ML20097A129

Cycle 9 Reload Rept

ML20097A129
Person / Time
Site:	Oconee
Issue date:	08/31/1984
From:	BABCOCK & WILCOX CO.
To:
Shared Package
ML15264A250	List: ML20097A129
References
BAW-1841, NUDOCS 8409130187
Download: ML20097A129 (53)

Text

1 L 1 8AW-1841

[' August 1984

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[ OCONEE UNIT 1. CYCLE 9

-- Reload Report --

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1 OCONEE UNIT 1, CYCLE 9 s

! -- Reload Report --

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q CONTENTS Page

1. INTRODUCTION AND

SUMMARY

. . . . . . . . . . . . . . . . . . . . . 1-1

2. OPERATING HISTORY ........................

2-1 3.- GENERAL DESCRIPTION ....................... 3-1

4. FUEL SYSTEM DESIGN . . . . . . . . . . . . . . . . . . . . . . . . 4-1 L

4.1. Fuel Assembly Mechanical Design . . . . . . . . . ..... 4-1 4.2. Fuel Rod and Gray APSR Designs . . . . . . . . . . . . . . . 4-2 4.2.1. Cladding Collapse . . . . . . . . . . . . . . . . . 4-2

( 4.2.2. Cladding Stress .................. 4-2 4.2.3. Cladding Strain .................. 4-3 x\ 4.3.

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Thermal Design . . . . . . . .,. ._. . . . . . . . . . . . . 4-3 Material Design ...................... 4-4 ct 4.5. Operating Experience . . . . . . . . . . . . . . . . . . . . 4-4

'- 5. ' NUCLEAR DESIGN . . . . . . . . . ... . . . . . . . . . . . . . . . 5-1 5.1. Physics Characteristics '

.................. 5-1 5.2. Analytical Input . . . . . . .: . . . . . . . . . . . . . . . 5-2 5.3. Changes in Nuclear Design . . . ' . ' .. . . . . . . . . . . . 5-2

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

7. ACCiDENTANDTRANSIENTANALYSIS . . . ... . . . . . . . . . . . . 7-1 7.1. General SafetyiAnalysis ......y '

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[. 7.2. Accident Evaluation .........'........... '

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8. PROPOSED MODIFICATIONS TO TECHNICAL SPECIFICATIONS . . . . . . . . 8-1 (I ~"

9. REFERENCES .;. . . . . . . . . . . . . . . . . . . . . . . . . . . 9-1

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• List of Tables i h,,

.I'd Table 4-1. Fuel Design Parameters and Diinansions . . . . . .' . . . . . . . 4-5 A Fuel ThermalcAnalysis Parameters for Oconee 1, Cycle 9 . . . . . 4-6 q ; 4-2.

15-1. Oconee 1 Cycles 8 and 9. Physics Parameters . . . . . . . . , . . 5-3

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5-2. Shutdown Margin Calculations for Oconee 1, Cycle 9 . . . . . . . 5-5 6-2 6-1. Thennal-Hydraulic Design Conditions .............. -

7-1. Comparison of Key Parameters for Accident Analysis . . . . . . . 7-4 y LOCA Limits for Oconee 1, Cycle 9 ............... 7-4 =

7-2.

7-3. Comparison of FSAR and Cycle 9 Accident Doses ......... 7-5 Q E

b List of Figures Figure {

3-1. Core Loading Diagram for Oconee 1, Cycle 9 .......... 3-3 E 3-2. Enrichment and Burnup Distributions for Oconee 1, Cycle 9 . . . 3-4 ii 3-3. Control Rod Locations for Oconee 1, Cycle 9 . . . . . . . . . . 3-5 [

3-4. BPRA Concentration and Distribution for Oconee 1, Cycle 9 . . . 3-6 1 -

4-1. Gray Axial Power Shaping Rod ................. 4-7 .

5-1. Oconee 1, Cycle 9 B0C Two-Dimensional Relative Power "

Distribution - Full Power, Equilibrium Xenon, Nonnal -

Rod Positions . . . . . . . . . . . . . . . . . . . . . . . . . 5-6 2 8-1. Core Protection Safety Limits for Oconee Unit 1, Cycle 9 . . . 8-2 y 8-2. Protective System Maximum Allowable Setpoints for .-

Oconee Uni t 1, Cycl e 9. . . . . . . . . . . . . . . . . . . . . 8-3 8-3. Rod Position Limits for Four-Pump Operation, 0 to 30 E-a

+10/-0 EFPD, Oconee 1, Cycle 9 ................ 8-4 8-4. Rod Position Limits for Four-Pump Operation, 30 +10/-0 to 250 t10 EFPD, Oconee 1, Cycle 9 .............. 8-5 na 8-5. Rod Position Limits for Four-Pump Operation After -

250 10 EFPD, Oconee 1, Cycl e 9 . . . . . . . . . . . . . . . . 8-6 8-6. Rod Position Limits for Three-Pump Operation, O to 30 _

+10/-0 EFPD, Oconee 1, Cycle 9 ................ 8-7 E

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8-7. Rod Position Limits for Three-Pump Operation, 30 '

+10/-0 EFPD to 250 10 EFPD, Oconee 1, Cycle 9 ........ 8-8 _

Rod Position Limits for Three-Pump Operation After 8-8.

250 10 EFPD, Oconee 1, Cycle 9 . . . . . . . . . . . . . . . . 8-9 r 8-9. Rod Position Limits for Two-Pump Operation, 0 to 30

+10/-0 EFPD, Oconee 1, Cycle 9 ................ 8-10 -

8-10. Rod Position Limits for Two-Pump Operation, 30 F .

+10/-0 to 250 10 EFPD, Oconee 1, Cycle 9 . . . . . . . . . . . 8-11 8-11. Rod Position Limits for Two-Pump Operation After .-

250 t10 EFPD, Oconee 1, Cycle 9 . . . . . . . . . . . . . . . . 8-12  ;

8-12. Power Imbalance Limits for 0 to 30 +10/-0 EFPD,  ?

Oconee 1, Cycle 9 . . . . . . . . . . . . . . . . . . . . . . . 8-13 8-13. Power Imbalance Limits, 30 +10/-0 to 250 10 L EF P D , Oc o n ee 1, Cy cl e 9 . . . . . . . . . . . . . . . . . . . . 8-14 -

8-14. Power Imbalance Limits After 250 10 EFPD, Oconee 1, Cycle 9 . . . . . . . . . . . . . . . . . . . . . . . 8-15

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1. INTRODUCTION AND

SUMMARY

This report justifies the operation of the ninth 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 9 operation of Oconee 1, this report employs analytical techniques and design bases established in reports that have been submitted to and accepted by the USNRC and its predecessor (see references).

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.- A brief summary of cycle 8 'and 9 reactor parameters related to powar capa-bility is included in section 5 of this report. All of the accidents analyzed in the Final Safety Analysis Report (FSAR)1 h, ave been reviewed for cycle 9 operation. In those cases where cycle 9 characteristics were con-servative compared to those analyzed for previous cycles, no new accident analyses were parformed.

Four of the batch 10 assemblies are gadolinia lead test assemblies (LTAs).

These assemblies are part of a' joint Duke Power / Babcock & Wilcox (B&W)/De-

,H. partment of Energy (DOE) program to develop and demonstrate an advanced fuel assembly design incorporating U0 2-Gd 023 for extended burnup in pressur-ized water r.eactors (PWRs). Reference 2 describes the LTAs. Four Mark BZ demonstration fuel assemblies containing Zi rcaloy-4 intemediate spacer grids will be reinserted for a third cycle of irradiation. The Mark BZ dem-onstration assemblies are described in reference 3. The sixty-four batch s . 11 fuel assemblies are the Mark BZ type. The analyses supporting full g batch implementation of Mark BZ fuel are presented in reference 4. Gray

\- axit.1 pcwer shaping rods (APSRs) have also been included in the cycle 9 de-sign. The gadolinia LTAs, ' the Mark BZ assemblies, and the gray APSRs will 7 not adversely affect cycle 9 operation.

The Technical Specifications have been reviewed, and the modifications re-quired for cyble 9 operation are justified in this report.

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Based on the analyses perfomed, which account for the postulated effects of fuel densification and the final acceptance criteria for emergency core cooling systems (ECCS), it has been concluded that Oconee Unit 1 can be op-erated safely for cycle 9 at the rated power level of 2568 MWt.

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2. OPERATING HISTORY '

i The reference fuel cycle for the nuclear and thermal-hydraulic analyses of

( Oconee 1, cycle 9 is the currently operating cycle 8. The cycle 9 design length of 410 effective full power days (EFPD) is based on a planned cycle I 8 length of 410 EFPD. No operating anomalies have occurred during previous L-cycle operations that would adversely affect fuel performance in cycle 9.

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3. GENERAL DESCRIPTION I

The Oconee Unit i reactor core and fuel design basis are described in de-tail in section 3' of the FSAR for Oconee Nuclear Station, Unit 1.1 The h

cycle 9 core contains 177-fuel assemblies, each of which is a 15x15 array of 208 fuel rods, 16 control rod guide tubes, and one incore instrument guide tube. The fuel consists of di shed-end, cylindrical pellets of uranium dioxide clad in cold worked Zircaloy-4. The standard Mark B fuel assemblies in ~ all batches .have an average fuel loading of 463.6 kg of uranium. The undensified 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 9. Nineteen of the batch 9 assemblies will be discharged at the end of cycle 8, along with 44 batch 8C assemblies and the batch 10A assembly. The remaining 49 batch C assemblies (designated 9B) and the fresh batch 11 Mark BZ assemblies 4 with initial enrichments of 3.28 and 3.314 wt % 235U , respectively, will be loaded into the central portion of the core. The four batch 10B gadolinia LTAs,2 with an initial enrichment of 4.00 wt % 235U , are in locations sym-(.- metrical to H13. The batch 10C fuel, with an initiai enrichment of 3.41 wt

% 235U , will mainly occupy the core periphery. Figure 3-2 is an eighth core map showing the assembly burnup and enrichment distribution at the be-ginning of cycle (B0C) 9.

h Reactivity is controlled by 61 full-length Ag-In-Cd control rods, 60 burn-able poison rod assemblies (BPRAs), and soluble boron shim. In addition to the full-length control rods, eight Inconel gray APSRs are provided for ad-ditional control of the axial power distribution. Since gray APSRs are be-ing utilized, there are eight control rods in group 7 and twelve in group 5 to reduce the negative offset response to the group 7 rod movement. The cycle 9 locations of the 69 control rods and the group designations are 3-1 Bat > cock &Wilcox a McDermott company

indicated in Figure 3-3. The core locations with the exception of group 5 and group 7 are identical to t;iose of the reference cycle. The cycle 9 lo-cations and concentrations of the BPRAs are shown in Figure 3-4.

The system pressure is 2200 psia and the core average densified nominal heat rate is 5.80 kW/ft at the rated power of 2568 MWt for the standard l Mark B fuel assemblies.

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Figure 3-1. Core Loading Diagram for Oconee 1, Cycle 9

FUEL TRANSFER CANAL i x

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. 10C 10C 11 10C 10C A LCS KC4 F K12 L11 10C 10C 10C 11 10C 11 10C 10C 10C 8 Lp3 K@2 Np3 F f28 F N13 K14 L13

= 10C 11 10C 11 98 108 98 11 10C 11 10C C N11 F Kt6 F P18 Cf8 PC6 F K18 F Mo4 10C 11 98 11 98 11 98 11 98 11 98 11 10C g 9 C19 F 894 F A07 F R8 F AB9 F D14 F Cf6 E hh h N Ak h P$f h Phh h F$f h hh g 10C 10C 11 98 11 98 11 98 11 98 11 98 11 10C 10C F E18 C12 F Gal F 013 F R36 F C83 F GIS F CB4 EC6

- 10C 11 98 11 98 11 98 10C 98 11 98 11 98 11 10C

G Dj9 F L14 F M14 F N14 NOS P94 F MR2 F L82 F 097 11 10C 108 98 11 98 10C 98 10C 98 11 98 108 10C 11 ~

W-=H F H11 HC3 F L15 M12 H14 E04 Fo1 F Hal HC5 F a H.IS

• .. H.13 10C 11 98 11 98 11 98 10C 95 11 99 11 98 11 10C K N@9 F F14 F Ett F 812 011 082 F E82 F F02 F N#7 "m 10C 10C 11 98 11 98 11 98 11 98 11 98 11 10C 10C L M18 012 F K81 F C13 F A18 F C83 F K15 F 004 M06

, M 9 L B B Rh hh hh 10C 11 98 11 98 11 98 11 98 11 98 11 10C N 010 F N82 F R87 F A88 F Rf9 F P12 F 006 10C 11 10C 11 98 108 98 11 10C 11 10C E12 F G06 B10 008 886

] 0 F F G10 F 005 p IOC 10C 10C 11 10C 11 IX 10C 10C i F33 G82 093 F E28 F D13 G14 F13 f 10C 10C 11 1CC 10C g FB5 G44 F G12 Fil I

5 Z j 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 Batch 1.0.

• Gado11nia LTA

Previous Cycle Location ** Mark BZ LTA

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Figure 3-2. Enrichment and Burnup Distributions for Oconee 1, Cycle 9 8 9 10 11 12 13 14 15 3.280 3.410 3.280 3.314 3.280 4.000 3.410 3.314 H Mark BZ* LTA*

29,099 16,462 21,669 0 23,414 15,840 17,365 0 3.280 3.314 3.280 3.314 3.280 3.314 3.410 K

21,633 0 21,594 0 22,811 0 16,423 3.280 3.314 3.280 3.314 3.410 3.410 L

22,649 0 20,926 0 13,150 17,014 l

3.280 3.314 3.410 3.410

" l 21,666 0 17,292 14,266 3.280 3.314 3.410 f

N 21,634 0 16,015 3.410 16,492 P

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X.XXX Initial Enrichment XXXXX B0C Burnup, mwd /mtU t

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s Figure 3-3. Control Rod Locations for Oconee 1, Cycle 9 X

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A B 4 6 4 h C 2 5 5 2 P

D 7 8 7 8 7 4

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l F 4 8 6 3 6 8 4 G 5 1 3 3 1 5 L H W- 6 7 3 4 3 7 6

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1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 X Group Number

,Grouc No. of Rods Function 1 8 Safety 2 8 Safety 3 8 Safety 4 9 Safety 5 12 Control

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Figure 3-4. BPRA Concentration and Distribution for Oconee 1, Cycle 9 l 8 9 10 11 12 13 14 15 H 1.40 I K 1.40 1.40 0.20 L 1.40 1.40 0.50 M 1.40 1.40 1.40 N 1.40 1.40 0.00 0 0.50 0.00 P 0.20

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R X.XX BPRA Concentration, wt % B4 C in A10 23 l

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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 9 are listed in Table 4-1. All the fuel assemblies are C- mechanically _ interchangeable. Four twice-burned Mark BZ demonstration fuel assen611es are included in batch 98. The Mark BZ uses Zircaloy as the ma-( terial for the six intermediate spacer grids. The Mark BZ demonstration as-sembly is described in reference 3, which concludes that reactor safety and perfonnance are not adversely affected by the presence of the four demon-stration ; assemblies. The four fuel assemblies in batch 10B are gadolinia LTAs. The mechanical design of the LTAs is described in reference 2. The 64 fuel assemblies in batch 11 are also Mark BZ assemblies. The mechanical

= . design of the Mark BZ fuel is described in reference 4. The balance of the fuel assemblies in the cycle 9 core are the standard Mark B type.

-Retainer assemblies will be.used on the two fuel assemblies that contain re-

{ generative neutron source assemblies and on the 60 batch 11 assemblies that contain BPRAs. The 62 retainers will be exposed for a fourth cycle of -ir-radiation during cycle 9. This additional cycle of irradiation is justi-fied based on an examination of retainers which have undergone three cycles

[ of irradiation. The . results of the examination meet criteria developed earlier in terms of wear and holddown force. These criteria ensure that h the retainers will perform in a safe and adequate manner in the areas of holddown force, stress, and fatigue during a fourth cycle of in-reactor use. These criteria were developed from analyses similar to those per-

{r formed in the original justification of the design and use of the retainer assemblies in references 5 and 6. The justification for the fourth cycle of irradiation is given in reference 7.

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4.2. Fuel Rod and Gray APSR Designs Mechanical evaluations of the fuel rods and gray APSRs are discussed below.

4.2.1. Cladding Collapse The fuel assemblies of batch 98 are more limiting than those of other batches because of their longer previous incore exposure time. The power history and fuel design parameters for the most limiting batch 9B fuel as-sembly were compared with those used in the generic Mark-B creep collapse analysis and were found to be enveloped. .The generic analysis was based on the methods and procedures described in reference 8 and is applicable to the batch 9B fuel design. The generic analysis predicts a collapse time of more than 35,000 effective full power hours (EFPH), which exceed the maxi-mum projected residence time of 29,359 EFPH (Table 4-1).

A detailed creep analysis was performed on the gadolinia-bearing fuel rods in the LTAs. The collapse time for these rods was greater than the maximum projected residence time.

The gray APSRs that are to be used in cycle 9 were designed to improve creep 'li fe. Cladding thickness and rod ovality control, which are the pri-mary factors controlling the creep life of a stainless steel material, have been improved to extend the creep life of the gray APSR. The minimum de-sign cladding thickness of the Mark B APSR is 18 mils, while that of the gray APSR is 24 mils. Additionally, the gap width between the end plug and Inconel absorber material was reduced. Finally, the ovality in the gap area will also be controlled to tighter tolerances. The gray APSR is shown in Figure 4-1.

4.2.2. Cladding Stress The stress parameters for the Oconee 1 standard fuel rods and the gadolinia-bearing fuel rods are enveloped by a conservative fuel rod stress analysis.

The following four assumptions were used in this analysis:

1. A lower post-densification internal pressure.

2. A lower initial pellet density.

3. A higher system pressure.

4. A higher thermal gradient across the cladding.

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L For design evaluation, the primary membrane stress must be less than two-thirds of the minimum speci fied unirradiated yield strength, and all

e. tresses (primary and secondary) must be less than the minimum specified un-f rradiated yield strength. In all cases, the margin is in excess of 30%.

The gray APSR design was analyzed to demonstrate that it meets specified de-( sign requirements. The APSR was analyzed for cladding stress due to pres-I sure, temperature, and ovality. It was found that the gray APSR nas suf-F ficient cladding and weld stress margins.

H 4.2.3. Cladding Strain I The fuel design criteria specify that the cladding average circumferential L

strain is not to exceed 1% inelastic strain. The pellet design is estab-( lished for a plastic cladding strain of less than 1% at the maximum design local pellet burnup and heat generation rate. These values are higher than the values the Oconee 1 UO2 fuel is expected to see. A strain analysis of the gadolinia fuel showed that the calculated strains for these rods are also below design limits. Thus, fuel rod cladding strain will not affect cycle 9 fuel performance.

The gray APSR was analyzed for cladding strain due to themal and irradia-

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tion swelling. The results of this analysis showed that no cladding strain is induced due to thennal expansion or irradiation swelling of the Inconel absorber.

4.3. Thermal Design All fuel rods in cycle 9 are thermally similar, except for the urania-gado-linia fuel in four LTAs. The analysis for reinserted batches 9B,10B, and

{ 10C and feed batch 11 fuel was perfomed with the TAC 029 code using the methodology described in reference 10. Nominal undensified input param-eters used in this methodology are presented in Table 4-2. Densification effects were accounted for in TAC 02.

Centerline fuel melt (CFM) limits of 20.5 kW/ft for 95% theoretical density (TD), pure U02 fuel, and 17.6 kW/f t for UO -Gd 2 023 fuel were predicted using the thermal code TAC 02. The fuel internal pressure in the highest burnup rod is predicted to reach the nominal reactor coolant system pressure of 4-3 Babcock & WHcom a McDermott company

l 2200 psia af ter 45,000 mwd /mtU for both pure U02 and UO2 -Gd23 0 . The maxi-mum burnup of any UO2 fuel rod in cycle 9 is less than 45,000 mwd /mtu; the highest burnup of any UO2 -Gd 023 fuel rod is less than 25,000 mwd /mtu.

4.4. Material Design The batch 11 fuel assemblies are not new in concept, nor do they utilize different component materials, except for the Zircaloy spacer grids and the Inconel 718 holddown springs. These materials have been used with success in similar reactor environments. Therefore, the chemical compatibility of all possible fuel-cladding-coolant-assembly interactions for the batch 11 fuel assemblies is acceptable.

4.5. Operating Experience B&W operating experience with the Mark B 15x15 fuel assembly has verified the adequacy of its design. As of April 30, 1984, the following experience has been accumulated for eight B&W 177-fuel assembly (FA) plants using the Mark B fuel assembly: j Ma F ""E' HWd mtU Cumulative net Current elec c Reactor cycle Incore Discharged output, MWh Oconee 1 8 34,499 50,598 48,808,138 Oconee 2 7 27,035 36,800 43,444,856 Oconee 3 8 35,123 35,463 45,200,486 Three Mile Island 5 25,200 32,400 23,840,053 g

Arkansas Nuclear 6 31,450 36,540 38,872,852 a One, Unit 1 Rancho Seco 6 30,500 38,268 33,923,457 l

Crystal River 3 5 23,170 29,900 27,083,428 Davis-Besse 4 28,520 32,790 19,237,628 (a)As of April 30, 1984.

(b)As of January 31, 1984.

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Table 4-1. Fuel Design Parameters and Dimensions Batch 9B Batch 108/10C Batch 11 FA type Mark B/ Mark GdB/ Mark BZ Mark BZ Mark B No. of FAs 45/4 4/60 64 Fuel rod 00, 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 Undensified active fuel-length 141.8 141.8

(' (nom. ), i n, 143.5/141.8 y Fuel pellet initial density (nom.), 95 95 95

(  % TD Fuel pellet 00 (mean specification), 0.3686 0.3686 0.3686 in.

r Initial fuel enrichment, wt % 2350 3.28 4.0/3.41 3.314 B0C burnup (avg), mwd /mtU 22,090 15841/15911 0 Cladding collapse time, EFPH >35,000 >35,000 >35,000 Estimated residence time (max.), 29,359 19,680 9,840 EFPH

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Table 4-2. Fuel Themal Analysis Parameters --

Oconee 1, Cycle 9 Batch 9B(a) 10B(b) 10C 11(c)

No. of assemblies 49 4 60 64 Nominal pellet density, % TD 95 95 95 95 Pellet diameter, in. 0.3686 0.3686 0.3086 0.3686 Stack height, in. 141.8 143.5 141.8 141.8 Nominal LHR(d) 0 2568 MWt, kW/ft 5.74 5.68 5.74 5.74 l

LHR to CL fuel melt, kW/ft 20.5 17.6(e) 20.5 20.5 (a) Includes four Mark BZ demonstration assemblies.

(b)Gadolinia LTAs.

(c) Sixty-four Mark BZ assemblies, including 60 BPRAs.

(d)LHR denotes linear heat rate.

(e) Reduced for gadolinia fuel.

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5. NUCLEAR DESIGN 5.1. Physics Characteristics Table 5-1 compares the core physics parameters of design cycle 9 with those

} 1 of the reference cycle 8. The values for both cycles were generated using PDQ0711-13 The average cycle 9 burnup will be the same as that of the cycle 8 design. Figure 5-1 illustrates a representative relative power dis-tribution for the beginning of cycle 9 at full power with equilibrium xenon f 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 criti-cal baron concentrations for cycle 9 are higher because of the gray APSRs and previous cycle reactivity carryover. The control rod worths differ between cycles due to the gray APSRs, changes in rod groupings for tran-sient banks 5 and 7, changer in radial fl ux, and burnup distributions.

This also accounts for the smaller ejected and stuck rod worths in cycle 9 compared to cycle 6 values. Calculated ejected rod worths and their ad-herence 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. These rod worths meet all safety criteria. The adequacy of the shut-down margin with cycle 9 stuck rod worths is demonstrated in Table 5-2.

The following assumptions were applied for the shutdown calculations:

1. Poison material depletion allowance.

2. 10% uncertainty on net rod worth.

3. Flux redistribution penalty.

Flux redistribution was accounted for since the shutdown analysis was cal-culated using a two-dimensional model. The reference fuel cycle shutdown margin is presented in the reload report for Oconee 1, cycle 814 The cycle 9 power deficits, dif ferential boron worths, and effective delayed 5-1 Babcock &Wilcox a McDermott company

neutron fractions differ from those for cycle 8 because of the higher criti-cal boron concentrations.

5.2. Analytical Input The constants used to compute core power distributions from incore detector g measammer.ts were obtained in the same manner for cycle 9 as for the ref- e eren:e cycle 8. The monitoring of power distributions in the LTAs is dis-cussid in reference 3.

5.3. Changes in Core Design Core design changes for cycle 9 are the use of gray APSRs and the introduc-I tion of 64 Mark BZ assemblies in addition to the four Mark BZ LTAs loaded g in cycle 7. Gray APSRs, which are longer and use a weaker Inconel ab- 4 sorber, replace the sil ver-indium-cadmium APSRs used in all previous cycles. Calculations with the standard three-dimensional model verified that these APSRs provide adequate axial power distribution control.

1 I

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5-2 Babcock &Wilcox a McDermott company

[

Table 5-1. Oconee 1 Cycles 8 and 9 Physics Parameters (a)

Cycle 8(b) Cycle 9(c)

Cycle length, EFPD 410 410 Cycle burnup, mwd /mtU 12,858 12,858 Avg. core burnup, E0C, mwd /mtU 24,183 24,724 Initial core loading, mtU 82.1 82.1 Critical horon, B0C, (No Xe), ppm

. HZPLG3, group 8 inserted 1602 1682 HFP(d) , group 8 inserted 1365 1454

, Critical boron, E0C, (eq Xe), ppm HIP, group 8 inserted 401 457 HFP, group 8 inserted 60 121 Control rod worths, HFP, BOC, % ak/k Group 6 0.98 1.09 Group 7 1.47 0.93 Group 8 0.42 0.10 Control rod worths, HFP, E0C, % ak/k Group 7 1.54 1.00 Group S 0.49 0.10

(

Max ejected rod worth, HZP, % ak/k(e)

B0C (N12) 0.59 0.35 EOC (L10) 0.48 0.36 Max stuck rod worth, HZP, % ak/k B0C (H14) 1.68 1.31 EOC (N12) 1.72 1.35 Power deficit, HZP to HFP, % ak/k

{ BOC 1.62 1.51 E0C 2.36 2.29 Doppler coeff., HFP, 10-5 (ak/k/F)

BOC, 100% power, no Xe -1.54 -1.53 E0C,100% power, eq Xe -1.78 -1.77 Moderator coeff., HFP,10-4 (ak/k/F)

B0C, (0 Xe, crit ppm, gp 8 ins) -0.67 -0.47 EOC, (eq Xe, 17 ppm, gp 8 ins) -2.85 -2.75 Boron worth, HFP, ppm /% ak/k r BOC (1425 ppm) 129 130 L E0C (17 pp;n) 110 111 5-3 hock &WIlcom

Table 5-1. (Cont'd)

Cycle 8(b) Cycle 9(c)

Xenon worth, HFP, % ak/k B0C (4 EFPD) 2.54 2.53 E0C (equilibrium) 2.67 2.66 Effective delayed neutron fraction, HFP B0C 0.00625 0.00621 E0C 0.00526 0.00526 (a) Cycle 9 data are for the conditions stated in this report. The cycle 8 core conditions are identified in reference 14.

(b) Cycle'8 data are based on a cycle 7 length of 420 EFPD.

(c) Cycle 9 data are based on a cycle 8 length of 410 EFPD.

(d)HZP denotes hot zero power (532F Tavg); HFP denotes hot full power (579F TavgI -

(e) Ejected rod worth for groups 5 through 7 inserted.

N 5-4 Babcock &WIlcos a MCDermott COmpafly

L Table 5-2. Shutdown Margin Calculations for Oconee 1, Cycle 9 BOC, % ak/k E0C, % ak/k Available Rod Worth Total rod worth, HIP (a) 8.14 8.82 Worth reduction due to burnup of poison material -0.42 -0.42

[ Maximum stuck rod, HZP(a) -1.31 -1.35 Net worth 6.41 7.05 Less 10% uncertainty -0.64 -0.71 Total available worth 5.77 6.34 Required Rod Worth Power deficit, HFP to HZP(a) 1.51 2.29 Max allowable inserted rod worth 0.30 0.50 Flux redistribution 0.81 1.20 Total required worth 2.62 3.99 Shutdown margin (total available worth minus total required worth) 3.15 2.35 Note: The required shutdown margin is 1.00% ak/k.

(a)HZP denotes hot zero power and HFP denotes hot full power.

~

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[ 1 5-5 Babcock &WIIcom a McDermott company

)

Figure 5-1. Oconee 1, Cycle 9 BOC (4 EFPD) Two-Dimensional Relative Power Distribution -- Full Power, Equilibrium Xenon, Normal Rod Positions 8 9 10 11 12 13 -14 15 0.848 1.059 1.042 1.203 1.061 1.193 1.130 0.850 H

Mark BZ* LTA*

K 0.996 1.169 1.077 1.210 1.073 1.193 0.639 L 1.053 1.198 1.096 1.277 0.969 0.432 M 1.097 1.213 1.098 0.702 N 1.040 1.039 0.465 0 0.572 P

R

• Demonstration assemblies.

Inserted Rod Group No.

X.XXX Relative Power Density 5-6 "*" & Mb8 a McDermott company

r k

k c

6. THERMAL-HYDRAULIC DESIGN

(

The thermal-hydraulic design evaluation supporting cycle 9 operation uti-lized the methods described in references 1, 5, and 15. Section 5 of ref-erence 4 demonstrates that a full . Mark BZ core and a full Mark B core pro-( vide practically the same departure from nucleate boiling (DNB) margin for both steady-state and transient conditions and that the current pressure-( temperature trip envelope is conservative for a full Mark BZ core.

-The cycle 9 transition core includes 64 fresh Mark BZ batch 11 fuel assem-blies, 60 of which will contain BPRAs. Batch 10B LTAs and the 4 demonstra-tion assemblies in batch 98 also incorporate Mark BZ spacer grids. The ef-fect 'of the higher pressure drop caused by the Mart BZ grids and by the BPRA retainers 7 in a predominantly Mark B core is a slightly lower flow in the Mark BZ assemolies. The DNB margin for the Mark BZ assemblies is re-duced as a result. To preserve the DNB margin, the radial-local design peaking is reduced to 1.67 for the Mark BZ assemblies. This peaking reduc-tion ensures a comparable DN8 margin for the limiting transient and for steady-state operations for the cycle 9 transition core and the Mark B ref-erence core. The 1.71 radial-local peaking still applies to the Mark B as-semblies. The 1.67 value is applicable to cycle 9 only.

( The maximum expected peaking during cycle 9 is 1.416. No fuel' rod bow pen-alty has been incorporated into the core departure from nucleate boiling l ratio (DNBR) limits, as justified by reference 16.

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> Table 6-1. Thermal-Hydraulic Design Conditions Cycle 8 Cycle 9 Power level, MWt 2568 2568 System pressure, psia 2200 2200 Reactor coolant flow, % design flow 106.5 106.5 Vessel inlet coolant temp. at 100% power, F 555.6 565.6 Vessel outlet coolant temp. at 100% power, F 602.4 602.4 Ref. design axial flux shape 1.5 cos 1.5 cos Ref. design radial-local peaking factor 1.71 1.67(a)

Active fuel length 140.3(b) 140.3(b)

Average heat flux at 100% power,10 3 Btu /h-ft2 176(c) 176(c) 3 Maximum }ocal heat flux at 100% power,10 Btu /h-ft 451(d) 441(e)

Critical heat flux (CHF) correlation B&W-2 BWC CHF correlation limit 1.30 1.18 Hot channel factors -- Enthalpy rise 1.011 1.011 Heat flux 1.014 1.014 Flow area 0.98 0.97 Minimum DNBR (112% power) 2.05(f) 1.74 (a) Applicable to cycle 9 only.

(b)Used in DNBR calculations and to calculate the next two items in this table.

(c) Based on an expanded rod OD of 0.43075 in.

(d) Based on a length of 140.3 in., rod OD of 0.43075 in., axial peak of 1.5', and radial-local peak of 1.71.

(e) Based on a length of 140.3 in., rod OD of 0.43075 in., axial peak of 1.5, and radial-local peak of 1.67.

If} Based on 52 open assemblies.

L 6-2 Babcock &WIIcom a McDermott company

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.7. ACCIDENT AND TRANSIENT ANALYSIS 7.1. General Safety Analysis

( _.Each FSAR1 accident analysis has been examined with respect to changes in cycle 9 parameters .to determine the effects of the cycle 9 reload and to J

ensure that thermal perfomance during hypothetical transients is not de-graded.

The effects of fuel. densification on the FSAR accident results have been evaluated and are reported in BAW-1388.15 Since batch 11 reload . fuel as-

-semblies contain fuel rods with a higher theoretical density than those con-sidered :in the reference 15 report, ' the conclusions in the reference are still valid.

7.2.- ' Accident Evaluation The key parameters 'in detemining the outcome of a transient can typically be classified in three major areas: core thermal parameters, thermal-hy-draulic parameters, and kinetics parameters, including the reactivity feed-back coefficients and control rod worths.

Core themal properties used in the FSAR accident analysis were design op-erating values based on calculational values plus uncertainties. Fuel h- themal analysis values for each batch in cycle 9 are compared in Table

, L4-2. The cycle 9 thermal-hydraulic maximum design conditions are compared h to the previous cycle 8 values in Table 6-1. These parameters are common to all of the accidents considered in this report. The key kinetics param-eters from the FSAR and cycle 9 are compared in Table 7-1.

A generic LOCA analysis for the B&W 177-FA, lowered-loop nuclear steam sup-ply (NSS) system has been perfomed using the final acceptance criteria ECCS evaluation model. This study is reported in BAW-10103, Rev. 3.17 The analysis in . BAW-10103 is generic since the limiting values of key param-eters for. all plants in this category were used. Furthemore, .the 1

7-1 Batscock&WI8com

combination of average fuel temperature as a function of LHR and the life-time pin pressure data used in the BAW-10103 l oss-of-coolant accident I (LOCA) limits analysis is conservative compared to those calculated for this reload. Thus, the analysis and the LOCA limits reported in BAW-10103 provide conservative results. Table 7-2 shows the bounding values for al-lowable LOCA peak LHRs for Oconee 1, cycle 9 as a function of burnup. The LOCA kW/ft limits have been reduced for low burnups to ensure conservative limits based upon an interim bounding analytical assessment of NUREG 0630 on LOCA and operating kW/f t limits 21 The Oconee 1, cycle 9 core contains four Mark BZ demonstration assemblies, I

four gadolinia LTAs, and 64 Mark BZ assemblies. As a result of material and geometrical differences, these assemblies have LOCA kW/f t limits that are lower in some cases than the standard Mark B limits. The four gadolinia LTAs are non-limiting. The Mark BZ LTAs were analyzed using LOCA kW/ft limits consistent with the 64 fresh Mark BZ assemblies.

It is concluded from the examination of cycle 9 core thermal and kinetics properties, with respect to acceptable previous cycle values, that this core reload will not adversely affect the ability of the Oconee 1 plant to operate safely during cycle 9. Considering the previously accepted design basic, used in the FSAR and subsequent cycles, the transiera evaluation of cycle 9 is considered to be bounded by previously accepted analyses. The initial conditions for the transients in cycle 9 are bounded by the FSAR 1, the fuel densification report 15, and/or subsequent cycle analyses.

The radiological dose consequences of the accidents presented in chapter 15 of the FSAR were recalculated using the specific parameters applicable to cycle 9. The bases used in the dose calculations are identical to those in the FSAR except that updated dose conversion factors were used. The use of the updated dose conversion factors resulted in reduced whole body dose g values. I Table 7-3 compares the revised FSAR dose values with those calculated spe-cifically for cycle 9. As can be seen from the table, some cycle 9 doses vary slightly from the FSAR values. However, all cycle 9 doses are either bounded by the values presented in the FSAR or are a small fraction of the 10 CFR 100 limits, i.e. below 30 rem to the thyroid or 2.5 rem to the whole j u

7-2 Babcock &WHcom f a McDermott company (

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l body. Thus, the radiological impact T of the accidents during cycle 9 are not significantly different than those described in chapter 15 of the FSAR.

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k Table 7-1. Comparison of Key Parameters for Accident Analysis FSAR and Predicted $

densification cycle 9 e Parameter report value value -

Doppler coeff,10-5 Ak/k/F BOC -1.17 -1.53 T E0C -1.33 -1.77 _

Moderator coeff, 10-4 Ak/k/F B0C +0.5 -0.47 .

E0C -3.0 -2.75 -

A All-rod group worth at HZP, _

% ak/k 10 8.14 y Initial boron cone'n at HFP, ppm 1400 1454(a)

Boron reactivity worth at 70F, -

ppm /1% ak/k 75 91 Max ejected rod worth at HFP, F.

% ak/k 0.65 0.22 r Dropped rod worth (HFP), % ak/k 0.46 0.20 5 W

w (a)The combined effect of boron concentration and boron worth is _

conservative for cycle 9. I Table 7-2. LOCA Limits for Oconee 1, Cycle 9 Linear heat rate, kW/ft -

Elevation, 0-30 +10/-0 30 +10/-0 to 250 10 EFPD ft EFPD 250 t10 EFPD to E0C 2 13.5 15.0 15.5 4 16.1 16.6 16.6 -_

6 17.5(a) 18.0 18.0 8 17.0 17.0 17.0 -

10 16.0 16.0 16.0 g (a)For the Mark BZ assemblies the LOCA limit is 16.5

=

kW/f t at the 6-f t elevation for the 0-30 +10/-0 EFPD window only.

7-4 Babcock &WHcom -

a McDermott company

s I

Table 7-3. Comparison of FSAR and Cycle 9 Accident Doses FSAR doses,(a) Cycle 9 doses, a rem rem

1. Fuel Handling Accident Thyroid dose at EAB, 2 h 0.50 0.50 Whole body dose at EAB, 2 h 0.028 0.010

2. Steam Line Break Thyroid dose at EAB, 2 h 0.20 0.20 Whole body dose at EAB, 2 h 0.002 0.001

3. Steam Generator Tube Failure Thyroid dose at EAB, 2 n 0.31 0.31 Whole body dose at EAB, 2 h 0.058 0.027

4. Waste Gas Tank Rupture Thyroid dose at EAB, 2 h 0.27 0.28 Whole body dose at EAB, 2 h 0.17 0.08

5. Control Rod Ejection Accident Thyroid dose at EAB, 2 h 1.44 1.40 Whole body dose at EAB, 2 h 0.004 0.002

.i Thyroid dose at LPZ, 30 days 1.57 1.55 Whole body dose at LPZ, 30 days (b) 0.002

6. Loss-of-Coolant Accident Thyroid dose at EAB, 2 h 5.0 4.97 Whole body dose at EAB, 2 h 0.010 0.005 Thyroid dose at LPZ, 30 days 5.5 5.51 1 Whole body ~ dose at LPZ, 30 days 0.010 0.007

7. Maximum Hypothetical Accident I Thyroid dose at EAB, 2 h 193 193 Whole body dose at EAB, 2 h 1.4 1.11 Thyroid dose at LPZ, 30 days 180 180 Whole body dose at LPZ, 30 days 0.62 0.44 (a)FSAR changed since cycle 7 reload.

(b)Not listed in FSAR.

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7-5 Babcock & WHcom a McDermott company

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8. PROPOSED MODIFICATIONS TO TECHNICAL SPECIFICATIONS The Technical Specifications have been revised for cycle 9 operation in accordance with the methods of references 18 through 21 to account for chanqes in power peaking and control rod worths as well as the inclusion of gray APSRs in the design.

Based on the Technical Specifications derived from the analyses presented in this report, the final acceptance criteria ECCS li. nits will not be ex-ceeded, and the thennal design criteria will not be violated. Figures 8-1 l

through 8-14 are revisions to previous Technical Specification limits. The allowable withdrawal limit for gray APSRs is 0-100% withdrawn at any time during cycle 9. Therefore the figures pertaining zo APSR withdrawal limits have been deleted.

l 8-1 Babcock & Wilcox a McDermott company i--. _ ..

, . . . . . . . . . . ~. _. . _ .. . . , - . .. .. _ . _ _ . ..

Figure 8-1. Core Protection Safety Limits for Oconee Unit 1, Cycle 9 (Technical Specification Figure 2.1-2A)

Themal Power Level, %

. 120

(-32.4,112) . (31.5,112) 110 (ACCEPTABLE M1 = 0.870 i 4 PUMP -

- 100 l M2 = -1.049 IOPERATION

(-53.9,93.3) l(-32.4,88.8). 90 (31.5, 88.8,)(53.9, 88.5)

ACCEPTABLE l l4&3 PUMP -

- 80 l0PERATION l

(-53.9,70.1) -

- 70 l l *

,(-32.4 61.1) ,(31.5,61.1)

ACCEPTABLE 60 l

14,3&2 PUMP -

- 50 1 OPERATION

(-53.9,42.4) l -

- 40 (53.9,37.6) g

- 30

- 20 ql cn UNACCEPTABLEi 7l d UNACCEPTABLE OPERATION n n l

- 10 Ml u n OPERATION A E f il

• i i i li i i '

i i i f J

-60 -40 -20 0 20 40 60 R'eactor Power Imbalance, %

CURVE RC FLOW (GPM) 1 374,880 2 280,035 3 183,690 2.1-7 8-2 Babcock &Wilcon a McDermott compariy

e Figure 8-2. Protective System Maximum Allowable Setpoints "

for Oconee Unit 1, Cycle 9 (Technical Specification Figure 2.3-2A)

Thennal Power Level, %

- - 120

(-17,108) . (17,108) -

M1*1 CCEPTABLE - - 100 l M2 = -1.209 ._

4 PUMP ,

l OPE}ATIO!i

(-40,85) (-17,80.67) _

80

! (40,80.19) .

ACCEPTABLE. _

UNACCEPTABLE 4 & 3 PUMP l UNACCEPTABLE OPERATION b

OPERATION OPQRATION

(-40,57.67)

/ ' --

60 l

(-17,52.92)I 1(17,52.92),

(40,52.86) l l ACCEPTABLE l l 4, 3 & 2 - - 40 PUMP l

(-40,29.92) OPERATION _ _

l (40,25.11)

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- 20, l-e A "J El E i i i cal I t i i l1 1 l l 1

-60 -40 -20 0 20 40 60 Reactor Power Imbalance, %

2.3-8 8-3 Babcock &Wilcox a McDermott company

I Figure 8-3. Rod Position Limits for Four-Pump Operation, O to 30 +10/-0 EFPD, Oconee 1, Cycle 9 (Technical Specification Figure 3.5.2-1A1) 100 -

(137,102) (270,102) ^ ,

OPERATION NOT ALLOWED SHUTDOWN (264,92) e MARGIN T LIMIT l 80 - (200,80)

I OPERATION RESTRICT g

2 r 60 ~

S (89,50)

/

OPERATION g 40 -

ACCEPTABLE E

E

$ 20 _

O (45,15) 0- ' '

0 100 200 300 GR 5 ' ' J 0 75 100 0 25 75 100 GR 7 I I '

0 25 100 3.5-15 8-4 Babcock &WHcom a McDermott company

L F

L Figure 8-4. Rod Position Limits for Four-Pump Operation, 30 +10/-0 to 250 t10 EFPD, Oconee 1, Cycle 9 (Technical Specification Figure 3.5.2-1A2) e L (264,^102) 100 -

OPERATION (177,102)

NOT ALLOWED SHUTDOWft (258,92) g MARGIN 80 N (226,80)

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OPERATION RESTRICT

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" 60 -

E (133,50) t y 40 _

E g OPERATION

[ -

ALLOWED

$ 20 _

b (70,15)

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0 100 200 300 GR 5 I I I Rod Index, % Withdrawn 0 75 100 GR 6 i i i

( 0 25 75 1

100 GR 7 I I I

[ 0 25 100

[ 3.5-15a 8-5 Bahcock &Wilcom a McDermott company

Figure 8-5. Rod Position Limits for Four-Pump Operation After 250 t10 EFPD, Oconee 1, Cycle 9 (Technical Specification Figure 3.5.2-1A3)

(216,1*) (262,102)

~

OPERATION SHUTDOWN NOT ALLOWED MARGIN (250,92)

LIMIT 80 -

.8 00 -

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E (150,50) t y 40 _

g OPERATION

- ACCEPTABLE I 20 -

(83,15)

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(0,8.0)

I I 0

0 100 200 300 Rod Index, % Withdrawn 0 75 100 GR 6 i i i i 0 25 75 100 GR 7 i i i 0 25 100 s

3.5-15b 8-6 Babcock &Wilcox a McDermott company

5 Figure 8-6. Rod Position Limits for Three-Pump Operation,

- 0 to 30 +10/-0 EFPD, Oconee 1, Cycle 9 L (Technical Specification Figure 3.5.2-2A1)

F b 100 -

x 2 SHUTDOWN

- MARGIN LIMIT (272,_77)

( $ 80 (137,77) (264,69) y r

" OPERATION OPERATION 60 -

NOT ALLOWED RESTRICTED (200,60) b 40 -

89,38)

( b OPERATION 20 -

{ ACCEPTABLE E <

(45,11.7)

( f 0

(0,9.3) 1 I 0 100 200 300 Rod Index, % Withdrawn GR 5 I I I 0 75 100 GR 7 I I I I

. 0 25 75 100 GR 71 8 1 0 25 100

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3.5-18

[ 8-7 Babcock & Wilcox a McDermott company f - - -

Figure 8-7. Rod Position Limits for Three-Pump Operation, 30 +10/-0 EFPD to 250 t10 EFPD, Oconee 1, Cycle 9 (Technical Specification Figure 3.5.2-2A2)

T OPERATION y 100 -

NOT ALLOWED E

g SHUTDOWN (177,77) (266,77) e, 80 -

MARGIN ^

(258,69 3 60 ~

OPERATION (242,60) g RESTRICT t

+*

40 5

133,38)

E E

20 OPERATION

!D ACCEPTABLE

$ (70,11.7) i ,6.8) 0 I I 0 100 200 300 GR 5 e I i Rod Index, % Withdrawn 0 75 100 I GR 6 i i 1 0 25 75 100 GR 7: I i 0 25 100 3.5-18a 8-8 Babcock &Wilcom a McDermott company

s Figure 8-8. Rod Position Limits for Three-Pump Operation P After 250110 EFPD, Oconee 1, Cycle 9 L (Technical Specification Figure 3.5.2-2A3)

L T OPERATION 100 -

RESTRICTED 7

g 0PERATION NOT ALLOWED (216,77) (266,,77)

( 80 -

250, 69)

( i 60 SHUTDOWN MAP, GIN L  %

242,60)

LIMIT a

40 -

150,38)

E Te 20 - OPERATION ACCEPTABLE 5 (83,11.7) 0 I '

0 100 200 300 GR 5 I I I O 75 100 GR 6 i i i  :

0 25 75 100 GR 7 e i i 0 25 100

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s s

3.5-18b

[ 8-9 Babcock &Wilcox a McDermott company

]

Figure 8-9. Rod Position Limits for Two-Pump Operation, O to 30 +10/-0 EFPD, Oconee 1, Cycle 9 (Technical Specification Figure 3.5.2-2A4)

T j 100 -

n.

- OPERATION NOT ALLOWED h 80 _

?u j 60 -

(137,52) (273,52)

I SHUTDOW OPERATION N

RESTRICTE 264,46)

$ 40 -

r1 g (200,40) c.

~

89,26) 20 -

b OPERATION

$ ACCEPTABLE 0 (O. 6.9) I i 0 100 200 300 GR 5 1 1 0 75 100 GR 6 : I i 1 0 25 75 100 GR 7 i i I l 0 25 100

]

3.5-18c 8-10 Babcock &Wilcox a McDermott company L . .

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5 Figure 8-10. Rod Position Limits for Two-Pump Operation, F 30 +10/-0 to 250110 EFPD, Oconee 1, Cycle L 9 (Technical Specification Figure 3.5.2-2AS)

~

L

{ $ 100 -

[ OPERATION

- NOT ALLOWED

( [

w 80 -

N 3 60 -

,% (177,52) (267,52)

% SHUTDOWN (258,46)

[ 40 -

MARGIN 242,40)

U* LIMIT OPERATION RESTRICTED

[ k (133,2

- 20 - (235,26) e OPERATION 3

{ e ( Nu,5.2) * (70, 8.5) ACCEPTABLE 0 100 200 300

{ Rod Index, % Withdrawn GR 5 i i i 0 75 100

[ GR 6 i e i i 0 25 75 100 GR 7 i i i 0 25 100

[

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[ 8-11 Babcock &Wilcox a McDermoit company

]

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Figure 8-11. Rod Position Limits for Two-Pump Operation After 250 t10 EFPD, Oconee 1, Cycle 9 (Technical Specification Figure 3.5.2-2A6) i T

I g 100 -

80 Y

E 60 -

SHUTDOWN (216,52) (268,52)

% MARGIN -

LIMIT j

40 -

(250,46) 242,40)

)

g OPERATION OPERATION S RESTRICT q NOT ALLOWED u (150,26 (235,26) J y 20 -

2 OPERATION ACCEPTABLE g%o,5.0) -(8h8 5) 0 100 200 GR 5 t i I Rod Index, % Withdrawn O' 75 100 GR 6i i i I O 25 75 100 GR 71 1 1 0 25 100

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3.5-18e 8-12 Babcock &WHcom a McDermott company

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Figure 8-12. Power Imbalance Limits for 0 to 30 +10/-0 EFPD, Oconee 1, Cycle 9 (Technical L Specification Figure 3.5.2-3A1)

(-19,102) (24,102) k -

- 100 b (-20,92) (25,92)

_ _ 99

( (-27,80)

T

$-- 80

'\h(31,80)

[

OPERATION @-- 70 ACCEPTABLE y

" - - 60 f 3 L

E 50 o

[ y 40 8

30

[ N 5.

- 20

- 10 t i I i i I I l l I i

-50 -40 -30 -20 -10 0 10 20 30 40 50 Axial Power Imbalance, %

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[ 8-13 Babcock &Wilcon a McDermott company i

Figure 8-13. Power Imbalance Limits, 30 +10/-0 to 250 t10 EFPD, Oconee 1, Cycle 9 (Technical Specification Figure 3.5.2-3A2)

(-22,102) (27,102) i 100

(-23,92) (28,92) 90 T

(-30,80) 5 - 80 (33,80) g -

OPERATION  % 70 ACCEPTABLE E

.8

" - - 60 Tu a: -

- 50 o

- 40 8

u E -

- 30 b

g -

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-50 -40 -30 -20 -10 0 10 20 30 40 50 Axial Power Imbalance, %

3.5-21a 8-14 Babcock &Wilcox a McDermott company

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Figure 8-14. Power Imbalance Limits After 250 10 EFPD, F Oconee 1, Cycle 9 (Technical Specification L Figure 3.5.2-3A3)

F L (-25,102) (27,102) f -

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(-26,92) (28,92)

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OPERATION I ACCEPTABLE g-- 70 e

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8-18 Bakock &Micou a McDermott company

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.g 5 9. REFERENCES VA .* N

- i'tj A.  :

p.(

1 Oconee Nuclear Station, Units 1, 2, and 3 - Final Safety Analysis Re- l

{.g .;

port, Docket Nos. 50-269, 50-270. and 50-287, Duke Power Company. /T y.g  ?

2 Gadolinia-Bearing Lead Test Assemblies Design Report, BAW-1772-P, I y: ] _

Babcock & Wilcox, Lynchburg, Virginia, June 1983. 965.f . . . . .

3 Mark BZ Dcmonstration Assemblies in Oconee 1, Cycles 7, 8, and 9, C.N..'

.. ~

- BAW-1661, Babcock & Wilcox, Lynchburg, Virginia, March 1981. .)g . (.: .~

4 Rancho Seco Cycle 7 Reload Report -- Volume 1 -- Mark BZ Fuel Assembly j j J . --

, Design Report, BAW-1721P, Babcock & Wilcox, Lynchburg, Vi rgi nia, April

.JV ?; ..

.%# "7 .'

1983. .: .

s,: . 5'-

3 5 BPRA Retainer Design Report, BAW-1496, Babcock & Wilcox, Lynchburg, Vir-  ?.,[- '

ginia, May 1978. .

1 6 J. H. Taylor (B&W) to S. A. Varga (NRC), Letter, "BPRA Retainer Reinser- [h

• tion," January 14, 1980. g.,

7 J. H. Taylor (B&W) to J. F. Stolz (NRC), Letter, "BPRA Retainer Reinser- . ..:

tion," July 13, 1984. /, , 5 4y 2 8 A. F. J. Eckert, H. W. Wilson, and K. E. Yoon, Program to Detennine In- $%

reactor Performance of B&W Fuels - Cladding Creep Collapse, BAW-10084A, T.

k Rev. 2, Babcock & Wilcox, Lynchburg, Virginia, October 1978. pp., ..

. w.; .

9 Y. H. Hsii, et al ., TAC 02 - Fuel Pin Performarice Analysis, BAW-10141PA, f ,A {S Rev.1, Babcock & Wilcox, Lynchburg, Virginia, January 1979. pjg-10 J. H. Taylor to J. S. Berggren, Letter, "B&W's Responses to TAC 02 Ques- T tions, Babcock & Wilcox, April 8,1982.

9j* ~5 . . -

[ 11 H. A. Hassan, W. A. Wittkopf, and W. A. Mullan, B&W Version of PDQ07 + .

Code, BAW-10117A, Babcock & Wilcox, Lynchburg, Virginia, January 1977. 7 3 ;3.%p, 12 J. J. Romano, Core Calculational Techniques and Procedures, BAW-10118A,  ; *1.ik a s .. 4, .c Babcock & Wilcox, Lynchburg, Virginia, December 1979. g.

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7 9-1 Babcock & Wilcox a McDermott company hh b

p=7

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13 M. R. Gudorf, G. E. Hanson, and J. R. Lojek, Assembly Calculations and Fitted Nuclear Data, BAW-10116A, Babcock & Wilcox, Lynchburg, Virginia, )

May 1977.

14 Oconee Unit 1, Cycle 8 Reload Report, BAW-1774, Babcock & Wilcox, Lynch-burg, Virginia, April 1983.

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

16 J. C. Moxley, et al., Fuel Rod Bowing in B&W Fuel Designs, BAW-10147P-A, ]

Rev.1, Babcock & Wilcox, Lynchburg, Virginia, May 1983.

17 R. C. Jones, J. B. Biller, and B. M. Dunn, ECCS Analysis of B&W's 177-FA Lowered-loop NSS, BAW-10103, Rev. 1, Babcock & Wilcox, Lynchburg, Vir-ginia, September 1975. )

18 H. A. Hassan, et al., Power Peaking Nuclear Reliability Factors, BAW-10119, Babcock & Wilccx, Lynchburg, Virginia, January 1977.

19 G. E. Hanson, Nonnal Operating Controls, BAW-10122, Babcock & Wilcox, Lynchburg, Virginia, August 1978.

20 C. W. Mays, Verification of the Three-Dimensional FLAME Code, BAW-10125A, Babcock & Wilcox, Lynchburg, Virginia, August 1976.

21 Bounding Analytical Assessment of NUREG 0630 on LOCA and Operating kW/ft Limits, 77-1141256-00, Babcock & Wilcox, Lynchburg, Virginia, May 1983.

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