thermal-hydraulic design evaluation supporting cycle 7 operation utilized the methods and models de-scribed in references 1, 11, and 12 except for the insertion of four hark BZ demonstration assemblies which contain six Zircaloy intermediate spacer grids.2 The cycle 7 maximu:r. design conditions remain unchanged from cycle 6 and are shown in Table 6-1. The four Marx BZ demonstration assemblies will be limited to a design peak of 1.558 (10% peaking reduction) to ensure that they are not thermally limiting, wh* e the 1,71 design radial-local peak remains valid for I all other assemblies la this cy-le. The thermal-hydraulic analysis did not consider the insertion of the Mark BZ assemblies. The analysis is conserva-tive because the Mark BZ assemblies have a higher resistance to flow and would thus tend to increase the flow in the limiting assembly.

A rod bow penalty has been calculated according to the procedure approved in reference 13. The burnup used is the maximum fuel assembly burnup of the batch that contains the limiting (maximum radial-local peak) fuel assembly. For cycle 7, this burnup is 17,649 mwd /mtU in a batch 9 assembly. The resultant net rod bow penalty af ter inclusion of the 1% flow area reduction factor credit is 0.2% reduction in DNBR. The thermal-hydrsulic design for cycle 7 includes a trargin more than 0.2% above the minimum DNBR of 1.30.

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6-1 Babcock s.Wdcox

l 8

Table 6-1. Thermal-Hydraulic P sign Conditions l

Cycle 6 11 Cycle 7 Power level, MWt 2568 2568 l System pressure, psia 2200 2200

( Reactor coolant flow, % design flow 106.5 106.5 Vessel inlet coolant temp, 100% power, F 555.6 555.6 Vescel outlet coolant temp, 100% power, F 602.4 602.4

( Ref design axial flux shap 1.5 cos 1.5 cos i

Ref design radial-local power peaking factor 1.71 1.71 Active fuel length, in. (a) (a) 3 2 Average heat flux, 100% power, 10 Btu /h-ft 176(b) 176(b)

CHF correlation BAW-2 BAW-2 Hot channel factors l Enthalpy rise 1.011 1.011 W Heat flux 1.014 1.014 ,

Flow area 0.98 0.93 '

Minimum DNBR with densification penalty 2.05 2.05

" See Table 4-2.

i (

Based on densified length of 140.3 in.

(

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6-2 Babcock & Wilcu  !

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7. ACCIDENT AND TRANSIENT ANALYSIS 7.1. General Safety Analysis Each FSAR I accident analysis has been examined with respect to changes in cycle 7 parameters to determine the effect of the cycle 7 reload and to ensure that thermal performance during hypothetical transients is not degraded.

I The ef fects of fuel densification on the FSAR accident results have been eval-uated and are reported in BAW-1388.12 Since batch 9 reload fuel assemblies contain fuel rods whose theoretical density is higher than those considered in the reference 12 report, the conclusions in that reference are still valid.

J.2. Accident Evaluation The key parameters that have the greatest effect on determining the outcome of I a transient can typically be classified in three major areas: core thermal pa-rameters, thermal-hydraulic parameters, and kiaetics parameters, including the reactivity feedback coefficients and control rod worths.

Core thermal properties used in the FSAR accident analysis were design operat -

ing values based on calculational values plus uncertainties. Fuel thermal analysis valuea for each batch in cycle 7 are compared in Table 4-2. The cycle 7 thermal-hydraulic maximum design conditions are compared to the previous cycle 6 values in Table 6-1.11 These parneters are common to all the accidents considered in this report. The key kinetics parameters from the FSAR and cycle 7 are compared in Table 7-1.

A generic LOCA analysia for the B&W 177-FA, lowered-loop NSS has been performed using the Final Acceptance Criteria ECCS Evaluation Model. This study is re-ported in BAW-10103, Rev 1.1" I The analysis in BAW-10103 is generic since the limiting values of key parameters for all plants in this category were 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 5

7-1 Babcock & Wilcox

I is conservative compared to those calculated for this r21oad. Thus, the anal-ysis and the LOCA limits reported in BAW-10103 provide conservative results for the operation of Oconee 1, cycle 7 fuel.

Table 7-2 shows the bounding vs. lues for allowable LOCA peak LHRs for Oconee 1 cycle 7 fuel after 50 EFPD. The LOCA kW/ft limits have been reduced for the first 50 EFPDs. 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 analysis l5 in ordm to account for mechanistic fuel densification. The limits for the first 50 EFPD are shown cre shown in Table 7-3.

It is concluded from the examination of cycle 7 core thermal and kinetics prop-erties, with respect to acceptable previous cycle values, that this core re-load will not adversely affect the ability of the Oconce 1 plant to operate g safely during cycle 7. Considering the previously accepted design basis used =

in the FEAR and subsequent cycles, the transient evaluation of cycle 7 is con-sidered to be bounded by previously accep.ed analyses. The initial conditions a for the transients in cycle 7 are bounded by the FSAR , the fuel densification 1

report 12, 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 report because even though the FCAR dose analyses used a conservative basis for the amount of plutonium fissioning in the core,, improvements in fuel management techaiques have increased the l

amount of energy produced by fissioning plutonium. Since plutonium-239 has different fission yields than uranium-235, the mixture of fission product nu-clides in the core <hanges slightly as the plutonium-239 to uranium-235 fission ratio changes, i.e., plutonium fissions produce mcre or some nuclides and less of others. Since the rad". alogical doses associated with each accident are im- W pacted to a different entent by each nuclide and by various mitigating factors .

and plant design features, the radiological consequences of the FSAR accidents  ;

were recalculated using the specific parameters applicable to cycle 7. The bases used in the dose calculations are identical to those presented in the FSAR except for the following two notable differences:

1. The fission yields and half-lives used in the new calcu-

~

lations are based on more current data.

I 7-2 Babcock 8.Wilcox

,- w '

}.,-

.h,'

1. }4'%i l MAGE EVALUATION f

TEST TARGET (MT-3) l.0 g a su EEE usir a 3

/ Ma i.25 ug

/ j 6" 4 $

+4+

• k ' h//

h,,,

4},i}.[O J

- _ _ ---_ x _ _ _ _ _ _ _ - _____ -

g . . _ . . _ . . _ ,

h/

g.; .,, ,

$<>a%, ,

h.:ty i NMI IMAGE EVALUATION .

TEST TARGET (MT-3) l.0 Ein m ,

== am g"n <

NJ l'l 1,1 'x.,[," EM +

1I / g '

l.25 " i.4 g

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4 sv +i, : _a

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\ IMAGE EVALUATION TEST TARGET (MT-3) e 1.0 l# tu E4 s M E g 2.2 .

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, I.lS;1'BE 3 l.25 1.4 I

_ 'lx.6 p4 <$ 4

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2. The steam generator tube rupture accident evt tuation con-siders the increased amount of steam released to the en-vironment via the main steam relief valves because of the slower depressurization due to the reduced 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 FSAR and those calcu-lated specifically for cycle 7. As can be seen from Table 7-4, some doses are slightly higher and some are slightly lower than the FSAR values; however, all I doses are well below the 10 CFR 100 limits of 300 Rem to the thyroid and 25 Rem to the whole body. The small increases in some doses are essentially off-set by reductions in other doses; in any event, these increases are within the range of untertainty normally associated with dose calculations. Thus, the radiological impacts of accidents during cycle 7 are not significantly differ-ent than those described in Chapter 14 of the FSAR, Cycle 7 contains a one-fuel assembly batch (4E) that will be irradiated to s50,000 mwd /mtU as part of the DOE extended burnup program. The doses asso-ciated with a fuel handling accident for this fuel assembly were calculated.

The results show that the thyroid and whole body doses are 0.28 and 0.013 Rem at the 2-hour exclusion distance. Both of these values are bounded by the fuel handling accident dose shown in Table 7-4. Therefore, operation of cycle 7 with this high burnup assembly poses no additional risk to the public.

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7-3 Babcock & Wilcox

8, Tabic 7-1. Comparison of Key Parameters for Accident Analysis FSAR and Predicted '

densification cycle 7 Parameter report value value Doppler coeff, 10-5 Ak/k/*F f BOC -1.17 -1.52 ,

EOC -1.33 -1.62 Moderator coeff, 10~" Ak/k/*F  ;

BOC +0.5 -0.48 ,

EOC -3.0 -2.87  ;

All-rod group worth at HZP,

% Ak/k 10 8.68 l Initial boron cone'n at HFP, ppm 1400 1464  !

isoron reactivity worth at 70F, W, ppm /1% Ak/k 75 86 Max ejected rod worth at HFP,

% Ak/k 0.65 0.28 l t

Dropped rod worth (HFP), l

% Ak/k 0.46 0.20 l r

Table 7-2. LOCA Limits, Oconee 1, Cycle 7, I; After 50 EFPD l Elevation, LHR limits, r ft kW/ft [

2 15.5 i 4 16.6 6 18.0 8 17.0 i 10 16.0 B

I! .

1 I! l 7-4 Babcock & Wilcox  !

8 Table 7-3. LOCA Limits, Oconee 1, Cycle 7, j 0-50 EFPD j

W Elevation, LHR limits, i ft kW/ft 2 14.5 i

4 16.1 l

6 17.5 1 8 17.0 1

10 16.0 1

!I il 4

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1 il lI 4

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I 7-5 Babcock & Wilcox I

1 l l l 8l Table 7-4. Comparison of FSAR and Cycle 7 Accident Doses l FSAR doses, Cycle 7 doses. l Accident Rem Rem W '

Steam generator tube f ailure lI g'

Thyroid at EAB 0.00034 0.31 Whole body at EAB 0.023 0.058  :

Steam line break Thyroid at EAB 0.19 0.20 Whole body at EAB 0.002 0.0016 l Fuel handling accident Thyroid at EAB 0.43 0.50  !

Whole body at EAB 0.027 0.028 i Rod ejection accident  !

Thyroid at EAB 0.19 0.21 gl Whole body at EAB 0.001 0.0005 5i '

Thyroid at LPZ 0.22 0.23 l

Whole body at LPZ (a) 0.0007 j Waste gas tank Thyroid at EAB 0.13 0.29 I Whole body at EAB 0.19 0.18 LOCA  !

Thyroid at EAB 4.6 4.9 l Whole body at EAB 0.01 0.010  !

Thyroid at LPZ 5.0 5.5 Whole body at LPZ 0.014 0.014 MHA ,

Thyroid at EAB 186. 193. gf Whole body at EAB 1.4 1.4 gj Thyroid at LPZ 144. 180.

Whole body at LPZ 0.65 0.62

(")Not reported in FSAR.

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7-6 Babcock & Wilcox

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8. PROPOSED MODIFICATIONS TO TECHNICAL SPECIFICATIONS The Technical Specifications have been revised for cycle 7 operation in accor-dance with the methods of references 16-18 to account for changes in power peaking and control rod worths. In addition,

1. The 0-50 EFPD operating limits on rod index, APSR position, I and axial power imbalance were established based on the in-terim LOCA linear heat rate limits, which account for mech-anistic fuci densification.15 After 50 EFPD, the FAC LOCA LHR limits " were used.

I I 2. The refueling boron requirement for cycle 7 has been in-creased from 1600 to 1835 ppmB as a esult of the increased

'I reactivity associated with the longer cycle. The Technical Specifications for the concentrated boric acid storage tank, core flood tanks, and borated water storage tank have been revised accordingly.

Based on the Teclinical Specifications derived f rom the analyses presented 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 8-17 are revisions to previous Technical Specification limits.

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I 8-1 Babcock & Wilcox

I cigure 8-1. Core Protection Safety Limits for Oconee Unit 1 THERNAL POWER LEVEL, %

- - 120 l

(-28,112) (28,112) 1 ACCEPTABLE 4 PUMP

(-45,100) OPERATION -

- 100 ( +45,100 )

g'

(

(-28,87.2) ' '

2 l ACCEPTABLE -

80 314 PUNP

(-45,74) OPERATION ( +45,74)  !

I

(-28,59.4) 3 _ _ 60 -

(2d,59.4) I! [

i ACCEPTABLE f' 2,3 & 4 PUNP l (-45,48) (+45,48)

OPERATION  ;

- 40 l l

I:!

- 20 l

I!!

I I I I I l l

-60 -40 -20 0 20 40 60 I Reactor Power imoalance, %

CURVE RC FLOW (GPM)  ;

I 374,880 2 280,035 ,

3 183,690 i

8-2 Babcock & Wilcox l i

8 Figure 8-2. Protective System Maximum Allowable Setpoints for Oconee Unit 1 THERMAL POWER LEVEL, %

- 120 (X,Y)

(X,Y)

(-15,108) -

- 110 (17,108)

M, = -0.90 g 4 PUNP M 2

1.0 l

OPERATION - 100 I I

(-36,89) -

- 90 (36,89) g

(-15,80.7) l(17,80.7) l 3&4 PUMP 80 j' l OPERATION l

- - 7,

(-36,62) l l (36,62)

[

~ 60 i l l ( 17,52.9)

I (-15,52.9)

~ ~

y2 ,3 & 4 I PUMP

! 0PERATION.

l

- 40 l

l l

(-36,34) (36,34) g I -

- 30 I

l

- 20 ,

.

e, i I +e +

I n

=

~

nl

1. l

- 10 gn l

n I i

-40 i i

-20 i i i 0

i Ii 20 i

40 i i Reactor P0wer Imnalance, %

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8-3 Babcock & Wilcox I

f

Figure 8-3. Rod osition Limits for Four-Pump Operation, 0-50 EFPD, Oconee 1, Cycle 7 00,102) 100

)f (281,92) -

POWFA LEVEL O 80 - (271,80). CUT 0FF "E

OPERATION RESTRICTED = 100% FP a.

. E E

5 60 -

SHUTOOWN MARGIN 3 LIMIT p E (88,50) . (200.50) .

• E E OPERATION 40

- NOT ALLOWE0 OPERATION ACCEPTABLE b

aE E 20 -

(0,11.8)(

5 a- (0.2 5)0 c i , ,

200 300 8 0 100 E- GR 5 I i i Rod Inaex (witndrawn)

" 75 100 0

$ GR 6 ' ' ' ,J 00' E O 25 7p  ;

S GR 7 0 25 100

___ E __ E .__ E ._-E __. . . _ . _

E _ ___ _ - __ _ _ _ _ ._. . . . . _ _ _ _. _ _ _ _ _ . .

l Figure 8-4. Rod Position Li.mits for Four-Pump Operation, 50-200 EFPD, Oconee I, Cycle 7

( 44.5,102 . a ( 300,102) 100 -

(277.5.102) y (274.5,92) )

JWER m

/ EVEL

80 - (264.5,80) . ;UT0FF = 1007, FP g OPERATION g RESTRICTED e SHUTDOWN g MARGIN L ittl T

$ 60 -

B E

3 _

(88,50) . ( 200. 50) .

' o OPERATION

" OPERATION ACCEPTABLE E NOT ALLOWED 5 40 -

b E

o 20 -

(0,11.8)c l g (0,2.5) c,  ! I i er 0 8

a 0 100 200 300

[ GR 5 i , i Roa Index (witnarann) g 0 GR 6 ' Y ' t 0 25 7,5 10p ,

GR 7 0 25 100 l

Figure 8-5. Rod Position Limits for Four-Pump Operation Af ter l 200 EFPD, Oconee 1, Cycle 7

( 0.102) (274.5.102) a (300.102) 100 -

(271,92)

POWER LEVEL (264.5.80) CUT 0FF = 100*, FP C 80 - OPERATION NOT ALLOWED E

Q OPERATION E RESTRICTED 5 60 -

E E (160,50) ( 200,50) 4  ; SHUTDOWN MARGIN E 40 -

g LIMIT OPERATION ACCEPTABLE O

b E

f 20 -

( '

(90,15)

(0,8.5)c p (0,2.5]C , , ,

{

E- GR 5 I 0

i 100 200 Rod inaer (witnarann) 300

" 0 75 100 GR 6 ' I I I

E R 0 25 1p0

}5--

0 0 25 100 M . . _ .

M . _ _

M . - -

M _.

M M M M M M M N

\il I' Ill l lIl i!

M m

M

)

7 7

~M 0

0 3 E

( L 0 0 B 0 0 A l 3 T i

3 1 P

M E C

C A

N O

M D P

F -

I T

A E ) R 7 E 0

5 7

P O

)

n ,

M 0 4

6 2

a r

w

( a

) n n 0 t 0

p 5 o 5 i i 1 2 i w M

t 0 (

a 0 r 2 x e D

( e p 0 d E n O T *  ; 0 2 I M p m

C I

R d

o u T R P S

- ) E 5 e 7 R i

] 0 M l e 7 5, N O

I T 4. T 7 4 A r 1 R R o ( E G M f s

P O

t i7 m 0 i e p 5 M

I Ll 1 2 c )

ny 5 oC 1 i ,

0 t ,

0 0 i1 M s oe 0

E 9

(

1 P e W

• n O - .

d o L L )

oc A N 8 5 RO M T O

N 0N 0I T 3

8 f

7, 0 N TGI 8 6 URM ( ,

- N HA 8 O SMLI M e r

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A 6

R u R G g E P )

i O 5 F 7 M 1 1

7 3

(

M - - - - - F 0

( c i 0

)

0 0 0 0 0 5 )0 0 8 6 4 2 5 5 1

3 9

2. R G

M 0%f E5g oSE ; gSf t E 0

(

0

(

M 2L o t$$7 g8x M

i ll l,ll ll

~

Figure 8-7. Rod Position Limits for Three-Pump operation, 50-200 EFPD, Oconee 1, Cycle 7 100 -

~

(144.5.77) (258,77) -

3 (300,77)

E j j OPERATION NOT g" ALLOWE0 OPERATION RESTRICTED

' ~ 60 -

i o SliUi00WN i

0 MARGIN j E LIMIT - (200,50) z o I w i,

$4 O

(86,36) .

OPERATION ACCEPTABLE t

20 -

(37,11.75) -

(90,15)

(0.9.35)c r (0,2. 5) c : , t i 0

@ 0 100 200 300 E

o ' Rod index (witndrann)

GR 5 1 I E- 0 75 100 , i e GR 6 75 100

$ 25 GR 7 ' '  ;

F 0 25 100 O

Figure 8-8. Rod Position Limits for Three-Pump Operation After 200 ZFPD, Oconee 1, Cycle 7 100 l-80 -

(220,77) (258,77) (300,77) c . .

3 E OPERATION o _

NOT ALLOWED, nPER.

$ 60 - REST 0 SHUTOOWN MARGIN (200,50) a LIMIT

?

• t 40 -

(160,38) -

g OPERATION ACCEPTABLE 5

y ~

(133,29) 0 -

OPERATl0N RESTRICTED

- (82,11.75)

(0.6.875)C (54,10)

(0,2.5) c , , ,

p 0 100 200 300 h GR 5 I t I Rau index (witnarawn) o 0 75 100

" GR 6 I I I 0 25 75 100

$ GR 7 i ' l is 0 25 100 0

1

rigure 8-9. Rod Position Limits for Two-Pump Operation, 0-50 EFPD, Oconee 1, Cycle 7 100 -

O 80 -

f E

E 60 -

OPERATION f

NOT ALLOWED (144.5,52) .

(205,52)

D

? E PERATION 000,50)

SHUTDOWN E $ 40 -

MARGIN LIMI RESTRICTED E

5 h

OPERATION ACCEPTABLE E 20 -

(37,8,5;

~

(0.6.9) W (0,2.5)0 I ' '

0 100 200 300

[

E GR 5 I i I Rod inder (witnarawn)

E- 0 7p Iq0 , ,

GR 6 h 0 25 75 100 5 0 25 100 a

e e m e es em em mm m W W WW m WW W _

EE m _

EE _

W

Figure 8-10. Rod Position Limits for Two-Pump Operation, 50-200 EFPD, Oconee 1. Cycle 7 100 -

n U 80 -

I E

n.

I G

j OPERATION 60 - NOT ALLOWE0

[

g (i44.5,52) (204,52) (300,52) o m

(200, 50)

SHUT 00NN PERATION C E 40 - MARGIN RESTRICTE0 E LIMIT OPERATION E (88,26) .

y ACCEPTABLE 20 -

- (90,15)

(37,8.5)

• I (0.6.9) e r (0,2.5) C i i i j p 0 0 100 200 300 l [ Rou Index (witndrawn)

GR 5 t I I

! E O 75 100 i e GR 6 I I

g 0 25 GR 7 ]5 1p0 l l

x 0 25 100 i

i 1

Figure 8-11. Rod Position Limits for Two-Pump Operation After 200 EFPD, Oconee 1, Cycle 7 100 -

O 80 -

E E

EE

% OPERATION NOT ALLOWE0 5 60 -

g (220,52) x 3 (300,52) 2  ; SHUT 00WN

,L _ MARGIN g 40 -

LIMIT S

f U -

(160,26) OPERATION E

' 20 - ACCEPTABLE OPERATION RESTRICTED i

( , 5 8,6.5)  ; i f 0

h 0 100 200 300 g gg 3 , , , Roa Inan N tnhn) 7 0 GR 6 ' '

E O 25 75 10,0 ,

[

0 25 100 M

n 4

Figure 8-12. Power imbalance Limits for 0-50 EFPD, 1

Oconee 1, Cycle 7 OPERATION RESTRICTED i

(-14,102) o (17.5,102) 100 - -

(-15,92) (17.5,92) 5 O jg (-25,80) 80 - -g (20,80)

!I i "c

. ~

lE

>g OPERATION 60 -

o.

ACCEPTABLE  %

cc it  :

E 40 - -E

=

i

' cL.

4

~

\  ;

ih "

5 i 20 - -

!I i.

} l  ! I f I

-30 -20 -10 0 10 20 f Axial Power imoalance (%)

i i

l l

lI l

l 8-13 Babcock & Wilcox il 1

Il l Figure 8-13. Power Imbalance Limits for 50-200 EFPD, Oconee 1, Cycle 7 Il l OPERATION RESTRICTED

( .17. 5.102) -

100 - -

3 (17.5,102) I

(-10,92) - _ _

(17.5,92)

(-30.5,80) . 80 - -y - ( 20,80 ) I,l a

- - Ee t '

E OPERATION 60 - -

o. -

ACCEPTA8LE  % i a:

-- g Z

40 - - '

s m I J'

20 - - g-i i i i i

-30 - 29 -10 0 10 20 1

Axial Power imoalance (%)

I I

8-14 Babcock & Wilcox

I Figure 8-14. Power Imbalance Limits After 200 EFPD, Oconee 1, Cycle 7 1

OPERAil0N RESTRICTED I

-I (-23.5,102) e . (15.5.102)

100 - -

(-24.5,92).

. (17,92)

I (-31,80) -

80 - -

c g - (20,80) 2 j

i  :

'E 5 IE 60 -- o

=

)I OPERATION _ _

ACCEPTABLE i Z 40 - -

I b

i E E

20 - -

l I l I f I i

-30 -20 -10 0 10 20 Axial Power imoalance (%)

i J

8-15 Babcock & Wilcox

I Figure 8-15. APSR Position Limits for 0-50 EFPD, Oconee 1, Cycle 7 (8.5,102) (35,102)

G 3 100 -

(35,92) OPERATION I

- (8. 5.92) -

RESTRICTED 80 C (0,80) .

(39.5,80) g O g d'

ae ,

a 60 -

.li U OPERATION ACCEPTABl.E (100,50)  !

a:

40 C

i I: t U

g E.

=

20 -

5l:I I I i  ; i O

0 20 40 60 80 100 f i

APSR Position (P'ercent Witnarawn)

T I

i i

f i

k 8-16 Babcock & Wilcox  !

i

I l Figure 8-16. APSR Position Limits for 50-200 EFPD, Oconee 1, Cycle 7 4

i (B.5,102) (32.5,102) e' ,l i 100 - -

DPERATION I

RESTRICTED 4

(8.5,92) (34,92)

g O

is E l E 80 ( (0.80) (41,80)

!'I ie i y -

E

] ,% 60 -

ij E -

OPERATION (100,50) 4 -

l E ACCEPTABLE

o l S 40 -

!E E 15 f -

l 20 -

i

' I ' ' ' ' ' ' ' I 0

I O 20 40 60 80 100 1

1 APSR Position (% witnarawn) lI i

lI i

lI

, 8-17 Babcock & Wilcow

!I

It l

Figure 8-17. APSR Position Limits After 200 EFPD, Oconee 1, Cycle 7 (8.5,102) (34.5.102) 100 -

i a; 3

~

< (8.5,92) - (36,92) OPERATION ,

RESTRICTED 3 80 < (0,80) . (42.5,80)  ;

f '

a - ,

=

u 60 -

B t E i

- t

, OPERATION (100,50) I e

5 40 - ACCEPTABLE 2 i b '

f i 20 -

L I '

0 ' l i I I e ' ' ' I 0 20 40 60 80 100 l APSR Position (% witnarann)  ;

Ili Il:

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I REFERENCES 1

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

2 Mark BZ Demonstration Assemblies in Oconee 1, Cycles 7, 8, and 9, BAW-1661, Babcock & Wilcox, Lynchburg, Virginia, March 1981.

8 BPRA Retainer Design Report, BAW-1496, Babcock & Wilcox, Lynchburg, Virginia, May 1978.

J. H. Taylor (B&W) to S. A. Varga (NRC), Letter, "BPRA Retainer Reinser-tion," January 14, 1980.

5 Program to Determiae In-Reactor Performance of B&W Fuels - Cladding Creep Collapse, BAW,10084A, Rev. 2, Babcock & Wilcox, Lynchburg, Virginia, October 1978.

6 Y. H. Hsii, et al. , TACO 2 - Fuel Pin Thermal Analysis Co le, NPGD-TM-469, Rev. 2, Babcock & Wilcox, Lynchburg, Virginia, December 1979.

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

e I B&W Version of PDQ07 Code, BAW-10117A, Babcock & Wilcox, Lynchburg, Virginia, January 1977.

I Core Calculational Techniques and Procedures, BAW-10118, Babcock & Wilcox, Lynchburg, Virginia, October 1977.

10 Assembly Calculations and Fitted Nuclear Data, BAW-10116A, Babcock & Wilcox, Lynchburg, Virginia, May 1977.

11 Oconee Unit 1 Cycle 6 Reload Report, BAW-1552, Jabcock & Wilcox, Lynchburg, Virginia, July 1979.

12 Oconee 1 Fuel Densification Report, BAW-1388, Rev. 1, Babcock & Wilcox, Lynchburg, Virginia, July 1973.

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I 13 L. S. Rubenstein (NRC) to J. H. Taylor (B&W), Letter, " Evaluation of In-terim Procedure for Calculating DNBR Reductions Due to Rod Bow," October 18, 1979.

1" ECCS Analysis of B&W's 177-FA Lowered-Loop NSS, BAW-10103, Rev. 1, Babcock

& Wilcox, Lynchburg, Virginia, September 1975.

15 J. H. Taylor (B&W) to L. S. Rubenstein (NRC) , Letter, September 5,1980.

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

17 Normal Operating Controls, BAW-10122, Babcock & Wilcox, Lynchburg, Virginia, August 1978.

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

Wilcox, Lynchburg, Virginia, August 1976.

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