Text

Cycle 10 - Reload Rept
	ML20149H145
Person / Time
Site:	Davis Besse
Issue date:	10/31/1994
From:	; BABCOCK & WILCOX CO.
To:	;
Shared Package
ML20149H138	List: ML20149H135; ML20149H140; ML20149H145;
References
	BAW-2223, BAW-2223-R01, BAW-2223-R1, NUDOCS 9411180070
	Download: ML20149H145 (73)
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' BAW-2223. Revision 1 October 1994 1

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DAVIS-BESSE NUCLEAR POWER STATION UNIT 1, CYCLE 10 -- RELOAD REPORT I

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1 I B&W Fuel Company P.O. dox 10935 Lynchburg, Virginia 24506-0935 I

B&W FuelCompany JW 9411180070 941111 PD. t ADOCK 05000346 P PDR

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CONTENTS E.a.jLe,

1. INTRODUCTION AND

SUMMARY

FOR INITIAL SIXTY-FOUR FEED CYCLE 10 .. 1-1 1A. INTRODUCTION AND

SUMMARY

FOR FINAL SIXTY-FOUR FEED CYCLE 10 . ... 1-2 I 2. OPERATING HISTORY FOR INITIAL SIXTY-FOUR FEED CYCLE 10 ...... 2-1 2A. OPERATING HISTORY FOR FINAL SIXTY-FOUR FEED CYCLE 10 . ...... 2-1

3. GENERAL DESCRIPTION FOR INITIAL SIXTY-FOUR FEED CYCLE 10 ..... 3-1 3A. GENERAL DESCRIPTION FOR FINAL SIXTY-FOUR FEED CYCLE 10 ...... 3-2

4. FUEL SYSTEM DESIGN FOR INITIAL SIXTY-FOUR FEED CYCLE 10 ..... 4-1 4.1. Fuel Assembly Mechanical Design . . . . . . . . ...... 4-1 4.2. Fuel Rod Design . . . . . . . . . . . . . . . . ...... 4-1 4.2.1. Cladding Collapse . . . . . . . . . . . ...... 4-1 4.2.2. Cladding Stress . . . . . . . . . . . . ...... 4-2 1- 4.2.3. Cladding Strain . . . . . . . . . . . . ...... 4-3 4.3. Thermal Design . . . . . . . . . . . . . . . . . ...... 4-3 4.4. Material Compatibility . . . . . . . . . . . . . ...... 4-4 I 4.5. Operating Experience . . . . . . . . . . . . . . ...... 4-4 4A. FUEL SYSTEM DESIGN FOR FINAL SIXTY-FOUR FEED CYCLE 10 . ...... 4-5 I 5. NUCLEAR DESIGN FOR INITIAL SIXTY-FOUR FEED CYCLE 10 . . ...... 5-1 5.1. Physics Characteristics . . . . . . . . . . . . ...... 5-1 1 5.2. Changer in Nuclear Design . . . . . . . . . . . ...... 5-1 5A. NUCLEAR DESIGN FOR FINAL SIXTY-FOUR FEED CYCLE 10 . . . ...... 5-2

6. THERMAL-HYDRAULIC DESIGN FOR INITIAL SIXTY-FOUR FEED CYCLE 10 ... 6-1 6A. THERMAL-HYDRAULIC DESIGN FOR FINAL SIXTY-FOUR FEED CYCLE 10 .... 6-2

7. ACCIDENT AND TRANSIENT ANALYSIS FOR INITIAL SIXTY-FOUR FEED CYCLE 10 7-1 7.1. General Safety Analysis . . . . . . . . . . . . ...... 7-1 7.2. Accident Evaluation . . . . . . . . . . . . . . ...... 7-1 I 111 i

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, ', sw Rev. 1 10/94 W CONTENTS (Cont'd)

Pace Il 7A. ACCIDENT AND TRANSIENT ANALYSIS I'OR FINAL SIXTY-FOUR FEED CYCLE 10 7-5

8. PROPOSED MODIFICATIONS TO CORE OPERATING LIMITS REPORT FOR INITIAL SIXTY-FOUR FEED CYCLE 10 . . . . . . . . . . . . . . . . . 8-1 8A. PROPOSED MODIFICATIONS TO CORE OPERATING LIMITS REPORT FOR FINAL SIXTY-FOUR FEED CYCLE 10 . . . . . . . . . . . . . . . . . . 8-3

9. STARTUP PROGRAM - PHYSICS TESTING . . . . . . . . . . . . . . . . 9-1 9.1. Precritical Tests 9.1.1.

Control Rod Trip Test . . . . . . . . . . . . . . .

9-1 9-1 l

5 9.1.2. RC Flow . . . . . . . . . . . . . . . . . . . . . . 9-1 9.2. Zero Power Physics Tests . . . . . . . . . . . . . . . . . . 9-1 9.2.1. Critical Boron Concentration . . . . . . . . . . . . 9-1 9.2.2. Temperature Reactivity Coefficient . . . . . . . . . 9-2 9.2.3. Control Rod Group / Boron Reactivity Worth . . . . . . 9-2 I.

i 9.3. Power Escalation Tests . . . . . . . . . . . . . . . . . . . 9-3 9.3.1 Core Symmetry Test . . . . . . . . . . . . . . . . . 9-3 9.3.2. Core Power Distribution Verification at Intermediate Power Level (IPL) and -100%FP 9-3 gj.

3J-9.3.3. Incore Vs. Excore Detector Imbalance correlation Verification . . . . . . . . . . . . . . 9-4 9.3.4. Hot Full Power All Rod Out Critical Boron Concentration . . . . . . . . . . . . . . . . . . . 9-5 9.4. Procedure for Use if Acceptance / Review Criteria Not Met . . 9-5

10. REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . 10-1 I

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List of Tables Table Pace 4-1. Fuel Design Parameters . . . . . . . . . . . . . . . . ...... 4-6 4-la. Creep Collapse Analysis for Final Sixty-Four Feed Cycle 10 . . . . 4-7 l 5-1. Davis-Besse Unit 1, Cycle 10 Physics Parameters . . . ...... 5-4 l

5-la. Comparison of Selected cycle 10 Physics Parameters . . ...... 5-6 l

I 5-2. Shutdown Margin Calculation for Davis-Besse, Initial Sixty-Four Feed Cycle 10 . . . . . . . . . . . . . . . ...... 5-7 5-2a. Shutdown Margin Calculation for Davis-Besse, Final ,

5-8 I l Sixty-Four Feed Cycle 10 . . . . . . . . . . . . . . . ......

6-1. Maximum Design Conditions, Cycles 9 and 10 . . . . . . ...... 6-3 7-1. Comparison of Key Parameters for Accident Analysis . . ...... 7-6 7-2. Bounding Values for Allowable LOCA Peak Linear Heat Rates . . . . 7-7 8-1. QUADRANT POWER TILT Limits . . . . . . . . . . . . . . . . . . . 8-16 8-2. Negative Moderator Temperature Coefficient Limit . . . . . . . . . 8-16 8-3. Power to Melt Limits . . . . . . . . . . . . . . . . . . . . . . . 8-16

, 8-4. Nuclear Heat Flux Hot Channel Factor - Fn . . . . . . . . . . . . 8-17 1

l 8-5. Nuclear Enthalpy Rise Hot Channel Factor - FL . . . . . . . . . . 8-19 l

I List of Ficures Fiaure Pace l

l 3-1. Davis-Besse Initial Sixty-Four Feed Cycle 10 Core Loading Diagram 3-4 3-la. Davis-Besse Final Sixty-Four Feed Cycle 10 Core Loading Diagram . . 3-5 l

r 3-7. Davis-Besse Batch 12 Zone Loading Pattern . . . . . . . . . . . . 3-6 5 3-3. Davis-Besse Initial Sixty-Four Feed Cycle 10 Enrichment and Burnup Distribution . . . . . . . . . . . . . . . . 3-7 i 3-3a. Davis-Besse Final Sixty-Four Feed Cycle 10 Enrichment and Burnup Distribution . . . . . . . . . . . . . . . . 3-8 3-4. Davis-Besse Cycle 10 Control Rod Locations . . . . . . . . . . . 3-9 3-5. Davis-Besse Cycle 10 BPRA Concentration and Distribution . . . . . 3-10 4-1. Mark-BIO Fuel Assembly Upper End Fitting and Leaf Spring . . . . . 4-8 i v I l B&W FuelCompany

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. E Rev. I 10/94 List of Ficures (Cont'd)

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Pace 1

5-1. Davis-Besse Initial Sixty-Four Feed Cycle 10 Relative Power -

Distribution at BOC (4 EFPD), Full Power, Equilibrium Xenon, Rods at "

Nominal'HFP Positions . . . . . . . . . . . . . . . . . . . . . . 5-9 5-la. Davis-Besse Final Sixty-Four Feed Cfcle 10 Relative Power Distribution at BOC (4 EFPD), Full Power, Equilibrium Xenon, Rods at Nominal HFP Positions . . . . . . . . . . . . . . . . . . . . . . 5-10 8-1. Regulating Group Position Alarm Setpoints, O to 25 +10/-O EFPD, l Four RC Pumps -- Davis-Besse 1, Cycle 10 . . . . . . . . . . . . . 8-4 8-2. Regulating Group Position Alarm Setpoints, 25 +10/-O to 460 110 EFPU. g Four RC Pumps -- Davis-Besse 1, Cycle 10 . . . . . . . . . . . . . 8-5 g 8-3. Regulating Group Position Alarm Setpoints, After 460 110 EFPD, l Four RC Pumps -- Davis-Besse 1, Cycle 10 . . . . . . . . . . . . . 8-6 8-4. Regulating Group Position Alarm Setpoints, O to 25 +10/-O EFPD, l B Three RC Pumps -- Davis-Besse 1, Cycle 10 . . . . . . . . . . . . 8-7 g

8-5. Regulating Group Position Alarm Setpoints, 25 +10/-O to 460 110 EFPD, i Three RC Pumps -- Davis-Besse 1, Cycle 10 . . . . . . . . . . . . . 8-8 Id 8-6. Regulating Group Position Alarm Setpoints, After 460 110 EFPD, l Three RC Pumps -- Davis-Besse 1, Cycle 10 . . . . . . . . . . . . 8-9 8-7. Control Rod Core Locations and Group Assignments for Davis-Besse 1, Cycle 10 . . . . . . . . . . . . . . . . . . . 8-10 8-8. APSR Position Alarm Setpoints . . . . . . . . . . . . . . . . . . 8-11 3 8-9. AXIAL POWER IMBALANCE Alarm Setpoints, O EFPD to EOC, Four RC Pumps -- Davie-Besse 1, Cycle 10 . . . . . . . . . . . . . 8-12 8-10. AXIAL POWER IMBALANCE Alarm Setpoints, O EFPD to EOC, Three RC Pumps -- Davis-Besse 1, Cycle 10 . . . . . . . . . . . 8-13 8-11. AXIAL POWER IMBALANCE Protective Limits . . . . . . . . . . . . . 8-14 8-12. Flux--AFlux/ Flow (or Power / Imbalance / Flow)

Trip Setpoints . . . . . . . . . . . . . . . . . . . . . . . . . . 8-15 8-13. Allowable Radial Peak for FL . . . . . . . . . . . . . . . . . . 8-19 "-

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1. INTRODUCTION AND

SUMMARY

FOR INITIAL SIXTY-FOUR FEED CYCLE 10 The original version of BAW-2223' was completed in July 1994. Since the i completion of that report, Toledo Edison has developed a revised fuel cycle l

l design for cycle 10. The new design provides ine'reased cycle lifetime and

= utilizes eight additional fresh fuel assemblies compared to the original cycle 10 design. This report describes the analyses and results of the redesigned g cycle 10.

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This report justifies operation of Davis-Besse Nuclear Power Station Unit 1 at

5 the rated core power of 2772 MWt for cycle 10. The required analyses are l

l included as outlined in the Nuclear Regulatory Commission (NRC) document,

" Guidance for Proposed License Amendments Relating to Refueling," June 1975.

l This report utilizes the analytical techniques and design bases that have been submitted to the NRC and approved by that agency.

l Cycle 10 reactor and fuel parameters related to power capability are summarized in this report and compared to those for cycle 9. All accidents analyzed in the j Davis-Besse Updated Safety Analysis Report2 (USAR), as applicable, have been reviewed for cycle 10 operation, and in all cases, the initial conditions of the transients in cycle 10 are bounded by previous analyses.

t Fuel assembly NJ053L, which will be reinserted from cycle 7, was recaged in a l Mark-B7 structural cage and now contains two stainless steel replacement rods, t

Recaged assembly NJ053L is now designated NJORAA and will receive its second W cycle of irradiation during cycle 10. The ef fect of the replacement rods on core performance is discussed in the applicable sections.

The Technical Specifications have been reviewed for cycle 10 operation. Based l on the reload report analyses performed, taking into account the emergency core B cooling system (ECCS) Final Acceptance Criteria and postulated fuel densification effects, it is concluded that Davis-Besse Unit 1, cycle 10 can be operated safely at its licensed core power level of 2772 MWt. The Core Operating Limits Report (COLR) changes for cycle 10 are included in section 8 of this report. *,

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E lA. INTRODUCTION AND

SUMMARY

FOR FINAL SIXTY-FOUR FEED CYCLE 10 During the cycle 9 refueling outage at Davis-Besse Unit 1, all fuel assemblies =

1 from cycle 9 were ultrasonically tested for leaking fuel rods. Three fuel

assemblies were identified as having leaking fuel rods. Two of these assemblies ,

were scheduled for reinsertion in the initial sixty-four feed cycle 10 desi.sn 3J' discussed in section 1 above. These two assemblies were not reconstituted or recaged and were discharged from the final cycle 10 core configuration. A fuel cycle design was completed for this revised configuration which is identified as the final sixty-four feed cycle 10 design. This report now also describes the evaluations that were performed to confirm that the final sixty-four feed cycle 10 design is conservatively bounded by the analyses that were completed for the initial sixty-four feed cycle 10 design. E I

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2. OPERATING HISTORY FOR INITIAL SIXTY-FOUR FEED CYCLE 10 I The reference cycle for the nuclear and thermal-hydraulic analyses of Davis-Besse Unit l is cycle 9, which achieved criticality on April 28, 1993 and shut down on October 1,1994 after achieving 500.79 EFPD. Power escalation began on April 30, 1993 and full power (2772 MWt) was attained on May 5, 1993.

During cycle 9 operation, no operating anomalies occurred that would adversely af fect fuel performance during cycle 10. Cycle 10 was analyzed to 520 ef fective !

full power days (EFPD) based on cycle 9 operation of 501 EFPD. The cycle 10 specific core protective limits and corresponding reactor protection system (RPS) setpoints for Flux--AFlux/ Flow or Power / Imbalance / Flow have been determined, and are applicable to 520 EFPD. All other cycle 9 RPS setpoints (reference 3) have been determined to be valid for cycle 10. The cycle 10 operating limits have

1. also been verified to 520 EFPD. The cycle 10 design includes an APSR pull and i

power coastdown.

The cycle 10 design minimizes the number of fuel assemblies that are cross core shuf fled to reduce the potential for quadrant tilt amplification. The cycle 10 shuffle pattern is discussed in section 3.

2A. OPERATING HISTORY FOR FINAL SIXTY-FOUR FEED CYCLE 10 The reference cycle for the verification of the nuclear and thermal-hydraulic analyses of the final sixty-four feed cycle 10 design remains cycle 9. However, selected parameters recalculated for the final sixty-four feed core design are ,

5 shown in comparison to their values for the initial sixty-four feed core design l where appropriate. The revised cycle 10 shuffle pattern is discussed in section 3A.

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3. GENERAL DESCRIPTION FOR INITIAL SIXTY-FOUR FEED CYCLE 10 The cycle 10 core consists of 177 fuel assemblies (FAs), each of which is a 15x15 array normally containing 208 fuel rode, 16 control rod guide tubes, and one incore instrument guide tube. All FAs in batches 9 and 10 have a constant nominal fuel loading of 468.25 kg of uranium. The batch 11 and 12 FAs have a constant nominal fuel loading of 468.56 kg of uranium. The fuel consists of dished-end cylindrical pellets of uranium dioxide clad in cold-worked Zircaloy-4.

I The undensified nominal active fuel lengths, theoretical densities, fuel and fuel rod dimensions, and other related fuel parameters may be found in Table 4-1 of this report.

Figure 3 1 is the core loading diagram for Davis-Besse Unit 1, cycle 10. Forty-eight batch 9B assemblies, 20 batch 10A assemblies, and 1 batch 8C assembly will l be discharged at the end of cycle 9. The remaining Satch a 10B and batch 11 FAs will be shuffled to their cycle 10 locations, with the core periphery locations occupied by batch 10B, batch 11, and batch 12 fuel assemblies. One batch 9D l assembly, discharged at the end of cycle 8, will be reinserted in cycle 10 as the center FA. Four batch 9A assemblies, discharged at the end of cycle 7, will be reinserted in cycle 10. One of these assemblies has been recaged with two I stainless steel replacement rods, and is designated NJORAA. This assembly will be loaded at core location G08, as shown in Figure 3-1. Batches 9A and 9D,10B, and 11 have initial enrichments of 3.38, 3. 69, and 3.77 wt %, respectively. The feed batch, consisting of 64 radially zone-loaded batch 12 assemblies withl ;

uranium enrichment of 4.06 wt % (average), will be inserted in the core interior in a symmetric checkerboard pattern. Implementation of radially zone-loaded enrichments within the fuel assembly results in a reduction in the radial peak pin power relative to an assembly with a uniform enrichment loading. Selected fuel rod locations within the assembly contain reduced enrichment fuel to obtain 5 this effect. The zone loading of the fuel rods in the batch 12 assemblies is illustrated in Figure 3-2. The cycle 10 shuf fle scheme is a very low leakager 3-1 i

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. E Rev. 1 10/94 (VLL) core loading. The VLL reload fuel shuffle scheme for cycle 10 will have a negligible effect on Nuclear Anstrumentation response for all aspects of reactor startup and subsequent power operation. Figure 3-3 is a quarter-core map showing each assembly's burnup at the beginning of cycle (BOC) 10 and its initial .

enrichment.

Cycle 10 is operated in a feed-and-bleed mode. The core reactivity is controlled by 53 full-length Ag-In-Cd control rod assemblies (CRAs), 48 burnable poison rod assemblies (BPRAs), and soluble boron. Eight of the BPRAs will be reinserted from cycle 5. Twelve standard control rods will be replaced in cycle 10 with 3 Extended Life Control Rods (ELCRAs) that are discussed in section 4.1. In addition to the full-length control rode, eight Inconel-600 axial power shaping rods (gray APSRs) are provided for additional control of the axial power distribution. The cycle 10 locations of the control rods and the group designations are indicated in Figure 3-4. The core locations and the rod group designations of the 61 control rods in cycles 9 and 10 are the same. The cycle 10 locations and concentrations of the BPRAs are shown in Figure 3-5. '

3A. GENERAL DESCRIPTION FOR FINAL SIXTY-FOUR FEED CYCLE 10 At the completion of cycle 9 operation, each individual fuel assembly was ultrasonically inspected for leaking fuel rods. As a result of this complete core inspection, three fuel assemblies were identified as containing one leaking fuel rod each. Of these"three assemblies, two were planned for reinsertion in cycle 10: one batch 10B assembly and one batch 11 assembly. In the initial sixty-four feed cycle 10 design, 64 batch 10 assemblies had been subdivided into B 20 batch 10A assemblies to be discharged at EOC-9 and 44 batch 10B assemblies scheduled for use in cycle 10. In the final sixty-four feed cycle 10 design,16 _

batch 10A assemblies have been discharged, while 48 batch 10B assemblies remain ,

in cycle 10. The batch 10 assembly with a leaking fuel rod and its three symmetric counterparts have been discharged as batch 10A assemblies. The batch 11 assembly with a leaking fuel rod and its three symmetric counterparts have .3 been discharged as batch 11A. The remaining 60 batch 11 assemblies are designated as batch 11B for cycle 10.

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Beyond these assembly substitutions, the final core design includes assembly shuffles that were made to minimize the impact on the initial sixty-four feed l design. Additional analyses have been performed on this final sixty-four feed design to assess its impact compared to the initial design. These discussions may be found as individual subsections in each section that follows.

Figure 3-la is the final sixty-four feed core loading diagram. A total of 16 full core locations were affected by this change, and as a result four fewer l

l3 batch 10A assemblies and four batch llA assemblies were discharged compared to

!g the initial sixty-four feed core loading diagram. The final cycle 10 shuffle i scheme is a very low leakage (VLL) core loading and will have a negligible effect on Nuclear Instrumentation response for all aspects of reactor startup and

( subsequent power operation. Figure 3-3a reflects these changes in the assembly burnup and initial enrichment distributions in cycle 10. No changes in control rod locations, control rod group designation, BPRA locations, or BPRA t

Concentrations were made in the final design.

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I.. Figure 3-la. Davis-Bains Final Sixty-Four Facd Cycle 10 Core Loading Diagram

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l 10B 108 10B 10B 10B A G13 P10 P08 PO6 G03 108 12 11B 12 10B 12 11B 12 10B B N12 F ID3 F P11 F L13 F ID3 108 12 11B 12 11B 12 11B 12 11B 12 10B C M14 F K02 F E12 F ED4 F K14 F 105 10B 12 11B 12 11B 12 1 12 11B 12 11B 12 10B D 005 F KO6 F M38 F K1 F K12 F PO7 F NO4 5 E 12 F

11B B09 12 F

11B L11 11B NO3 108 107 F 12 10B 109 11B N13 11B 106 12 F

11B B07 J

F I F 10B 007 108 1 11 C10 12 12 F

11B 11B D09 12 11B C12 10B 11B H13 12 12 F

11B 10B KOS 9A F

12 11B 11B 008 12 11B CD4 10B llB 105 12 12 F

11B 11B CD6 12 10B 009 10B G L14 F 105 F G01 F M10 HD4 IDS F G15 F N11 F ID2 HW- F N N D0h f0 F F -Y 7 8 7 108 12 11B 12 IOD 12 11B 9A 11B 12 IOB 12 11B 12 10B K F14 F DOS F ED1 F Fil E12 ED6 F K15 F D11 F F02 7

L 108 (D7 11B 010 12 F

11B H11 llB 012 11B (D8 12 F

10B 008 F 12 11B 103 11B 004 11B N07 12 F g 10B C09 12 11B 12 11B 11B 108 12 10B 115 11B 12 11B 12 M F P09 F E10 D03 107 F FO9 D13 FOS F PO7 F 10B 12 11B 12 11B 12 11B 12 31B 12 11B 12 10B N D12 F ID9 F GO4 F G06 F ID8 F G10 F C11 10B 12 118 12 11B 12 11B 12 11B 12 10B

60,000 >60,000 >60,000 >60,000 >55,000 -

Maximum pin burnup, mwd /mtU 36,520 49,075 53,842 45,963 25,676 El:

ge Nominal linear j; heat rate at 2772 MWt, kW/ft 6.14 6.14 6.14 6.25 6.25 Minimum linear heat rate to l um melt, kW/ft 20.5 20.5 20.5 22.1 22.3 l Calculated using method from reference 5.

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Table 4-$a. Creen Collapse Analysis for Final S$xtv-Four Feed Cvele 10 I Creep Collapse Burnup EOC 10 (520 EFPD)
Max. FA Burnup BATCH No. of Assemblies (mwd /mtU) (mwd /mtU) 9A 4 >60,000 34,251 I
l 9D 1 >60,000 47,370 I 10B 48 >60,000 50,047 5 11B 60 >56,700 43,835 E 12 64 >50,970 24,452 I
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5. NUCLEAR DESIGN FOR INITIAL SIXTY-FOUR FEED CYCLE 10 5.1. Physics Characteristics Table 5-1 compares the core physics parameters for the cycle 9 and 10 designs.
I The values for cycles 9 and 10 were generated with the NEMO code'.
in core physics parameters are to be expected between the cycles due to the Differences changes in fuel and burnable poison concentrations which create changes in flux and burnup distributions. Figure 5-1 illustrates a representative relative power distribution for BOC-10 at full power with equilibrium xenon, group 7 inserted to nominal HFP position, and gray APSRs partially inserted.
The ejected rod worths in Table 5-1 are the maximum calculated values.
Calculated ejected rod worths and their adherence to criteria are considered at l all times in life and at sll power levelt in the development of the rod position limite presented in section 8. The adequacy of the shutdown margin with cycle 10 rod worths is shown in Table 5-2. The following conservatisms were applied i for the shutdown calculations: l
.I 1. Poison material depletion allowance.
2. 10% uncertainty on net rod worth.
l= 3. Xenon transient allowance.
4. A maximum power deficit.
The xenon transient allowance was taken into account to ensure that the ef fects 1
of operational maneuvering transients were included in the shutdown analysis.
5.2. Chances in Nuclear Desian l
The design changes for cycle 10 include increased enrichments, Mark-B10AZL assembly design, and radially zone-loaded fuel (see Figure 3-2). These changes {
I are incorporated in the physics model. The impact of the two stainless steel rods was also evaluated and determined not to significantly impact core reactivity, stuck rod worth, or ejected rod worth.
Twelve standard control rods will be replaced by Extended Life Control Rods (ELCRAs). The new control rods consist of the same poison material, but have a 5-1 B&W FuelCompany
EI Rev. 1 10/94 lI 5
smaller poison diameter and a longer poison length. The cladding material of the .l standard rods is stainless steel 304. Inconel 625 is used to clad the ELCRAs.
Although the resulting change in rod worth is small, these changes are incorporated into the analysis.
No significant operational or procedural changes exist with regard to axial or radial power shape, xenon, or tilt control. The stability and control of the core with APSRs withdrawn has been analyzed. The calculated stability index without APSRs is -0.0260 h , which demonstrates the axial stability of the core. l d
The operating limits (COLR changes) for the reload cycle are given in section 8.
5A. NUCLEAR DESIGN FOR FINAL SIXTY-FOUR FEED CYCLE 10 since the final design introduces only minor modifications to the initial core design, no significant changes were expected in the physics parameters predicted for the initial sixty-four feed cycle 10 design. This was demonstrated in the -
evaluation by recalculating a limited number of important physics parameters (see ,
Table 5-la). These parameters represent those which are likely to be sensitive to the core changes or may affect core operating limits. The balance of key physics parameters are qualitatively evaluated based on these calculations.
The largest changes in core response were the decreases in ejected rod worth and stuck rod worth at BOC. However, a decreasing value for these parameters is less limiting and hence, the initial sixty-four feed design values were bounding. The rod worths were also affected by the final design changes. Referring to Table 5-la, the HFP group 6 worth increased by 0.03 %Ak/k (+3.6%), while group 7 5 decreased by 0.02 %Ak/k (-1.8%). These magnitudes are consistent with changes introduced with the final design where assembly shuffles have occurred at or near the corresponding core locations with regulating control rods. With exposure, the magnitude of these changes decreased. The overall impact on total rod worth was limited to a maximum increase of 0.03 %Ak/k.
The balance of the physics parameters are virtually unchanged. The critical boron concentrations did not change at BOC. At other times in life, increases of 1 or 2 ppm were observed. These increases are insignificant changes. The Doppler temperature coefficient remains the same. The moderator temperature coef ficient becomes more positive by 0.01 x 10-2 % Ak/k/*F at BOC, while remaining 5-2 B&W FuelCompany g
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unchanged at EOC. The more positive BOC moderator coef ficient is due to the decrease in core leakage, but is well within its allowable limit. The EOC steam line break temperature coefficient is unchanged. An updated BOC power
[5 distribution may be found in Figure 5-la. The updated stability index without APSRs is -0.0261 hr d.
2 In addition to these parameters, the shutdown margin calculation was updated as i
shown in Table 5-2a. The final sixty-four feed design increases the shutdown margin by 0.15 %Ak/k at Boc and 0.03 %Ak/k at EoC. These increases illustrate I that the shutdown margin analysis for the initial design is bounding.
l These comparisons indicate that the perturbations caused by the final sixty-four feed design do not significantly alter the core neutronic behavior. For many of l the important parameters, the p. vsics data are the same as those for the initial design. Further, all differences are minor and inconsequential to the core response as shown in Tables 5-la and 5-2a. Therefore, the final sixty-four feed design does not change any conclusions of the initial nuclear analysis.
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Rev. 1 10/94 73 ble 5-1. Davis-Besse Unit 1. Cycle 10 Physics Parameters Cycle 9 Cycle 10")
Cycle length, EFPD 500 520 Cycle ournup, mwd /mtU 16,719 17,383 Average core burnup - 520 EFPD S), HWd/mtU 31,910 34,408 Initial core loading, mtU 82.9 82.9 Critical boron") - 0 EFPD, No Xe, ppm HZP 1,840 1,973 HFP 1,646 1,754 Critical boron") - 520 EFPD*), 100%FP Xe, ppm HZP 239 196 HFP W S) 5 ")
Control rod worths - HFP, 4 EFPD, %Ak/k Group 6 1.06 0.84 Group 7 1.06 1.14 Group 8 0.15 0.14 Control rod worths - HFP, 520 EFPD*), %Ak/k -
Group 7 1.14 1.15 Group 8 NA NA Max ejected rod worth - HZP, %Ak/k 0 EFPD, Group 5 at 15%WD, 0.55 0.74 Groups 6-8 inserted") (N-12) 520 EFPD S), Groups 5-7 inserted (N-12) 0.49 0.56 W l Hax stuck rod worth - HZP, %Ak/k g 1
0 EFPD (N-12) 0.72 0.78 g l
520 EFPD*~ '
(N-12, Cy9; H-13, Cy10) 0.66 0.70 Power deficit * - HZP to HFP, 100%FP Xe, %Ak/k l 4 EFPD -1.80 -1.87 ~
S 520 EFPD) -3.37 -3.16 Doppler coeff* - HFP, 10 4% dk/k/'F '
O EFPD, No Xe* -1.59 -1.55 520 EFPD S), 100%FP Xe, O ppm *) -1.90 -1.70 Moderator coeff* - HFP, 10-2 %Ak/k/'F 1 0 EFPD, No XeW -0.63 -0.(1 l 520 EFPD S), 100%FP Xe, O ppm *) -3.43 -3.33 Temperature coeff* - HZP, 10-2 % Ak/k/*F l 520 EFFD*), 100%FP Xe, Grps 1-7 In, M13 Out*, O ppm -2.61 -2.70 l 5-4 B&W FuelCompany
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i Rev. 1
3 l 10/94 l Table 5-1. Davis-Besse Unit 1. Cycle 10 Physics Parameters l l
l Boron worth * - HFP, ppm /%Ak/k 0 EFPD 142 150 l 520 EFPD*), 100%FP Xe 115 118 f l
Xenon worth" - HFP, %Ak/k l 4 EFPD 2.62 2.60 520 EFPD), S 100%FP Xe 2.78 2.77 i
Effective delayed neutron fraction" - HFP l 4 EFPD 0.00632 0.00628 l lI l
520 EFPD S', 100%FP Xe 0.00526 0.00527
") Based on cycle 8 length of 458.9 EFPD (actual) and cycle 9 length of 501 EFPD.
) Calculated at 500 EFPD for cycle 9. f
"' Control rod group 8 is inserted for calculation at 0 EFPD and withdrawn for calculation at 520 EFPD. j
Power coastdown to 500 EFPD at 5 ppm for cycle 9 and to 520 EFPD at 5 ppm for F cycle 10. l l
") Cycle 9 value calculated with rod groups 5-7 inserted. (
l l m All calculations done with control rod groupe 1-7 at 100%WD and control rod l
group 8 at nominal HFP position, unless otherwise noted.
I
Cycle 10 values were calculated at 1831 ppm (includes allowances for B10 atom l fraction variation and reactivity anomalies); cycle 9 values were calculated at 1666 ppm (estimated prior to final determination of reactivity bias and j
{
i 1
does not include allowances).
(
These values were calculated with the control rods at rod index 260%WD.
Cycle 10 stuck rod location was M13; cycle 9 stuck rod location was N12. l r
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I 5-5 B&W FuelCompany
Table 5-la. Comparison of Selected Cycle 10 Physics Parameters Initial Final
,Cvele 10 pycle 10* "
Cycle length, EFPD 520 520 Cycle burnup, mwd /mtU 17,383 17,384
]
Average core burnup - 520 EFPD, mwd /mtU 34,408 34,491 Initial core loading, mtU 82.9 82.9 Critical boron - 0 EFPD, No Xe, ppm HFP 1,754 1,754 l
3 Control rod worths - HFP, 4 EFPD, %Ak/k Group 6 0.84 0.87 Group 7 1.14 1.12 Group 8 0.14 0.14 Max ejected rod worth - HZP, %Ak/k 0 EFPD, Groups 5-8 inserted (N-12) 0.85 0.76 Max stuck rod worth - HZP, %Ak/k 0 EFPD (N-12) 0.78 0.66 520 EFPD (M-13) 0.70 0.69 ,
i Power deficit *' - HZP to HFP, 200%FP Xe, %Ak/k 2 j 4 EFPD -1.87 -1.86 i 520 EFPD -3.16 -3.16 l
l Doppler coeff*' - HFP, 10 % Ak/k/*F I O EFPD, No Xe* -1.55 -1.55 g 520 EFPD, 100%FP Xe, O ppm * -1.70 -1.70 g Moderator coeff*' - HFP, 10-2 %Ak/k/'F 0 EFPD, No Xe* -0.61 -0.60 520 EFPD, 100%FP Xe, O ppm * -3.38 -3.38 l
m Temperature coeff - HZP, 102 %Ak/k/ F 520 EFPD, 100%FP Xe, Grps 1-7 In, M13 Out, O ppm -2.70 -2.70 l '
Xenon worth *' - HFP, %Ak/k 4 EFPD 2.60 2.60 l
M Based on cycle 8 length of 458.9 EFPD (actual) and cycle 9 length of 501 EFPD.
S' All calculations done with rod groups 1-7 at 100%WD and APSR at nominal HFP position, unless otherwise noted.
Cycle 10 values were calculated at 1831 ppm (includes allowances for B oi atom fraction variation and reactivity anomalies).
l
These values were calculated with the control rods at rod index 260%WD. g B.
5-6 B&W FuelCompany H
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l f Rev. 1 10/94 l
I Table 5-2. Snutdown Margin Calculation for Davis-Besse, Initial Slxtv-Four Feed Cvele 10 I
}
BOC
%Ak/h 470 EFPD Group 8 in EOC. %Ak/k 520 EFPD Grouc 8 out l
Available Rod Worth l
Total rod worth, HZP 6.33 7.00 7.02 l Worth reduction due to burnup of poison material -0.09 -0.10 -0.11 i Maximum stuck rod worth, HZP -Q,78 -0.69 -0.70 l Net Worth 5.46 6.21 6.21 I
{
Less 10% Uncertainty Total available worth
-0.55 4.91
-0.62 5.59
-0.62 5.59 l Recuired Rod Worth Power deficit, HFP to HZP 1.87 3.13 3.16 l Xenon transient allowance 0.30 0.30 0.30 Max allowable inserted rod worth 0.,36 0.54 0.,55 Total required worth 2,53 3.97 4.01 l l Shutdown Marcin j I Total available minus total required 2.38 1.62 1.58 l ILqt e_: Required shutdown margin is 1.00% Ak/k.
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Table 5-2a. Shutdown Margin calculation for Davis-Besse, g Final Sixtv-Four Feed Cvele 10 g 1
EOC. %6k/k BOC, 470 EFPD I'd, 520 EFPD
%6k/k Grouc 8 in Groue 8 cut Available Rod Worth Total rod worth, HZP 6.36 7.03 7.04 Worth reduction due to burnup of poison material -0.09 -0.10 -0.11 Maximum stuck rod worth, HZP 71 66 -0.67 -0.69 Net Worth 5.61 6.26 6.24 Lens 10% Uncertainty -0.56 -0.63 -0.62 Total available worth 5.05 5.63 5.62 Recuired Rod Worth Power deficit, HFP to HZP 1.86 3.13 3.16 Xenon transient allowance 0.30 0.50 0.30 Max allowable inserted rod worth 0.36 0.54 0.55 l Total required worth 2.52 3.97 4.01 Shutdown Marcin Total available minus I total required 2.53 1.66 1.61 '
Notes Required shutdown margin is 1.00% Ak/k.
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R3v. 1 This Figure Superseded by Figure 5-1a
Figure 5-1. Davis-Besse Initial Sixty-Four Feed Cycle 10 Relative Power Distribution at BOC (4 EFPD), Full Power, Equilibrium Xenon, Rods at Nominal HFP Positions l I 8 9 10 11 12 13 14 15 7
H 0.625 0.829 0.861 1.336 1.240 1.303 0.725 0.295 I K 0.829 0.943 1.196 1.039 1.362 1.192 1.163 0.329 8
L 0.861 1.19- 1.121 1.184 1.163 1.308 0.862 0.244 M 1.336 1.037 1.184 1.214 1.386 1.202 0.950 7 8 7 N 1.240 1.363 1.165 1.387 1.234 1.236 0.405 0 1.303 1.188 1.306 1.202 1.237 0.577 I P 0.404 0.725 1.153 0.859 0.949 I
y -
0. m e. m e.2 3
..... . 1.ti.. . .. O... t, 5-9 I
B&W FuelCompany
Figure 5-la. Davis-Besse Final Sixty-Four Feed Cycle 10 Relative Power Distribution at BOC (4 EFPD), Full Power, ii
'i Equilibrium Xenon, Rods at Nominal HFP Positions 8 9 10 11 12 13 14 15 7 1 H 0.625 0.830 0.862 1.341 1.254 1.338 0.801 0.312 K O.830 0.944 1.198 1.042 1.372 1.213 1.199 0.341 8
L O.862 1.194 1.120 1.182 1.163 1.313 0.869 0.248 I
M 1.180 1.205 1.370 1.185 0.940 1.341 1.039 7 8 7 N 1.254 1.372 1.158 1.370 1.203 1.188 0.390 O 1.338 1.209 1.309 1.187 1.194 0.479 I-P 0.801 1.191 0.868 0.942 0.394 I
R O.312 0.339 0.247 I
x Inserted Rod Group Number x.xxx Relative Power Density .
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R2v. 1 10/94 i
6.0 THERMAL-HYDRAULIC DESIGN FOR INITIAL SIXTY-FOUR FEED CYCLE 10 E
The thermal-hydraulic design evaluation supporting cycle 10 operation utilized the methods and models described in reference 9. Cycle 10 is the first Davis-Besse fuel cycle to implement Statistical Core Design (SCD) methodology (reference 10). This thermal-hydraulic analysis technique provides additional DNBR margin by statistical combination of core and fuel parameter uncertainties.
l l The Davis-Besse cycle 10 SCD-based DNB design criterion was a 1.40 Thermal Design
!W Limit (TDL), based on the BWC CHF correlation (reference 11). With the implementation of the batch 12 Mark-B10AZL fuel, the cycle 10 core is a full Zircaloy-grid core. The BWC CHF correlation is, therefore, applicable for the entire core. Utilization of the SCD methodology supports an increase in design (radial x local) peaking factor from 1.714 to 1.795.
l The incoming batch 12 fuel, designed with zone-loaded fuel and a cruciform leaf type holddown spring, is hydraulically and geometrically similar to the fuel remaining in the core from previous cycles. The cycle 10 core contains 48 BPRAs and 68 open fuel assemblies with unplugged control rod guide tubes. The sixteen l Mark-B10AZL open fuel assemblies have optimized control rod guide tubes which i
lower the core bypass flow rate. The resulting core bypass flow for this a configuration is 6.4%. The analysis basis for the cycle 10 SCD-based core model l was a full Mark-B9/ Mark-B10 core with a core bypass flow rate of 5.3%. The 1.40 TDL provides 6.6% retained DNBR margin relative to the 1.313 Statistical Design
[ Limit (SDL). Application of the core model to the cycle 10 configuration required the use of 2.9% retained DNBR margin to offset the impact of the slightly higher cycle 10 core bypass flowrate.
Analyses for the limiting DNB transients (one-pump coastdown, four pump coastdown, and locked rotor events) were re-evaluated with SCD methodology to justify the increased design (radial x local) peaking f actor. The three-to-two pump locked rotor remains the limiting loss of flow event. Steady-state DNB analyses were performed to define core protective limits. No rod bow penalty or
( 6-1 B&W FuelCompany g
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E=
Rev. 1 10/94 densification power spike factors were considered for DNB analyses in the cycle 10 core (as discussed in reference 9). The batch 9A recaged fuel assembly containing two stainless steel replacement rods was shown to be acceptable, based on the methodology described in reference 12.
Table 6-1 provides a summary comparison of the DNB analysis parameters for cycles 9 and 10.
6A. THERMAL-IIYDRAULIC DESIGN FOR FINAL SIXTY-FOUR FEED CYCLE 10 The final cycle 10 core configuration bypass flowrate remains unchanged from that of the initial configuration. The core flowrate for the final cycle 10 configuration is bounded by the core flowrate used in the generic DNB analyses.
Therefore, the DNB results generated in the Statistical Core Design (SCD)-based DNB analyses for the initial cycle 10 configuration remain valid and applicable for the final cycle design.
The batch 9A recaged fuel assembly containing two stainless steel replacement rods wus confirmed to remain accept *.,le, based on the methodology described in BAW-2149A.
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E.. Rnv. 1 Table 6-1. Maximum Desian conditions. Cveles 9 and 10 cvele 9 cvele 10 Design power level, MWt 2772 2772 Nominal core exit pressure, psia 2200 2200 Minimum core exit pressure, psia 2135 2135 Reactor coolant flow, gpm 380,000 380,000 Core bypass flow, % 8. 4 ") 5.3 *)
DNBR modeling Crossflow Crossflow w/SCD Reference design (radial x local) power peaking factor I 1.71 1.795 Reference design axial flux shape 1.65 chopped 1.65 chopped cosine cosine Hot channel factors Enthalpy rise 1.011 1.011 1 Heat flux Flow area 1.014 0.97 N/A")
0.97 Active fuel length, in. 140.6 140.6 Avg heat flux at 100% power, 10 Btu /h-ft 5 2 1.89 1.89 Max heat flux at 100% power, 10 Btu /h-ft:
5 5.35 5.60 CHF correlation BWC BWC CHF correlation DNB limit 1.18 1.40 TDL Minimum DNBR I at 102% power at 112% power 1.78 1.55 2.02 1.79
") Calculated for the actual cycle 9 core configuration.
) Used in the analysis.
)
The hot channel factor for heat flux is no longer applied in DNB I calculations as discussed in reference 9.
6-3 B&W FuelCompany
I.. R3v. 1 ,
10/94 I
7. ACCIDENT AND TRANSIENT ANALYSIS FOR INITIAL SIXTY-FOUR FEED CYCLE 10 7.1 General Safety Analysis Each USAR accident analysis has been examined with respect to changes in the cycle 10 parameters to determine the effects of the cycle 10 reload and to ensure that thermal performance during hypothetical transients is not degraded. The effects of fuel densification on the USAR accident results have been evaluated I and are reported in reference 13.
I The radiological dose consequences of the USAR Chapter 15 accidents have been evaluated using conservative radionuclide source terms that bound the cycle specific source terms for Davis-Besse cycle 10. The dose calculations were performed consistent with the assumptions described in the Davis-Besse USAR but used the more conservative source terms (which bound future reload cycles) . The results of the dose evaluations showed that of fsite radiological doses for each accident were below the respective acceptance criteria values in the current NRC Standard Review Plan (NUREG-0800).
The ef fects of inadvertent loading of a fuel assembly into an improper position have been evaluated. This type of misplacement would be detected with the incore detectors during startup tests.
7.2 Accident Evaluation The key parameters that have the greatest effect on determining the outcome of a transient can typically be classified in three major areas: (1) core thermal, (2) thermal-hydraulic, and (3) kinetics parameters, including the reactivity feedback coefficients and control rod worths.
Fuel thermal analysis parameters from each batch in cycle 10 are given in Table 4-1. The cycle 10 thermal-hydraulic maximum design conditions are presented in Table 6-1. A comparison of the key kinetics parameters from the USAR and cycle 10 is provided in Table 7-1.
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E Rev. 1 3 10/94 gl The EOC moderator temperature coefficient listed in Table 7-1 for cycle 10 is the 3-D, hot full power (HFP) temperature coefficient. An evaluation was performed to verify the acceptability of the more negative cycle 10 moderator temperature coefficient for all USAR accidents excluding steam line breaks. The results of the evaluation were acceptable for all USAR accidents, excluding steam line breaks, for a moderator temperature coefficient as negative as -4.0 x 102
% Ak/k/*F. M The steam line break accident was evaluated based on the total reactivity change from 532*F to the minimum temperature reached during the event ( 510*F) . The '
temperature coefficient used in safety analysis of the steam line break is
-3.10 x 10-2 % A k/ k/*F. This value is based on the sum of the moderator density, control rod worth degradation and Doppler reactivity, over the temperature range f rom 532*F to slo *F. The combined temperature coef ficient for EOC-10 is shown in section 5, and in Table 7-1, as -2.70 x 10-2 % A k/ k/*F. Since the safety analysis l value for the EOC temperature coefficient is more negative than the cycle 10 ;
value, the steam line break analysis remains bounding for cycle 10.
With the implementation of SCD, the design (radial x local) peak was increased l to 1.795. A review was performed of the non-LOCA safety analyses and the LOCA analyses to determine the effect of the increased radial x local peaking on the results of those analyses. Most of the safety analyses were performed with a W systems thermal-hydraulic computer code to determine the primary and/or secondary system (s) response (s). Because the core was modeled as a lumped average channel, l changes in radial and/or axial peaking factors will have no effect on those calculated system responses (i.e. , pressures, temperatures, mass flow rates, core average power). Since the predictions of the primary and secondary system responses and the average core power are unaffected by the change in radial x local peaking factor, the Chapter 15 events can be divided into four categoriec:
1. Those events that are bounded by the pump coastdown event with respect to minimum DNBR,
2. The locked rotor event, j
3. Events that require offsite dose calculations, and
4. The control rod ejection event.
7-2 B&W FuelCompany E. .l.
Rev. 1 10/94 I
The rod ejection event is separated because although it has an acceptance criterion based on DNB, it is limiting in terms of peak fuel enthalpy. The DNB-J imiting events are discussed in section 6.
An evaluation of the rod ejection event was performed for cycle 10 to determine L the maximum allowable qF that results in peak fuel enthalpy less than 210 cal /gm.
The limit value is well below the acceptance criterion given in Regulatory Guide h 1.77, but was conservatively chosen for this evaluation because it represents a threshold value for zirconium-water reaction. The integrated neutron power from the limiting case in the USAR, assuming an ejected rod worth of 0.65 tak/k from l BOC, HFP conditions, was used to determine the maximum allowable F q assuming an l p adiabatic heat-up of the fuel. The maximum allowable Fq was determined to be k 3.43. This bounds the increased design (radial x local) peaking factor, 1.795, given the design axial peak of 1.65. Furthermore, it was determined that the percentage of pins in DNB reported in the USAR remains bounding for an Fq of 3.43. An evaluation was also performed fo; a rod ejection with three operating reactor coolant pumps from the maximum power level of 80.6 percent full power.
That evaluation determined that the ejection of a control rod assembly worth 0.65 tak/k from the three-pump condition is bounded by the four-pump case.
Generic loss-of-coolant accident (LOCA) analyses for the B&W 177-FA raised-loop I nuclear steam system (NSS) have been pe- 'ormed to calculate allowable LOCA line::r L
heat rate (LHR) limits for the Mark-B8A and Mark-B10AZL fuel types. The final acceptance criteria B&W ECCS evaluation model techniques and assumptions, as described in BAW-10lO4P, Rev. 5", were used in the analyses. The application of the evaluation model" included the effects of the NUREG-0630 fuel pin rupture curves, FLECSET reflooding heat transfer coefficient calculations, and the BWC CHF correlation. For batches 9A, 9D, and 10, analyses to justify an LHR limit of 12.0 kW/f t for cycle 10 were performed using bounding TACO 3 fuel performance data. A tabulation showing the maximum allowable LHR limits for these batches is given in Table 7-2.
For batches 11 and 12, the cycle-specific fuel temperature and internal pin pressure as functions of burnup for cycle 10 were found to be bounded by the fuel performance data assumed in the Mark-B10AZL LOCA analyses, which utilized TACO 3 e
fuel performance data. The LOCA analyses assumed a different pre-pressure than 7-3 B&W FuelCompany
l Rev. 1 10/94 58 that used in cycle 10, but the increased pre-pressure was evaluated to be .
acceptable. The batch 11 fuel is Mark-BBB, but the Mark-B10AZL LOCA limits are applicable to this batch as well. A tabulation showing the maximum allowable LHR limits for batches 11 and 12 is given in Table 7-2. .
The LOCA analyses were also reviewed for the effect of the increased design (radial x local) peaking factor. The LOCA limits were established using an axial peaking factor of 1.7. Based on a sensitivity study performed in reference 14, a reduction in axial peaking factor (and thus, an increase in radial peaking f actor to maintain the same peak LHR limit) would result in lower PCTs. The LOCA limits, therefore, represent an upper limit on the axial peaking factor, as long as the initial fuel pin conditions remain bounded by the analysis. The radial peaking factor may be increased as long as the LOCA limits are not exceeded at any elevation.
The impact of the two stainless steel rods in assembly NJORAA on power peaking was evaluated based on full-core power distributions. An evaluation of the peaking changes in both assembly NJORAA and adjacent fuel assemblies was :
performed to verify that the fuel rod peaking increases meet the criteria 4
established in reference 13 for minimum DNBR prediction and LOCA analysis initial conditions. The maximum increase in the peak pin power over the entire fuel cycle operation was less than 0.1%. Based on the results of the peaking increases for affected fuel rods, no additional DNBR or ECCS analyses were required.
It is concluded by the examination of cycle 10 core thermal, thermal-hydraulic, and kinetics properties, with respect to acceptable previous cycle values, that this core reload will not adversely affect the ability to safely operate the Davis-Besse plant during cycle 10. Considering the previously accepted design basis used in the USAR and subsequent cycles, the transient evaluation of cycle .
10 is considered to be bounded by previously accepted analyses. The initial Es conditions of the transients in cycle 10 are bounded by the USAR and/or subsequent cycle analyses.
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I 7A. ACCIDENT AND TRANSIENT ANALYSIS FOR FINAL SIXTY-FOUR FEED CYCLE 10 The physics evaluation of the final sixty-four feed cycle 10 core showed that the physics parameters used in the safety analysis evaluation for the initial sixty-four feed cycle 10 were either essentially unchanged or bounding. The final sixty-four feed cycle 10, therefore, has no adverse effect on the previous safety '
analysis evaluation.
The batch burnup data were examined for the initial and final sixty-four feed cycle 10 designs. No significant differences in the burnup data were found and i the accident doses for the initial sixty-four feed cycle 10 remain valid and
'W applicable for the final sixty-four feed cycle 10 design.
The RPS offset protective limits were unchanged for the final sixty-four feed I cycle 10, and the detector uncertainty is unchanged. The initial sixty-four feed cycle 10 RPS power / imbalance / flow trip setpoint and safety limit calculations,
.I therefore, remain valid.
The boron BOC concentrations were unchanged for the final sixty-four feed cycle W 10 design. The minimum required BWST boron concentration for the initial sixty-four feed cycle 10 remains valid.
I The fuel performance evaluation was shown to remain applicable for the redesigned core. The allowable LOCA limits are based on the fuel performance evaluation and therefore, the allowable LOCA linear heat rate limits for the initial cycle 10 core remain valid for the final cycle 10 core.
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l Rev. 1 ,
10/94 ,l l
Table 7-1. Comotrison of Kev Parameters for Accident Analysis l 1
USAR Cycle 10 Parameter value value ,
BOL") Doppler coeff, 10~3, % Ak/k/*F -1.28 -1.55 EOL*) Doppler coeff, 10-3, % Ak/k/'F -1. 4 5(*) -1.70 '
BOL moderator coeff, 10-2, % Ak/k/'T +0.13 -0.61 EOL moderator coeff, 10 2, %Ak/k/*F -4.0 -3.38 EOL temperature coef f (532 to 510*F) -3.10 -2.70 10'3, %Ak/k/F All rod bank worth (HZP), %Ak/k 10.0 6.32 l I
100 152 Boron reactivity worth (HFP),
ppm /%Ak/k I E]
E Hax ejected rod worth (HFP), %Ak/k O . 6 5(* O.31 l .
Max dropped rod worth (HFP), %Ak/k O.65 <0.20 Initial boron cone (HFP), ppm 1407 1831") l
") BOL denotes beginning of life.
) EOL denotes end of life. =
4') -1.77 x 10-' % Ak/k/'F was used for steam line f ailure analysis.
Calculational uncertainty (15%) is applied to the limit in the design analysis when determining cycle-specific regulating group position limits.
") Includes allowances for B atom variations and reactivity anomalies.
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7-6 B&W FuelCompany g[ .
I Rev. I 10/94 Table 7-2. Boundino values for Allowable LOCA Peak Linear Heat Rates Mark-B8A Fuel Tvoe Allowable Peak LHR for Specified Burnup, kW/ft Less Than After Core 24,500 24,500 W Elevation, ft mwd /mtU mwd /mtU 0 12.8 9.6 0.375 12.8 9.6 1 16.0 12.0 2 16.0 12.0 4 15.75 12.0 I 6 8
10 16.5 17.25 17.0 12.0 12.0 12.0 I 11 11.625 12 17.0 13.6 13.6 12.0 9.6 9.6 Mark-B8B and Mark-B10AZL Fuel Tvoes Allowable Peak LHR for Specified Burnup, kW/f t*
I Core glevation, ft.
Less Than 20,000 mwd /mtU 20,000 mwd /mtU 47,016 mwd /mtU 47,557 mwd /mtt) 49,180 mwd /mtU 55,131 mwd /mtU 0 14.24 13.6 13.6 13.44 12.96 11.2 I 0.375 1
2 14.24 17.8 17.8 13.6 17.0 17.0 13.6 17.0 17.0 13.44 16.8 16.8 12.96 16.2 16.2 11.2 14.0 14.0 I 4 6
8 17.8 17.5 18.0 16.2 16.2 16.8 16.2 16.2 16.8 16.2 16.2 16.8 16.2 16.2 16.2 14.0 14.0 14.0 10 18.3 17.0 17.0 16.8 16.2 14.0 1 11 11.625 18.3 14.64 17.0 13.6 17.0 13.6 16.8 13.44 16.2 12.96 14.0 11.2 12 14.64 13.6 13.6 13.44 12.96 11.2 Linear interpolation between burnup points to calculate the Allowable LHR limit is allowed.
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Rev. 1 10/94 4
8. PROPOSED MODIFICATIONS TO COhG OPERAT NG LIMITS REPORT FOR INITIAL SIXTY-FOUR FEED CTCis' 10
. The Core Operating Limits Report (COLR) has been revised for cycle 10 operation to accommodate the influence of the cycle 10 core redesign on power peaking, l reactivity, and control rod worths. Revisions to the cycle-specific parameters were made in accordance with the requirements of NRC Generic Letter 88-16 and Technical Specification 6.9.1.7. The core operating limits were determined from a cycle 10 cpecific power distribution analysis using NRC approved methodology
provided in the reference to A chnical Specification 6.9.1.7.
A cycle 10 specific analysis was conducted to generate the axial power imbalance
" protective limits, corresponding trip setpoints, and the Limiting conditions for Operation (rod index, axial power imbalance, and quadrant tilt), based on the 1
d NRC-approved methodology described in reference 9. The analysis incorporates the DNB peaking limits based on the allowable increase in design (radial x local) peaking provided by the Statistical Core Design methodology described in reference 10. The ef fects of gray APSR repositioning were included explicitly in the analysis. The analysis also determined that the cycle 10 core operating limits provide protection for the oveQower condition that could occur during an I overcooling transient because of nuclear instrumentation errors. The maximum allowable LOCA linear heat rate limits used in the analysis are based on the ECCS analysis described in section 7.2. Table 8-4 provides the burnup- and elevation-dependent linear heat rate limits for each incore segment. They are the basis for the rod index and axial power imbalance operating limits required by Technical Specifications 3.1.3.6. and 3.2.1. The two stainless steel rods in assembly NJORAA were determined to have an insignificant impact on the cycle 10 core operating limits.
The measurement system-independent rod position and axial power imbalance limits determined by the cycle 10 analysis were error adjusted to generate alarm setpoints for power operation. These setpoints are implemented in the COLR starting with cycle 10. Figures 8-1 through 8-10 are revisions to the operating l 8-1 I
B&W FuelCompany
. . as Rev. I 10/94 limits contained in the COLR and have been adjusted for instrument error to provide alarm setpoints. Figures 8-11 and 8-12 are the core protective limits l and RPS imbalance trip setpoints, both of which have been relocated to the COLR starting with cycle 10. Figure 8-13 provides the allowable radial peaking factors .
to be used in the calculation of the F1 limits. The regulating group position " '
alarm setpoints in Figures 8-1 and 8-4 at 0%FP for the first operational window were set based on ejected rod worth analysis. The analysis showed that the group 5 insertion limit is restricted to 17.8%wd (Figure 8-1) and 18.1%wd (Figure 8-4) for hot zero power (HZP) conditions during this period. Therefore, the restricted regions below 26.5%FP (Figure 8-1) and 19.9%FP (Figure 8-4) are based on the ejected rod criterion, and the restricted regions above 26.5%FP (Figure 8-1) and 19.9%FP (Figure 8-4) are based on the power peaking criteria (allowable LOCA linear heat rate and initial condition DNB peaking limits). The analysis showed that the rod index restriction due to ejected rod worth can be removed W after 25 EFPD, at which time the HZP ejected rod worth decreases to 0.736%Ak/k.
The 3-RCP axial power imbalance alarm setpoints provided in Figure 8-10 are based on the 4-RCP, full-power LOCA LHR limits, which were verified to be applicable for 3-RCP operation. Table 8-1 presents the quadrant power tilt setpoints for cycle 10, Table 6-2 provides the negative moderator temperature coef ficient limit for cycle 10, and Table 8-3 provides minimum linear heat rate to melt (kW/ft) limits. Table 8-4 provides the Fg limits and Table 8-5 provides the FL limits, both of which have been relocated to the COLR starting with cycle 10. The Fn limits reflect the two different active fuel lengths and respective allowable E linear heat rates as functions of incore segment (core elevation) and burnup.
The F1 relationship defined in Table 8-5 ensures acceptable DNBR performance using Statistical Core Design Methodology in the event of the limiting condition I and Il transient. The family of curves in Figure 8-13 preserves initial condition DNB margin in the form of allowable initial condition peaking.
Allowable F1 values can be determined based on particular axial peaks at a given axial elevation for either three or four RC pump operation.
Based on the analyses and operating limit revisions described in this report, the )
Final Acceptance Criteria ECCS limits will not be exceeded, nor will the thermal design criteria be violated.
8-2 saw roeicarevy [,
BA. PROPOSED HODIFICATIONS TO CORE OPERATING LIMITS REPORT FOR FINAL SIXTY-FOUR FEED CYCLE 10 A power distribution evaluation of the final sixty-four feed core was performed to verify that the core protective and operating limits for the initial sixty-four feed core are applicable to the final core design. The evaluation was performed in accordance with the approved methods described in BAW-lOl79P-A.
A Margins to the F, and F1 peaking limits were determined to be bounded by limits set for the original analysis. The reactivity parameters pertaining to
{ determination of rod index limits based on shutdown margin were examined using inputs for the final core design. No changes to the shutdown margin-based rod index limits were required. The BOC hot zero power (HZP) ejected rod worth I (0.764%Ak/k) is within the analysis limit. Therefore, the restrictions placed on the rod index limits for the initial sixty-four feed cycle 10 are bounding with respect to all applicable criteria. The two stainless steel rods in assembly NJORAA were determined to have an insignificant impact on the core power distribution for the final cycle 10 design. Based on the results of the evaluation, it is concluded that the core protective limits, trip setpoints, and operating limits provided for the initial cycle 10 design are valid and applicable for operation with the final cycle 10 design.
k 8-3 B&W FuelCompany
DAVIS-BESSE I Figuro 8-1 ,
E' Rev. I 10/94 S Figure Regulatinq Group Position Alarm Setpoints Oto 25+10/-O EFPD, Four RC Pumps Davis-Besse I, Cycle 10 lE This Figure is referred to by Technical Specifications 3.l.3.6 and 3.I.3.8 IIO i POWER LEVEL / r i00 ____ CUTOFF = 100% ,f, ,f___.__
j.
x ; / !/ l y90 i / /
a.
80 I
/ ,/! =
3 r
k [ UNACCEPTABLE OPERATION r
/
I i hJ 70 j; l I IOPERATION F jfRESTRICTED1 a 60 ,
gm '" ,, ,_
j ir [i) ];
W MARGIN " / '
A-Q LIMIT jQ f x 50 / /
f f o
40 I~~' / / i c l i j ACCE TABLE
' OPERATION 8 30 l % e 2r ,
[
I L ____OPER'ATION
[ 20 :$,RESTRIC,TEDio _ipp
- g
__.g j re c
___4___ n__ jf 5 O,
O 25 50 75 100 125 150 175 200 225 250 275 300 Rod Index (% Withdrawn)
O 75 100 E I I l g~
GROUP 5 0 25 75 100 1:
I I GROUP 6 I l lJ 5'
O 25 100 I I I GROUP 7 i
A Rod Group overlap of 25*5% between sociuentlat withdrawn groups 5 Note I and 6, and 6 and 7 shall be maintained Note 2: Instrument error is accounted for in these setpoints.
1 b~
E DAVIS-BESSE I g , Figure 8-2 I
Rev. 1 10/94 I Figure Regulating 25+10/-O Group to 460 Davis-Besse I, Cycle 10
lO EFPD, Position Four Alarm RC PumpsSetpoints This Figure is referred I to by Technical Specifications 3.l.3.6 and 3.l.3.8 IIO i00 :::: POWER LEVEL / r CUTOFF = 100%
t. ,
,/ ,
/-----
T / }
q 90 ; ,r 1
6 s r U 0. / l
_; 80 -
j 1 G .---
l W 70 T l I
F
---/n i j OPERATIONRESTRICTEDI
. o 60 gmw ___
l y)
W MARGIN: ;FP e 17 y LIMIT ; \l / / '
F T 50 \b l /
+ (
f f 40 i t / /
ER u 30 OP AT e ..r e f b 20 i /
' 7 e , ,
f' g 10 4_,e e O Ri O 25 50 75 100 125 150 175 200 225 250 275 300 Rod Index (% Withdrawn)
- O 75 100 1 I I GF.OUP 5 0 25 75 100 l I I J GROUP S O 25 100 1 I I OROUP 7 Note la A Rod Group overlap of 25*5% between sociuen t i al wi thdrawn groups 5 and 6, and 6 and 7, shall be maintained Note 2: Instrument error is accounted for in these setpoints.
8-5 B&W FuelCompany
DAVIS-BESSE I 3,y, g:
3 Figuro 8-3 10/94'. as '
Figure Regulating Group Positlon Atarm Setpoints E After 460110 EFPD, Four RC Pumps --
W Davis-Besse I, Cycle 10 This Figure is referred to by Technical Specifications 3.l.3.6 and 3.l.3.8 .
IIO POWER LEVEL ?1""
I, I00 -::: CUTOFF = l00% L--_
,- f---_
g- /. j-.---_
90 j 7 t'-
G. l---)f l
a BO p i--_ ( ---
k j _-___
Y 70 / ll I
H
/ ----
OPERATION
/ , RESTRICTED a 60 SHUTDOWN--- -[ ----
LD MARGIN -
~;P /[ s' Q LIMIT -l / ./
a 50 ; ,7
,e
+ ) , '_
,: ~_.__
40 ._____
'j j
+-

.. UNACCEPTABLE

~~~~~
l 7# E 5 ' OPERATION --} ----3 hC ____
u 30 _____
g_g a ::::: J -
b 20 -____.
p f
u ',L.-
o s g 10 ____.
f I
O.
O 25 50 75 100 125 150 175 200 225 250 275 300 Rod Index (% Withdrawn)
O 75 100 I I I .
GROUP 5 0 25 75 I00 I
I GROUP 6 i 1
]
0 25 100 I I I GROUP 7 Note I: A Rod Group overlap of 25*5% between s o ciu e n t l e t withdrawn groups 5 and 6, and 6 and 7, shall be maintained Note 2: Instrument error is accounted for in these setpoints.
8-6 B&W Fuel Company B.l;
Figuro 8-4 "$J,4 Figure Regulating Group Position Alarm Setpoints I O to 25+10/-O EFPD, Three RC Pumps --
Davis-Besse I, Cycle 10 This FI is referred
. tobyT!urechnical Speelfications 3.l.3.6 and 3.l.3.8 t
i00 C
8 g 90 O
Q. , l 80 I
i d
y E:: POWER LEVEL
CUTOFF = 75% /
/ f/ '
E 70 /i / I w r
!Y u l
E 1
a 60 / /
til F
UNACCEPTABLE OPERATION r
/
$ 50 ,ehRATION -
- / :RESTR,ICTEDj I
40 m
MARGIN , 7 l
" g #
Jr' o 30 LIMIT ' e' -
'f ACCEPTABLE s.
I o t s
O OPERATION
n. 20
- -J 4
/
I c e
$ 10 o
.-OPERATION
5tRESTRICTEO -dM
---N 7-l v
r M7 i O 5k h u--
0 25 50 75 100 125 150 175 200 225 250 275 300 Rod Index (% Withdrawn)
O 75 100 I
l l
GROUP 5 I
0 i
25 I
I 75 I
i00 I
t GROUP 6 0
1 25 I
100 I
GROUP 7 I Note 1: A Rod Group overlap of 25*5% between s o ciu e n t i a l withdrawn groups 5 and 6, and 6 and 7, shall be maintained Note 2: Instiument error is accounted for in these setpoints.
8-7 B&W FuelCompany
DAVIS-BESSE I Figuro 8-5 Rg;,j. ,
g Figure i-Regulating 25+10/-O Group to 450 *lO EFPD,Positlon Three Alarm Setpoints RC Pumps --
Davis-Besse I, Cycle IO This Figure is referred to by Technical Specifications 3.l.3.6 and 3.1.3.8 110 100 C E g 90 3 O !
Q.
80 E d
y h :: POWER LEVEL
CUTOrF = 75% -- e
/ f fMZ1 g
g 70 y' ' ,/ g
& / l 6 a 60 i / /
tti UNACCEPTABLE OPERATION e
[ l A
H 1
<C g 59 / OPERATION
, _____ ,f RESTRICTEDj 40 '
/ Ur a
_ _ .SHUTDCfdN . J ?
?
MARGIN -
y /
5 i LIMIT #,
u 30 ~----
/
iei i / ACCEPTABLE g e
e i IV OPERATION b 20 5,#
c g L !pf g y .---. -
f3 g 10 ____
jy- ;
O O 25 50 75 100 125 150 175 200 225 250 275 300 Rod Index (% Withdrawn) 76
'?
GROUP 5 0 25 75 100 ll.
i 1 I '
GROUP S E3_
O 25 100 5-l l I GROUP 7 Note 1 A Rod Group overlap of 25*5% between s ociuen t i al wi t hdrawn groups 5 and 6, and 6 and 7, shall be maintained Note 2: Instrument error is accounted for in these setpoints. W 8-8 B&W FuelCompany J[
f t
DAVIS-BESSE I Rev. 1 E . Figuro 8-6 10/94 5 . .
1
! Figure Regulating Group Position Alarm Setpoints I After 460*lO EFPD, Three RC Pumps --
Davis-Besse I, Cycle 10 l
This Fi is referred
, to byT!urechnleal SpectfIcatIons l
3.l.3.6 and 3.l.3.8 6
110 >
l. l l00 I [
90 -
t [
80
' ~'
a i l ---- POWER LEVEL r A) h 3E:: CUTOFF '= 75%' f[ f w 70 E
l l ,/,'/
h _ . ! !
g 60 i / /
W F
UNACCEPTABLE OPERATION /
/ I 4
E 50 OPERATION -1 i
l RESTRICTED j
/
l *~
o 40 SHUTDdM MARGIN LIMIT iii 7, --
/
/
e j#
s -
P o 30
,r; f L ----------ACCEPTABLE e l 2 V'! OPERATION b 20 ,. ,i - i '
L v e J r 10 -
u $ ::37;[
~
O O 25 50 75 100 125 150 175 200 225 250 275 300 '
Rod Index (% Withdrawn)
O 75 100 I I I GROUP 5 O 25 75 100 l I I I GROUP 6 0 25 100 l I I GROUP 7 Note I: s e r.
Aand Rod 6,Group and overlap 6 and of 7,25*5% shall betweenma be int ai ned ;uent i al withdrawn groups 5 Note 2: Instrument error is accounted for in these setpoints.
8-9 B&W FuelCompany
DAVIS-BESSE I Figure 8-7 3 y, 1, gl 10/94 . ,
1 Figu.e Control Rod Core Locations ,,
and Group Assignments l!
Davis-Besse I, Cycle 10 This FI is referred to byT!urechnical Specificotion 3.l.3.7 4 -N I
A 4 6 4 B
2 5 5 2 C
8 7 8 7 D 7 5 5 2 E 2 6 8 4 4 8 6 3 F
1 1 5 G 5 7 6 M
3 H W- 6 7 3 4 1 1 5 K 5 3 6 8 4 i4 8 6 L
5 2 il i 2 5 8 7 8 7 7
N l 2 5 5 2 0 l l 4 6 4 P l l R l 1 1
,1 Z
10 11 12 13 14 15 2 3 4 5 6 7 8 9 1
I 3 Group Number Group No. of Rods Function 1 4 Safety 2 8 Safety i 3 4 Safety RJ 4 9 Safety 3-5 12 Control 6 8 Control 7 8 Control 8 APSRs
_8_
Total 61 I
8-1o B&W FuelCompany
/94 DAVIS-BESSE I Figure 8-8 -
Figure APSR Position Alarm Setpoints This Figure is referred to by Technical Specification 3.1.3.9 Before APSR Pull - 0 EFPD to 460 +/- 10 EFPD, Three or Four RC n_ um_os o_neration*
( Lower Setpoint: 0 %WD
[ Upper Setpoint: 100 %WD After APSR Pull - 460 +/- 10 EFPD to End-of-Cycle Three or Four RC pumps operation
l Insertion Prohibited per Tech Spec 3.1.3.9
Power restricted to 77% for 3 pump operation 8-11 B&W FuelCompany
DAVIS-BESSE I Figuro 8-9 E
I?)s4 - -
Figure AXIAL POWER IMBALANCE ALorm Sotpoints E O EFPD to EOC, Four RC Pumps --
5 Davls-Besse I, Cycle 10 This Figure le referred to by Technicol SpeelfIcatlon 3.2.1 l10 I
) l00 i l ,
\
t
\
i 5 go
/ ,' L 1
^
) /
---80 i }
CTED ,
g,_ !l RESYRICTED I .
REGION L
I F.__so El i er -
i '
"];
! $---so !
. i
[
l 1
$---40 ' l l
i c !
i i a e -
$---30 i
l l
1 20 i
l-l OPERATING PERMISSIBLE I E
~
REGION gO ; g 31
-40 -30 -20 -10 O 10 20 30 40 AXIAL POWER IMBALANCE % )
LEGEND FULL INCORE EXCORE Note I: Instrument error is accounted for in these setpoints.
8-12 B&W FuelCompany ,3
DAVIS-BESSE I Rev. 1
. Figure 8-10 1o/94 Figure AXIAL POWER IMBALANCE Alarm SetpoInts O EFPD to EOC, Three RC Pumps --
Davis-Besse I, Cycle 10 This Figure is referred to by Technical Specification 3.2.I F
I, I10 .
+
lO0 .
90 i
i 80,
^ 1 i < .
't
[
GI N / j ---70 'g gy{RCTED._
, / / L ,
/ J =
{ ( y---60 ) ,
. E , 4
! k !
l j ---50 j '
[ i.
t t
i < 1
! m---40 ,!
i' 4
U---30 j l! c :!
I l o
!I P
20
} h $
3 'j PERMISSIBLE ll i OPERATING j REG"ON 10 .i
.i
{ 4
! l!
i! '!
[ -40 -30 -20 -1O O IO 20 30 40 AXIAL POWER IMBALANCE % .
LEGEND FULL INCORE EXCORE - - - - - -
Note I: Instrument error is accounted for in these setpoints.
8-13 B&W FuelCompany .
..-w.
R0v. 1- ,
as l'O/94 DAVIS-BESSE I 31 Figure 8-1I Ei Figure AXIAL POWER IMBALANCE Protective Limits Thle to by Flo'ure is referred Technical Specification 2.l.2
% RATED THERMAL POWER I
120 4 P N LIMIT (37.09,i12)
(-37.09,i12)
--810
(-51.45,100) --100 (55.21,96.OS 3 PUMP LIMIT 90
(-37.09,89.80) (37.09,89.80)
-- 80 g=
(55.21,73.8s)
(_si.45,77.so)
-- 70 60
- 50
_ 40 UNACCEPTABLE UNACCEPTABL OPERATION OPERATION
- - 30
- - 20 ACCEPTABLE OPERATION FOR SPECIFIEO RC PUMP COMBINATION -- 10 1 i 1 i f 1 I f I f f f 1
-60 -50 -40 -30 -20 -10 O 10 20 30 40 50 60
}
AXIAL POWER IMSALANCE %
Required Measured Reactor Coolant Flow to Ensure Pumps Operatin9 Flow, gpm Compliance, gpm 4 380.000 389,500 290.957 "
3 283.860 8-14 B&W FuelCompany gi'
i 1
,- . DAVIS-BESSE I Ra;}9 Figure 8-12 Figure FL ux-- A F L ux/F L ow (or Power /Imbatonce/ Flow)
Trip Se1 points This Figure is referred L to by Technical Specification 2.2.I
% RATED THERMAL POWER
(-17,108) (17,108)
--I10 _ .
M2=-I.O Ml=l.O --100 4 PUMP LIMIT Curve shows trip
- -90 setpoint for on
(-38,87)< (38,87) o 2gproximately"
(-17,80.6) (17,80.6) or e m
--80 operation (283,860 gpm).
e The actual setpoint
--70 3 PUMP will be calculated EXAMPLE by the Reactor Protection System and will be directly
(-38,59.6) -
-60 ,
proportional to the (38,59.6) actual flow with three pumps.
-50
--40 UNACCEPTABLE UNACCEPTABLE OPERATION OPERATION _
39 ACCEPTABLE -
-20 OPERATION FOR SPECIFIEO RC PUMP COMBINATION
--10 I I I I I I f I I f f I
-60 -50 -40 -30 -20 -10 O 10 20 30 40 50 60 AXIAL POWER IMBALANCE %
8-15 B&W FuelCompany
l R2v, 1 SI g
l 10/94 Table 8-1 l!
Table OUADRANT POWER TILT Limits This Table is referred D'
to by Technical Specification .
3.2.4 ;
Steady-state Steady-state Limit for Limit for QUADRANT POWER TILT THERMAL THERMAL Transient Maximum as measured by: POWER s 60% POWER > 60% Limit Limit
(%) (%) (%) (%)
Symmetrical Incore 6.8 4.4 10.03 20.0 detector system Table 8-2 Table Necative Moderator Tempera";ure Coef ficient_ Limit This Table is referred to by Technical Specification 3.1.1.3c ;
d Negative Moderator Temperature -
Coefficient Limit -3.73 X 10 ak/k/'F l ~
(at RATED THERMAL POWER)
Table 8-3 Table Power to Melt Limits This Table is referred to by Technical Specification Bases B2.1 Batch 9A Batch 9D Batch 10B Batch 11 Batch 12 I
Fuel Assembly l Type Mark-B8A Mark-B8A Mark-B8A Mark-B8B }hrk-B10AZL g Minimum linear i heat rate to i melt, kW/ft 20.5 20.5 20.5 22.~ 22.3 1 I l I
8-16 I
B&W FuelCompany B.:
Rev. 1 10/94 Table 8-4 Table Nuclear Heat Flux Hot Channel Factor - F9
This Table is referred to by Technical Specification 3.2.2 Heat Flux Hot Channel Factor F 9 Fn shall be limited by the following relationships:
Fo $ LHR^""(Bu) / [ LHR**
P ) (for P $ 1.0)
LHR^" "(Bu): See the Tables below LHR^W = 6.139 kW/ft for Mark-B8A fuel . . .
LHR^VU = 6.253 kW/ft for Mark-BBB fuel ain" = 6.253 kW/ft for Mark-B10AZL fuel P = ratio of THERMAL POWER / RATED THERMAL POWER Bu = Fuel Burnup (mwd /mtU)
Batch 9A (Mark-B8A) LHR^#
kW/ft Less than After Axial Seament 24,500 mwd /mtU 24,500 mwd /mtU 1 12.8 9.6 2 15.1 11.3 3 15.7 12.0 4 15.7 12.0 5 16.4 12.0 .
b 6 17.0 12.0 7 16.2 11.4 8 13.6 9.6
(
J Y
8-17 ,
B&W FuelCompany
Rav. 1 10/94
{
Table 8-4, continued "" l Batches 9D and 10B ' Mark-B8A) LHR***' kW/ f t j; l
85 After Axial Secment 24,500 mwd /mtU l'
1 9.6 2 11.3 3 12.0 .
4 12.0 5 12.0 l 6 12.0 7 11.4
'8 9.6 Batch 11B (Mark-BBB) LHR^**' kW/ft After Axial Seament 20,000 mwd /mtU 1 13.6 2 16.0 l
3 16.2 g!
4 16.2 E '
5 16.2 6 16.6 7 16.2 8 13.6 Batch 12 (Mark-B10AZL) LHR***' kW/ft Less Than After 20,000 mwd /mtU 20,000 mwd /mtU Axial Seament 1 14.2 13.6 2 16.8 16.0 3 17.7 16.2
=
4 17.5 16.2 5 17.5 16.2 6 17.8 16.6 7 17.4 16.2 .
8 14. 13.6 I
I 8-28 saw Fuelcompany
T::blo 8-5 Table NuclearEnthalovRiseHotChannelFactor-F$
R3v. 1 10/94 I This Table is referred to by Technical specification i
3.2.3 Enthalov Rise Hot Channel Factor Ph F$n i ARP (1 + 0. 3 (1 - P/P ) ]
ARP = Allowable Radial Peak, see Figure I P = THERMAL POWER / RATED THERHAL POWER and P $ 1.0 P, = 1.0 for 4-RCP operation P = 0.75 for 3-RCP operation Figure 8-13 Figure Allowable Radial Peak for F5H 2.1 ; ;;; ; ;
V,.
- hU '**
x'i, i Q- gii Y "
. l g j, i g 1
< 2.0 gg,, c -y I
==
v A,4,.7,
,, r
'sO; j -
g .g, g
U ' 'A '(
I
l9l ; lll A,. i N N ,, ;
0. -
J i ,m N N i
g
_., N A '.s A_-__-
; ;'s--
N N I o l
l.8 N.,
f 28. Ih?
o 'N '
N % l
'% x N, r----
I T s
e 1.7
%, m s, -
A s m.
A 's 1 1 56.24"5 '
V,,
4W ll I
_O ,,
c ,,
D * ,m l ,%
o I.6 ,-. . .1 .i t 84.3,6 h":
I a , y
<C ,
',N__-_
1 : - wr--
I.5 i . .. 1 3, I
_ I,12.48";
I l.4 1.1 1.2 1.3 1.4 I.5 1.6 1.7 1.8 I.9 Axial Peak
Based on an active core height of 140.6 inches. Linear interpolation and extrapolation above 112.48 inches are acceptable. For axial heights <28.12 inches, the value at 28.12 inches will be used.
8-19 B&W FuelCompany
I,*, Rsv. 1 10/94 I
9. STARTUP PROGRAM - PHYSICS TESTING i The planned startup test program associated with core performance is outlined below. These testa vorify that core performance is within the assumptions of the safety analysis and provide information for continued safe operation of the unit.
9.1. Precritical Tests 1
9.1.1. Control Rod Trio Test Precritical control rod drop times are recorded for all control rods at hot full-flow conditions before zero power physics testing begins. Acceptance criteria state that the rod drop time from fully withdrawn to 75% inserted shall be less than 1.58 seconds at the conditions above. 1 It should be noted that safety analysis calculations are based on a rod drop from 1
fully withdrawn to two-thirds inserted. Since the most accurate position 1
indication is obtained from the zone reference switch at the 75% inserted position, this position is used instead of the two-thirds inserted position for data gathering.
9.1.2. PC Flow I Reactor coolant flow with four RC pumps running will be measured at hot standby conditions. The measured flow shall be within allowable limits.
9.2. Zero Power Physics Teste 9.2.1. Critical Boron concentration once initial criticality is achieved, equilibrium boron is obtained and the critical boron concentration determined. The critical boron concentration is calculated by correcting for any rod withdrawal required to achieve the all rods out equilibrium boron. The acceptance criterion placed on critical boron concentration is that the actual boron concentration shall be within i 50 ppm boron of the predicted value.
I I
g B&W FuelCompany
Rrv. 1 . $!
1 10/34
m 9.2.2. Temoerature Reactivity Coefficient B,'
E' The isothermal HZP temperature coefficient is measured at approximately the a21-rods-out configuration. During changes in temperature, reactivity feedback may be compensated by control rod movement. The change in reactivity is then I calculated by the summation of reactivity associated with the temperature change. l The acceptance criterion for the temperature coefficient is that the measured value shall not differ from the predicted value by more than 10.2 x 102 ,
% Ak/k/*P .
The moderator temperature coef ficient of reactivity is calculated in conjuncticn {
with the temperature coef ficient measurement. Af ter the temperature coef ficient has been measured, a predicted value of fuel Doppler coefficient of reactivity is subtracted to obtain the moderator temperature coef ficient (MTC). This value shall be less than +0.9 x 10-2 % Ak/k/'F. The MTC is also e:trapolated to full power conditions.
9.2.3. Control Rod Grouo/ Boron Reactivity Worth Individual control rod group reactivity worths (groups 5, 6, and 7) are measured at hot zero power conditions using the bcron/ rod swap method. This technique consists of deborating the reactor coolant system and compensating for the reactivity changes from this deboration by inserting individual control rod groups 7, 6, and 5 in incremental steps. The reactivity changes that occur during these measurements are calculated based on reactimeter data, and 5 incremental rod worths are obtained from the measured reactivity worth versus the change in rod group position. The incremental rod worths of each of the controlling groups are then summed to obtain integral rod group worths. The acceptance criteria for the control rod group worths are as follows:
1. Individual group 5, 6, 7 worths predicted value - measured value x 100% shall be < 15%
predicted value ,
li
2. Sums of groups 5, 6, and 7:
oredicted value - measured value predicted value x 100% shall be < 10%
Il I
I 9-2 B&W FuelW g
r R v. 1
.I,', 10/94 The boron reactivity worth (differential boron worth) is measured by dividing the total inserted rod worth by the boron change made for the rod worth test. The j acceptance criterion for measured differential boron worth is as follows:
]
loredicted value - measured valuel l
I l predicted value l x 100% shall be < 15%
I The predicted rod worths and differential boron worth are taken from the ATOM. l s
9.3. Power Escalation Tests 9.3.1. Core Symmetry Test The purpose of this test is to evaluate the symmetry of the core at low power during the initial power escalation following a refueling. Symmetry evaluation is based on incore quadrant power tilts during escalation to the intermediate power level. The absolute values of the quadrant power tilts should be less than the COLR limit.
t 9.3.2. Core Power Distribution Verification at Intermediate Power Level (IPL) and -100% FP l
Core power distribution tests are performed at the IPL and approximately 100%
l l
full power (FP). Equilibrium xenon is established prior to the -100% FP test. !
The test at the IPL (40-80 %FP) is essentially a check of the power distribution l in the core to identify any abnormalities before escalating to the -100% FP plateau. Peaking factor criteria are applied to the IPL core power distribution j results to determine if additional tests or analyses are required prior to -100%
FP operation.
The following acceptance criteria are placed on the IPL and -100% FP tests:
1. The maximum LHR shall be less than the LOCA limit.
2. The value obtained from extrapolation of the worst-case maximum LHR to the g next power plateau overpower trip setpoint shall be less than the fuel melt limit, or the extrapolated value of imbalance must fall outside the RPS power / imbalance / flow trip envelope.
3. The maximum FL value shall not exceed the limits specified in the COLR.
4. The measured radial (assembly) peaks for each 1/8 core fresh fuel location l shall be within the following limits:
9-3 B&W FuelCompany
Rrv. 1 -
10/94 predicted value - measured value x 100% more positive than -3.8% t predicted value 5.
The measured total (segment) peaks for each 1/8 core fresh fuel location l l shall be within the following limits:
predicted value - measured valu_2 x 100% more positive than -4.8%
predicted value The following review criteria also apply to the core power distribution results ,
i at the IPL and at -100% fps
6. The 1/8 core RMS of the dif ferences between predicted and measured radial l (.
(assembly) peaking factors should be less than 0.05. )
5-
7. For all 1/8 core locations, the (absolute) difference between predicted and measured radial (assembly) peaking factors should be less than 0.10.
Items 1 and 3 ensure that the initial condition limits are maintained at the IPL l and -100% FP.
Item 2 establishes the criteria whereby escalation to full power may be accomplished without exceeding the safety limits specified by the safety analysis 3" with regard to DNBR and linear heat rate.
Items 4 and 5 are established to determine if measured and predicted power l distributions are within allowable tolerances assumed in the reload analysis.
Items 6 and 7 are review criteria, established to determine if measured andl predicted power distributions are consistent.
9.3.3. Incore vs. Excore Detector Imbalance Correlation Verification Imbalances, set up in the core by control rod positioning, are read simultaneously on the incore detectors and excore power range detectors. The excore detector of f set versus incore detector of f set slope shall be greater than 0.96 and the y-intercept (excore offset) shall be between -2.5% and 2.5%. If either of these criteria are not met, gain amplifiers on the excore detector signal processing equipment are adjusted to provide the required slope and/or intercept.
j '
I I
9-4 ggg pjggggggggy
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Rav. 1 10/94
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9.3.4.
Hot Full Power All Rods Out Critical Boron Concentration j The hot full power (HFP) all rods out critical boron concentration (AROCBC) is determined at -100% FP by first recording the RCS boron concentration during equilibrium, steady state conditions.
Corrections to the measured RCS boron concentration are made for control rod group insertion and power deficit (if not at 100%FP) using predicted data for CRG worth, power Doppler coefficient, and 1
differential boron worth. A correction may also be made to account for the I '
observed difference between the measured and predicted AROCBC at zero power. The review criterion placed on the HFP AROCBC is that the measured AROCBC should be l within i 50 ppm boron of the predicted value.
9.4. Procedure for Use if Acceptance / Review Criteria Not Met I If an acceptance criterion ("shall" as opposed to "should") for any test is not met, an evaluation is performed before continued testing at a higher power plateau is allowed. This evaluation is performed by site test personnel with I participation by B&W Nuclear Technologies technical personnel as required.
Further specific actions depend on evaluation results. These actions can include repeating the tests with more detailed test prerequisites and/or steps, added tests to search for anomalies, or design personnel performing detailed analyses of potential safety problems because of parameter deviation. Power is not escalated until evaluation shows that plant safety will not be compromised by such c.scalation.
If a review criterion ("should" as opposed to "shall") for any test is not met, an evaluation is performed before continued testing at a higher power plateau is recommended. This evaluation is similar to that performed to address failure of I
an acceptance criterion.
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10. REFERENCES I 1. Davis-Besse Nuclear Power Station Unit 1, Cycle 10 Reload Report, BAW-2223, B&W Fuel Company, Lynchburg, Virginia, dated July 1994.
2. Davis-Besse Nuclear Power Station No. 1, Updated Safety Analysis Report, Docket No. 50-346.
3. Davis-Besse Nuclear Power Station Unit 1, Cycle 9 Reload Report, BAW-2180, B&W Fuel Company, Lynchburg, Virginia, dated December 1992.
4. TACO 3: Fuel Pin Thermal Analysis Computer Code, BAW-10162P-A, Babcock &
Wilcox, Lynchburg, Virginia, dated November 1989.
I 5. Program to Determine In-Reactor Performance of B&W Fuels - Cladding Creep Collapse, BAW-10084P, Rev. 2, dated October 1978.
Babcock and Wilcox, Lynchburg, Virginia,
6. TACO 2: Fuel Performance Analysis, BAW-10141P-A, Rev. 1, Babcock & Wilcox, Lynchburg, Virginia, dated June 1983.
7. Fuel Rod Gas Pressure Criterion (FRGPC), BAW-10183P, B&W Fuel Company, Lynchburg, Virginia, dated July 1991 (SER dated February 22, 1994).
8. NEMO- Nodal Expansion Method Optimized, BAW-10160-A. Rev. 1, B&W Fuel company, Lynchburg, Virginia, dated March 1993.
9. Safety Criteria and Methodology for Acceptable Cycle Reload Analysis, BAW-10179P-A, B&W Fuel Company, Lynchburg, Virginia, dated August 1993.
10. Statistical Core Design for BKW-Designed 177-FA Plants, BAW-10187P-A, B&W Fuel Company, Lynchburg, Virginia, dated March 1994.
11. BWC Correlation of Critical Heat Flux, BAW-10143P-A, Babcock & Wilcox, Lynchburg, Virginia, dated April 1985,
12. Evaluation of Replacement Rods in BWFC Assemblies, BAW-2149-A, B&W Fuel Company, Lynchburg, Virginia, dated September 1993.
13. Davis-Besse Unit 1 Fuel Densification Report, EAW-1401, Babcock & Wilcox, Lynchburg, Virginia, dated April 1975.
14. B&W's ECCS Evaluation Model, BAW-10104P, Rev. 5, Babcock & Wilcox, Lynchburg, Virginia, dated April 1986.
15. ECCS Evaluation of B&W's 177-FA Raised-Loop NSS, BAW-10105. Rev. 1, Babcock & Wilcox, Lynchburg, Virginia, dated July 1975.
I 2-B&W FuelCompany

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