ML20024C092

Cycle 4 - Reload Rept.

ML20024C092
Person / Time
Site:	Davis Besse
Issue date:	05/31/1983
From:	BABCOCK & WILCOX CO.
To:
Shared Package
ML20024C071	List: ML20024C070 ML20024C081 ML20024C092
References
BAW-1783, TAC-51965, NUDOCS 8307120246
Download: ML20024C092 (69)

Text

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BAW-1783 May 1983 DAVIS-BESSE NUCLEAR POWER STATION UNIT 1, CYCLE 4 - RELOAD REPORT BABC0CK & WILC0X Utility Power Generation Division P. O. Box 1260 Lynchburg, Virginia 24505 8307120246 830705 PDR ADOCK 05000346 P pm

CONTENTS Page

1. INTRODUCTION AND

SUMMARY

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

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

3. GENERAL DESCRIPTION ........................ 3-1 4 F UE L SYS T EM DE S I GN . . . . . . . . . . . . . . . . . . . . . . . . 4-1 4.1. Fuel Assembly Mechanical Design .............. 4-1 4.2. Fuel Rod Design ...................... 4-2

4. 2.1. Cladding Co11cpse ................. 4-1 4.2.2. Cladding Stress .................. 4-1

4. 2. 3. Cladding Strain .................. 4-2 4.3. Th e rmal De s i g n . . . . . . . . . . . . . . . . . . . . . . . 4-2 4.4. Ma t e ri al C ompa t i b i l i ty . . . . . . . . . . . . . . . . . . . 4-2 4.5. Ope rati ng Expe ri ence . . . . . . . . . . . . . . . . . . . . 4-2

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

6. TH E RMAL -H Y DR AU L IC DE S I GN . . . . . . . . . . . . . . . . . . . . . 6-1

7. ACCIDENT AND TRANSIENT ANALYSIS ................. 7-1 7.1. General Safety Analysis .................. 7-1

7. 2. Accident Evaluation .................... 7-1

8. PROPOSED MODIFICATION TO TECHNICAL SPECIFICATIONS ........ 8-1

9. STARTUP PROGRAM - PHYSICS TESTING ................ 9-1 9.1. Precritical Tests ..................... 9-1 9.1.1. Control Rod Trip Test ............... 9-1 9.1.2. Reacto r C ool a nt Fl ow . . . . . . . . . . . . . . . . 9-1

9. 2. Zero Power Physics Tests . . . . . . . . . . . . . . . . . . 9-2 9.2.1. Critical Boron Concentration . . . . . . . . . . . . 9-2

9. 2. 2. Temperature Reactivity Coefficient . . . . . . . . . 9-2 9.2.3. Control Rod Group Reactivity Worth . . . ,, . . . . . 9-2 9.2.4 Ejected Control RLd Reactivity Worth . . . . . . . . 9-3

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CONTENTS (Cont'd)

Page 9.3. Powe r E sc al a t io n Te s t s . . . . . . . . . . . . . . . . . . . 9-3 9.3.1. Core Power Distribution Verification at N40, s75, and s100 FP With Nominal Control Rod Position ... 9-3 9.3.2. Incore Versus Excore Detector Imbalance Correlation Verification at s4D% FP ........ 9-5 9.3.3. Temperature Reactivity Coefficient at s100% FP , . . 9-5 9.3.4 Power Doppler Reactivity Coefficient at s100% FP . . 9-5 9.4. Procedure for Use When Acceptance Criteria Are Not Met . . . 9-6 R EF E R E NC E S . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-1 List of Tables Table 4-1. F uel Des i gn Pa ramete rs . . . . . . . . . . . . . . . . . . . . . 4-4 4-2. Fuel The rmal Analysi s Paramete rs . . . . . . . . . . . . . . . . 4-5 5-1. Davis-Besse Unit 1, Cycle 4 Physics Parameters . . . . . . . . . 5-3 5-2. Shutdown Margin Calculation for Davis-Besse Unit 1. Cycle 4 .. 5-5 6-1. Davis-Besse Cycles 3 and 4 Thermal-Hydraulic Design Conditions ....................... 6-2 7-1. Comparison of Key Parameters for Accident Analysis . . . . . . . 7-3 7-2. Bounding Values for Allowable LOCA Peak Linear Heat Rates ... 7-3 8-1. Reactor Protection System Instrumentation Trip Setpoints . . . . 8-2 List of Fiqures Figure 3-1. Davis-Besse Cycle 4 Full Core Loading Diagram ......... 3-3 3-2. Enrichment and Burnup Distribution for Davis-Besse Unit 1, Cycle 4 ............................ 3-4 3-3. Control Rod Locations for Davis-Basse Unit 1, Cycle 4 ..... 3-5 5-1. BOC (4 EFPD), Cycle 4 Two-Dimensional Relative Power Distribution - Full Power, Equilibrium Xenon, APSRs Inserted . . 5-6 8-1. Reactor Core Safety Limit ................... 8-11 8-2. Trip Setpoint for Flux - AFlux/ Flow .............. 8-12 8-2a. Reactor Core Safety Limit ................... 8-12a 6-3. Regulating Group Position Limits, O to 24+10/-0 EFPD, Four RC Pumpf Davis-Besse 1, Cycle 4 ............... 8-13 8-4 Regulattte Group Position Limits, 24+10/-0 to 15010 EFPD, Four RC Pumps - Davi s-Besse 1, Cycle 4 . . . . . . . . . . . . . 8-14

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Figures (Cont'd)

Figure Page 8-5. Regulating Group Position Limits After 150t10 EFPD, Four RC Pumps - Davis-Besse 1, Cycle 4 ............ 8-15 8-6. Regulating Group Position Limits, O to 24+10/-0 EFPD, Three RC Pumps - Davi s-Besse 1, Cycl e 4 . . . . . . . . . . . . 8-16 8-7. Regulating Group Position Limits, 24+10/-0 to 150t10 EFPD, Three RC Pumps - Davi s-Besse 1, Cycl e 4 . . . . . . . . . . . . 8-17 8-8. Regulating Group Position Limits After 150t10 EFPD, Three RC Pumps - Davi s-Besse 1, Cycle 4 . . . . . . . . . . . . 8-18 8-9. APSR Position Limits, O to 24+10/-0 EFPD, Four RC Pumps -

Davis-Besse 1, Cycle 4 .................... 8-19 8-10. APSR Position Limits, 24+10/-0 to 15010 EFPD, Four RC Pumps - Davis-Besse 1, Cycle 4 ................ 8-20 8-11. APSR Position Limits After 150t10 EFPD, Four RC Pumps -

Davis-Besse 1, Cycle 4 .................... 8-21 8-12. APSR Position Limits, 0 to 24+10/-0 EFPD, Three RC Pumps -

Davis-Besse 1. Cycle 4 .................... 8-22 8-13. APSR Position Limits, 24+10/-0 to 150t10 EFPD, Three RC Pumps - Davis-Besse 1, Cycle 4 ................ 8-23 8-14. APSR Position Limits After 150t10 EFPD, Three RC Pumps -

Davis-Besse 1, Cycle 4 .................... 8-24 8-15. Axial Power Imbalance Limits, O to 24+10/-0 EFPD, Four RC Pumps - Davi s-Bessa 1, Cycl e 4 . . . . . . . . . . . . . . . 8-25 8-16. Axial Power Imbalance Limits, 24+10/-0 to 15010 EFPD, Four RC Pumps - Davis-Besse 1, Cycle 4 ............ 8-26 8-17. Axial Power Imbalance Limits After 150t10 EFPD, Four RC Pumps - Davis-Besse 1, Cycle 4 ............ 8-27 8-18. Axial Power Imbalance Limits, 0 to 24+10/-0 EFPD, Three RC Pumps - Davi s-Besse 1, Cycl e 4 . . . . . . . . . . . . 8-28 8-19. Axial Power Imbalance Limits, 24+10/-0 to 150t10 EFPD, Three RC Pumps - Davi s-Besse 1, Cycl e 4 . . . . . . . . . . . . 8-29 8-20. Axial Power Imbalance Limits After 150t10 EFPD, Three RC Pumps - Davi s-Besse 1, Cycl e 4 . . . . . . . . . . . . . . . 8-30 8-21. Control Rod Core Locations and Group Assignments -

Davis-Besse 1, Cycle 4 .................... 8-31 9

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1. INTRODUCTION AND

SUMMARY

This report justifies operation of the Davis-Besse Nuclear Power Station Unit 1 at the rated core power of 2772 MWt for cycle 4 The required analyses are included as outlined in the Nuclear Regulatory Commission (NRC) document,

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

This report utilizes the analytical techniques and design bases documented in several reports that-have been submitted to the NRC and approved by that agency.

Cycle 4 reactor and fuel parameters related to power capability are summar-ized in this report and compared to cycle 3. All accidents analyzed in the Davis-Besse Final Safety Analysis Report (FSAR) have been reviewed for cycle 4 operation, and in cases where cycle 4 characteristics were conservative compared to cycle 1, no new analyses were perfonned.

Retainersl and neutron sources will remain in the core. The effects on con-tinued operation without orifice rod assemblies (0 ras) and with the retainers have been accounted for in the analysis performed for cycle 4 The Technical Specifications have been reviewed and modified where required for cycle 4 operation. Based on the analyses performed, taking into account the emergency core cooling system (ECCS) Final Acceptance Criteria and post-ulated fuel densification effects, it is concluded that Davis-Besse Unit 1, cycle 4 can be operated safely at its licensed core power level of 2772 MWt.

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2. OPERATING HISTORY The reference cycle for the nuclear and thermal-hydraulic analyses of Davis-Besse Unit 1 is the currently operating cycle 3, which achieved criticality on August 29, 1982. Power escalation began on Septerber 1,1982 and full power (2772 MWt) was reached on October 29, 1982. During cycle 3 operation, no operating anomalies occurred that would adversely affect fuel performance during cycle 4. The duration of cycle 3 and .the planned duration of cycle 4 are 268 and 240 effective full power days (EFPD) respectively.

A quadrant power tilt that was larger than that experienced in previous cy-cles was measured at the beginning of cycle 3 in quadrant WX. To reduce the potential for tilt amplification, the cycle 4 design minimizes the number of assemblies that are cross-core shuf fled. The cycle 4 shuffle pattern is discussed in section 3.

The axial power shaping rods (APSR) were pulled at 200 EFPD to increase the lifetime of cycle 3. The APSR pull coupled with a power coastdown will re-sult in a cycle 3 length of approximately 268 EFPD.

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l 3 GENERAL DESCRIPTION Th2 Davis-Besse Unit I reactor core is described in detail in chapter 4 of th2 FSAR2 for the unit. The cycle 4 core consists of 177-FAs, each of which is a 15x15 array containing 208 fuel rods,16 control rod guide tubes, and enn incore instrument guide tube. All FAs in batches 4, 5, and 6 have a con-stant nominal fuel loading of 468.25 kg of uranium. Batches 10 and 28 have a I fuel loading of 472.24 kg of uranium. The fuel consists of dished-end cylin-

! drical pellets of uranium dioxide clad in cold-worked Zircaloy 4. The undens-ified nominal active fuel lengths, theoretical densities, fuel and fuel rod dimensions, and other related fuel parameters may be found in Tables 4-1 and 4-2 of this report.

Section 2 addresses the tilt amplification that occurred in cycle 3. Refer-snce 3 provides guidelines for a fuel shuffle method that reduces the number of assemblies that are cross-core shuffled. The cycle 4 design expanded upon this method so that only eight fuel assemblies are cross-core shuffled. This will minimize any carryover effects from tilts in previous cycles.

Figure 3-1 is the core loading diagram for Davis-Besse Unit 1, cycle 4. Twen-ty-five batch IC assemblies and 60 batch 3 assemblies will be discharged at

- the end of cycle 3. The batch 4 and 5 assemblies will be shuffled to their cycle 4 locations. Batches 4 and SA have an initial uranium-235 enrichment of 3.04 wt %. Batch 5B has an initial enrichment of 2.99 wt %. Seventeen batch ID assemblies with an initial enrichment of 1.98 wt % and 20 batch 28 assemblies with an initial enichment of 2.63 wt % will be reinserted in cycle 4 A feed batch consisting of 48 batch 6 assemblies with uranium enrichment of 2.99 wt % will be inserted in cycle 4 and occupy the periphery of the

- ccre. Figure 3-2 is a quarter-c' ore map showing each assembly's burnup at the bsginning of cycle '(B0C) 4 and its initial enrichment.

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l Cycle 4 is operated in a feed-and-bleed mode. The core reactivity control is supplied mainly by soluble boron and supplemented by 53 full-length Ag-In-Cd In addition to the full-length control rods, l control rod assemblies (CRAs).

eight axial power shaping rods (APSRs) are provided for additional control of l the axial power distribution. The cycle 4 locations of the 61 control rods and the group designations are indicated in Figure 3-3. The core locations ,

I of 61 control rods for cycle 4 are identical to those of reference cycle 3.

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Figure 3-1. Davis-Besse Cycle 4 Full Core Loading Diagram FUEL TRANSFER CANAL )

x 1

6 6 6 6 6 A

6 6 6 28 58 28 6 6 6 8 L4 R8 L12

, Cy2 Cy2 6 58 4 10 58 4 58 1D 4 58 6 C M2 K3 08 N3 F10 N13 F8 K13 M14 Cy1 evi 6 58 58 4 28 4 58 4 28 4 58 58 6 0 811 A7 L5 C6 05 08 011 C10 L11 G15 85 Cy2 Cy2 6 4 4 58 4 SA 4 SA 4 58 4 4 6 E

C9 E10 G1 86 L1 L8 L15 810 A9 E6 C7 6 6 10 28 4 4 10 58 10 4 4 28 10 6 6 p H6 F3 F2 G9 ES 812 G9 K9 F14 F13 H12 Cy1 Cy2 Cy1 Cy1 Cy2 Cy1 6 28 58 4 5A 10 58 28 58 10 SA 4 58 28 6 G D10 C12 E4 A10 G7 84 C11 014 E11 A6 E12 C4 D6 Cy2 Cy1 Cy2 Cy1 Cy2 6 58 4 58 4 58 28 10 28 58 4 58 4 58 6

% H15 F6 H4 H10 02 [3 L6 M13 N14 H6 H12 L10 H1 Cy2 Cy1 Cy2 6 28 58 4 5A 10 58 2B $8 10 SA 4 58 28 6 K N10 012 M4 R10 MS N2 05 P12 K9 R6 Mit 04- N6 Cy2 Cy1 Cy2 Cy1 Cy2 6 6 10 28 4 4 10 58 10 4 4 28 10 6 6 t H4 L3 L2 G7 K7 P4 M11 K7 L14 L13 H10 Cy1 Cy2 Cy1 Cy1 Cy2 Cy1 6 4 4 58 4 5A 4 SA 4 58 4 4 6 M 09 M10 R7 P6 F1 F8 FIS P10 K15 M6 07 6 58 58 4 28 4 58 4 28 4 58 58 6 N fil K1 F5 06 N5 N8 Mll 010 Fil R9 PS Cy2 Cy2 6 58 4 10 58 4 58 10 4 58 6 0 E2 G3 L8 03 L6 D13 N8 G13 E14 Cy1 Cy1 6 6 6 28 58 28 6 6 6 P F4 A8 F12 Cy2 Cy2 6 6 6 6 6 A

I 2

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

$y =S5nIe tftd r5mcycle1

- Cy2 = reinserted fras cycle 2 i

4 s

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

1 Figure 3-2. Enrichment and Burnup Distribution for Davis-Besse Unit 1, Cycle 4 8 9 10 11 12 13 14 15 1 1.98 2.63 2.99 3.04 2.99 3.04 2.99 2.99

" 12,644 24,006 6,602 19,798 11.579 19,764 8,340 0 2.63 2.99 1.98 3.04 3.04 2.99 2.63 2.99 24,006 6,586 13,254 6,533 19,646 10,184 23,008 0

- 2.99 1.98 3.04 3.04 2.63 1.98 2.99 2.99 6,602 12,983 19,670 16,248 23,499 13,416 0 0 3.04 3.04 3.04 2.99 3.04 3.04 2.99 19,828 6,527 16,237 8,012 17,950 21,080 0 2.99 3.04 2.63 3.04 2.99 2.99 2.99 N

11,568 19,617 23,501 17,930 8,007 8,906 0 l

l 3.04 2.99 1.98 3.04 2.99 2.99 0

! 19,764 10,141 13,560 21,087 8,889 0 l

2.99 2.63 2.99 2.99 2.99 1 8,335 23,007 0 0 0 R l 0 0 0

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x. x x Initial enrichment l

xx.xxx- BOC burnup, MWt 3-4

Figure 3-3. Control Rod Locations for Davis-Besse Unit 1, Cycle 4 X

l A

i 8 4 7 4 I 8 8 3 4 C 8 5 8 7 0 7 5 5 1 E 1 7 3 8 4 F 4 8 3 2 2 8 G 8 5 7 -Y HW- 7 5 7 3 7 2 2 8 K 6 3 8 4 L 4 8 3 7 5 1 N 1 5 5 8 7 N 7 8 0 1 S 6 i P 4 7 4 R

I 9 10 11 12 13 14 15 I 2 3 4 5 6 7 8 GROUP NO. OF RODS FUNCTIONS 8 SAFETY X- - GROUP NUNGER 1 2 4 SAFETY 3 5 SAFETY 4 8 SAFETY 5 8 CONTROL 6 8 CONTROL 7 12 CONTROL 8 l APSRs TOTAL

• 61 3-5

4. FUEL SYSTEM DESIGN 4.1. Fuel Assembly Mechanical Design The types of FAs and pertinent fuel parameters for Davis-Besse Unit 1, cycle 4 are listed in Table 4-1. All Mark-B (Mk-B) FAs are identical in concept and are mechanically interchangeable. Retainer assemblies will be used on two FAs that contain the regenerative neutron sources. The justification for the design and use of retainer assemblies is described in references 1 and 4 4.2. Fuel Rod Design The fuel rod design and mechanical evaluation are discussed below.

4.2.1. Cladding Collapse Due to its previous incore exposure time, the fuel of batch 2B is more limit-ing than batches 10, 4, SA, SB, and 6. The batch 2B assembly power histories were analyzed to detennine the most limiting three-cycle power history for creep collapse. This power history was compared to a generic analysis to en-sure that creep ovalization will not affect the fuel performance during Davis-Besse Unit 1, cycle 4 The generic analysis was based on reference 5 and is applicable to the batch 2B design.

The creep collapse analysis (Table 4-1) predicts a collapse time longer than 35,000 effective full power hours (EFPH), which is longer than the expected residence time of 21,840 EFPH.

4.2.2. Cladding Stress The Davis-Besse Unit 1, cycle 4 stress parameters are enveloped by a conserva-tive fuel rod stress analysis. The methods used for the analysis of cycle 4 have been used in the previous cycles.

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4.2.3. Cladding Strain The fuel design criteria specify a limit of 1.0% on cladding plastic tensile circumferential strain. The pellet is designed to ensure that plastic clad-ding strain is less than 1% at design local pellet burnup and heat generation rate. The design values are higher than the worst-case values the Davis-Besse Unit 1, cycle 4 fuel is expected to see. The strain analysis is also based on the upper tolerance values for the fuel pellet diameter and density, and the lower tolerance for the cladding inside diameter (ID).

4.3. Thernal Design All fuel in the cycle 4 core is thermally similar. The cycle 4 thernal analy-ses represent a change in analytical method. The analyses for the inconing batch 6 fuel have been perforned with the TAC 026 code using the analysis methodology described in reference 7. This nethodology uses nominal undensi-fied input parameters provided in Table 4-2. Densification effects are ac-counted for in the TAC 02 densificatior, aodel. The TAC 02 analyses also apply to the hatch 5B fuel since this fuel is identical in design to the batch 6 fuel. Reinserted FAs fron batches 10, 2B, 4, and SA were evaluated using TAFY38 analyses perforned for prior cycles.

The thermal design evaluation for the cycle 4 core is summarized in Table 4-2. Linear heat rate (LHR) capabilities are based on centerline fuel melt (CFM) with core protection limits based on a 20.4 kW/f t LHR to CFM. The TAC 02 analyses performed for batches 5B and 6 demonstrate that 20.5 kW/ft is the CFM limit for this fuel. Using TAFY3, the fuel internal pressure has been evaluated for the highest burnup fuel rod and is predicted to be less than the noninal reactor coolant system pressure of 2200 psia. The naxinun burnup of any fuel rod during cycle 4 is less than 42,000 mwd /ntU.

4.4 Material Conoatibility The ca1patibility of all possible fuel cladding - coolant assembly interac-tions for batch 6 FAs is identical to that of present fuel.

4.5. Operatina Experience Operating experience with the Mark-B 15x15 FA has verified the adequacy of its design. As of February 28, 1983, the following experience has been ac-cumulated for eight Babcock & Wilcox (88W) 177-FA plants using the Mark-B FA:

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Max FA burnup,(a) mwd /mtU Cumulative net Current electric Reactor cycle Incore Discharged output,(b) MWh Oconee 1 7 48,010 40,000 41,241,515 Oconee 2 6 27,240 36,800 36,475,957 Oconee 3 7 22,975 35,450 37,022,597 Thrce Mile 5 25,000 32,400 23,840,053 Island 1 Arkarisas Nuclear 6 23,160 36,540 34,949,454 Unit 1 Rancho Seco 5 37,883 37,730 29,933,402 Crystal River 3 4 28,110 29,900 22,081,044 Davis-Besse 3 28,820 25,326 12,898,260

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(a)As of February 28, 1983.

(b)As of October 31, 1982.

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

l Table 4-1. Fuel Design Parameters Batch 28 4 5A 58 6 10 Mk-84A Mk-B4A Mk-B4A Mk-B4A Mk-B4A Mk-B4A F A type 17 20 44 8 40 48 Number of assemblies t Fuel rod OD, in. 0.430 0.430 0.430 0.430 0.430 0.430 l Fuel rod ID, in. 0.377 0.377 0.377 0.377 0.377 0.377 ,

Spri ng Spri ng Spring Spri ng Spri ng Spring Flexible spacer type Rigid spacer type Zr-4 Zr-4 Zr-4 Zr-4 Zr-4 Zr-4 Undensified active 143.44 143.20 143.20 fuel length, in. 143.5 143.5 143.44 Fuel pellet (mean) 0.3686 dia. , in. 0.3675 0.3675 0.3697 0.3697 0.3686 Fuel pellet initial 96 96 94 94 95 95 density, % TD mean Initialwtfuel ment,  % 23Inrich-O 1.98 2.63 3.04 3.04 2.99 2.99 Estimated residence 12,000 12,000 5,760 time, EFPH 14,736 21,840 19,104 Cladding collapse >3 5,000 >3 5,000 time, EF PH >35,000 >35,000 >3 5,000 >3 5,000 J

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Table 4-2. Fuel Thermal Analysis Parameters Batch 1D/2B 4/5A SB 6 Nunber of assemblies 17/20 44/8 40 48 Initial density, % TD 96 94 95 95 Pellet diameter, in. 0.3675 0.3697 0.3686 0.3686 Nominal stack height, in. 143.5 143.44 143.2 143.2 Enrichnent, wt % 235U 1.98/2.63 3.04 2.99 2.99 LHR capability, kW/ft to CFM 20.4 20.4 20.5 20.5

- Densified fuel parameters (a)

TAFY3 Code Analysis Only Pellet diameter, in. 0.3651 0.3648 0.3649(b) 0.3649(b)

Fuel stack height, in. 143.14 141.65 142.13 142.13 Average fuel temperature, F 1340 1355 1464(c) 1464(c)

Nominal LHR, kW/ft at 2772 MWt 6.14 6.21 6.19 6.19 l

(a)Densification to 96.5% TD assumed for TAFY3 analysis.

(b)This data is provided for comparative purposes only and does not represent parameter values used in TACO 2 analyses.

(c)BOL, TAC 02 code.

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5. NUCLEAR DESIGN 5.1. Physics Characteristics Table 5-1 compares the core physics parameters of cycle 3 with those of cycle

4. These values were generated using PDQ07 9 -11 for both cycles. Since the ccre has not yet reached an equilibrium cycle, differences between the cycles in core physics parameters are to be expected. Figure 5-1 illustrates a rep-r:sentative relative power distribution for the BOC at full power (FP) with equilibrium xenon and group 8 inserted.

Due to the difference in design cycle lengths, the critical boron concentra-tiens for cycle 4 differ from those of reference cycle 3. Because of differ-ent isotopic distributions, cycle 4 control rod worths, ejected rod worths, and stuck rod worths differ from those of cycle 3. 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 all times in life and at all power levels in the development of the rod position limits presented in s:ction 8. The adequacy of the shutdown margin with cycle 4 rod worths is shown in Table 5-2. The following conservatisms were applied for the shut-down calculations:

1. Poison material depletion allowance.

2. 10% uncertainty on net rod worth.

3. Flux redistribution penalty.

Flux redistribution was taken into account since the shutdown analysis was calculated using a two-dimensional model. The cycle 4 moderator coefficients and the power deficits from hot zero power (HZP) to hot full power (HFP) are similar to those for cycle 3. The differential boron and xenon worths are also similar in both cycles. The effective delayed neutron fraction for cy-cle 4 show a decrease with burnup (also shown in reference cycle 3).

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5.2. Changes in Nuclear Design There are no significant core design changes between the reference cycle and the cycle 4 designs, although the cycle 4 core was shuffled in a manner to minimize the carryover effect on quadrant tilt. The same calculational meth-ods and design information were used to obtain the important nuclear design parameters. No significant operational or procedural changes exist with re-7 gard to axial or radial power shape, xenon, or tilt control.

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Table 5-1. Davis-Besse Unit 1, Cycle 4 Physics Parameters Cycle 3 Cycle 4 Cycle length, EFPD(a) 268 240 Cycle burnup, mwd /mtU 8,929 8,014 Average core burnup - E0C(b), mwd /mtU 20,191 18,924 Initial core loading, mtU 83.2 83.0 Critic boron - B0C, No Xe, ppm HZP 1 1,231 1,250 HFP J Group 8 inserted 1,015 1,042 Criticgl)

HZPld boron - EOC, Eq. Xe, ppm 284 337 HFP(d) 10 42 Control rod worths - HFP, BOC, % ak/k Group 6 0.93 1.02 Group 7 1.52 1.73 Group 8 0.31 0.27 Control rod worths - HFP, E0C % ak/k Group 7 1.53 1.74 Group 8 NA 0.35 Max ejected rod worth - HZP, % ak/k (location)

B0C O.78 0.85(e)

Groups 5-8 inserted (N-12) (N-12)

EOC(d)

Groups 5-7 inserted 0.72 0.85 (N-12) (N-12)

Max stuck rod worth - HZP, % ak/k (location)

BOC 1.44 1.70 (N-12) (L-14)

E0C 1.'25 1.54 (L-14) (L-14)

Power deficit, HZP to HFP, Eq. Xe, % ak/k B0C (4 EFPD) -1.79 -1.77 EOC -2.34 -2.36 Doppler coeff - HFP,10-5 ak/k/ F BOC , No Xe,1042 ppn, Group 8 inserted -1.46 -1.47 E0C, Eq. Xe,10 ppm, Group 8 inserted -1.63 -1.58 Moderator coeff - HFP,10-4 ak/k/ F BOC, No Xe,1042 ppm, Group 8 inserted -1.13 -1.00 E0C, Eq Xe,10 ppm, Group 8 inserted -2.89 -2.87 Boron worth - HFP, ppm /% ak/k BOC (1042 ppab) 110 108 E0C (10 ppn) 97 97 Xenon worth - HFP, % ak/k BOC (4 EFPD) 2.67 2.67

E0C (equilibrium) 2.73 2.74 5-3

-. . -. . . - - = . . .

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l Table 5-1. (C ont 'd) l l

Cycle 3 Cycle 4 Effective delayed neutron fraction - HFP BOC 0.00595 0.00598 4 E OC 0.00530 0.00538 l l

(a)EFPD denotes effective full power days.

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(b)EOC denotes end of cycle.

(c)HZP denotes hot zero power (532F T a vg); HFP denotes hot full power (584F core Ta vg)*

(d) Group 8 is withdrawn at E0C 3 and inserted at E0C 4.

(e) Ejected rod worth at the rod insertion limit.

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Table 5-2. Shutdown Margin Calculation for Davis-Besse Unit 1, Cycle 4 BOC , E0C ,

% ak/k % ak/k Available Rod Worth Tctal rod worth, HZP(a) 7.69 7.89 Worth reduction due to burnup of -0.16 -0.19 p31 son material Maximum stuck rod, HZP -1.70 -1.54 Net worth 5. 83 6.16 Less 10% uncertainty -0.58 -0.62 Total available worth 5.25 5.54 Required Rod Wortn Power deficit, HFP(a) to HZP 1.77 2.36 Max allowable inserted rod worth 0.51 0.64 Flux redistribution 0.73 1.15 Total required worth 3.01 4.15 Shutdown Margin t

Total available minus total required 2.24 1.39 Nite: Requi red shutdown margin is 1.00% ak/k.

. (a)HZP denotes hot zero power (532F Tavg); HFP deno:es hot full power (584F core Tavg)-

5-5

l 1

Figure 5-1. BOC (4 EFPD), Cycle 4 Two-Dimensional Relative Power Distribution Xenon, APSRs - F(asuT1 Power, Equilibrium Inserted 1

8 9 10 11 12 13 14 15 l l

l 0.816 0.885 1.293 1.142 1.179 1.043 1.155 0.986 H

l l

K 0.889 1.206 0.982 1.286 1.000 1.108 0.909 0.937 .

8 L 1.298 0.988 1.034 1.045 0.725 0.829 1.240 0.774 N 1.143 1.285 1.040 1.155 0.%7 0.909 1.022 N 1.180 1.001 N8 0.724 0.967 1.184 1.029 0.714 0 1.004 1.109 0.829 0.911 1.035 0.811 P 1.156 0.910 1.241 1.024 0.715 l

0.986 0.937 0.775 1

1 Inserted rod group number x,x x Relative power density 9

(a) Calculated results from two-dimensional pin-by-pin PDQ07 .

5-6

I l

l

6. THERMAL-HYDRA'!LIC DESIGN The fresh batch 6 fuel is hydraulically and geometrically similar to the

'other fuel loaded into the cycle 4 core. The thermal-hydraulic design eval-uation-supporting cycle 4 operation is based on the methods and models de-scribed in references 13 and 14 The cycle 4 thermal-hydraulic design is idontical to that of cycle 3. The thermal-hydraulic design conditions for cy-cles 3 and 4 are summarized in Table 6-1.

Previous fuel cycle evaluations included the calculation of a rod bow penalty for each fuel batch based on the highest fuel rod burnup in that batch. A rod bow topical report 15, 'which addresses the mechanisms and resulting local conditions of rod bow, has been submitted to and approved by the NRC. The topical report concludes that rod bow penalty is insignificant and is offset by the reduction in power production capability of the FAs with irradiation.

< Therefore, no departure from nucleate boiling ratio (DNBR) reduction due to fuel rod bow need be considered for cycle 4.

6-1

I Table 6-1. Davis-Besse Cycles 3 and 4 Thermal- l Hydraulic Design Conditions l

Design power level . MWt 2772 2200 System pressure, psia Reactor coolant flow, % design 110 i

Vessel inlet / outlet coolant temp.,100% power F 557.7/606.3 Ref design radial-local power peaking factor 1.71 Ref design axial flux shape 1.5 cosine with tails Hot channel factors 1.011 Enthalpy rise q Heat flux (Fy)(F ) 1.014 Flow area 0.98 l Active fuel length See Table 4-2 Avg heat flux,100% power, Btu /h-ft2 1.89 x 105 (a)

Max heat flux,100% power, Btu /h-ft 2 4.85 x 105(a)

BAW-2 Critical heat flux (CHF) correlation Minimum DNBR, (% power) 1.79(112%)

(a)With thermally expanded fuel rod OD of 0.43075 inch.

I i

l l

6-2

7. ACCIDENT AND TRANSIENT ANALYSIS 7.1. General Safety Analysis Each FSAR2 accident analysis has been examined with respect to changes in the cycle 4 parameters to determine the effects of the cycle 4 reload and to en-sure that themal perfomance during hypothetical transients is not degraded.

The effects of fuel densification on the FSAR accident results have been eval-uated and are reported in reference 13.

The radiological dose consequences of the FSAR chapter 15 accidents based on cycle 4 iodine and noble gas inventories have been evaluated. These doses are either bounded by the FSAR values or are a small fraction of the 10 CFR 100 limits.

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 re-activity feedback coefficients and control rod worths.

Fuel thermal analysis parameters from each batch in cycle 4 are given in Tabl e 4-2. A comparison of the cycle 4 thermal-hydraulic maximum design con-ditions to the previous cycle values is presented in Table 6-1. A comparison of the key kinetics parameters from the FSAR and cycle 4 is provided in Table 7-1.

A generic loss-of-coolant accident (LOCA) analysis for B&W 177-FA raised-loop nuclear steam systems (NNSs) has been performed using the Final Acceptance Criteria ECCS Evaluation Model.16 The combination of average fuel tempera-ture as a function of linear heat rate (LHR) and the lifetime pin pressure data used in the LOCA limits analysis is conservative compared to those calcu-lated for this reload. Thus, the analysis and the LOCA limits reported in j 7-1

reference 16 provide conservative results for the operation of Davis-Besse Unit 1, cycle 4 fuel. A tabulation showing the bounding values for allowable LOCA peak LHRs for Davis-Besse Unit 1, cycle 4 fuel are provided in Table 7-2.

It is concluded by the examination of cycle 4 core thermal, thermal-hydrau-lic, and kinetics properties, with repsect to acceptable previous cycle val-ues, that this core reload will not adversely affect the ability to safely operate the Davis-Besse Unit 1 plant during cycle 4 Considering the previ-ously accepted design basis used in the FSAR and subsequent cycles, the tran-sient evaluation of cycle 4 is considered to be bounded by previously accepted analyses. The initial conditions of the transients in cycle 4 are bounded by the FSAR and/or the fuel densification report.

l 7-2

\

Table 7-1. Comparison of Key Parameters for Accident Analysis FSAR and l densif'n report Cycle 4 Pa rameter value value BOL(a) Doppler coef f,10-5, ak/k/'F -1.28 -1.47 E0L(b) Doppler coef f,10-5, ak/k/*F -1.45(C) -1.58 BOL moderator coef f,10-4, - ak/k/'F +0.13 -1.00 I E0L moderator coef f,10-4, ak/k/'F -3 . 0 -2.87 All rod bank worth (HZP), % ak/k 10.0 7.69 Boron reactivity worth (HFP), ppm /1% ak/k 100 108 Max ejected rod worth (HFP), % ak/k 0.65 0.46 Max dropped rod worth (HFP), % ak/k 0.65 0.20 Initial boron conc (HFP), ppm 1407 1042 (a)BOL denotes beginning of life.

(b)E0L denotes end of life.

(c)-1.77 x 10-5 Ak/k/ F was used for steam line failure analysis.

Table 7-2. Bounding Values for Allowable LOCA Peak Linear Heat Rates Allowable Allowable Core peak LHR, peak LHR, elevation, first 24 EFPD, balance of cycle, ft kW/ft kW/ft 2 15.5 16.5 4 16.8 17.2 6 18.0 18.4 8 17.5 17.5 10 17.0 17.0 7-3

8. PROPOSED MODIFICATION TO TECHNICAL SPECIFICATIONS The Technical Specifications have been revised for cycle 4 operation to ac-count for changes in power peaking and control rod worths. The effects of NUREG-0630 have been incorporated into the operating limits. Figures 8-1 through 8-20 are revisions to the previous cycle Technical Specifications.

Based on these Technical Specifications the final acceptance criteria ECCS limits will not be exceeded and the thermal design criteria will not be vio-lated. .

i 8-1

Table 8-1. Reactor Protection System Instrumentation Trip Setpoints Table 2.2-1 Functional unit Trip setpoint Allowable values

1. Manual reactor trip Not applicable. Not applicable.

2. High flux <104.94% of RATED THERMAL POWER with <104.94% of RATED THERMAL POWER with l Tour pumps operating Tour pumps operatingI '

<79.85% of RATED THERMAL POWER with <79.85% of RATED THERMAL POWER with three pumps operating Three pumps operatingI l

3. RC high temperature <618 F <618 F#

4. Flux -- aflux/fl owfl) Trip setpoint not to exceed the lim- Allowable values not to exceed the it line of Figure 2.2-1 limit line of Figure 2.2-II m m5 RC 1ow pressure (l) J_1983.4 psig 11983.4 psig* )_1983.4 psig**

r'o S

6. RC high pressure <2300 psig <2300.0 psig* <2300.0 psig**

7. RC pressure-temperature (l) 1(12.60 Tout F - 5662.2) psig }_(12.60 Tout F - 5662.2) psigI l

8. High flux {gumber of RC <55.1% of RATED THERMAL POWER with pumps on( i <55.1% of RATED THERMAL POWER with one pump operating in each loop one pump operating in each loopI

<0.0% of RATED THERMAL POWER with <0.0% of RATED THERMAL POWER with Two pumps operating in one loop and Iwo pumps operating in one loop and no pumps operating in the other loop no pumps operating in the other loopI

<0.0 of RATED THERMAL POWER with no <0.0% of RATED THERMAL POWER with no pumps operating or only one pump op- pumps operating or only one pump op-erating eratingI

9. Containment pressure high H psig <4 psigI

SAFETY LIMITS BASES ,

The reactor trip envelope appears to approach the safety limits more closely than it actually does because the reactor trip pressures are measured at a lo-cation where the indicated pressure is about 30 psi less than core outlet pressure, providing a more conservative margin to the safety limit.

The curves of Figure 2.1-2 are based on the more restrictive of two thermal limits and account for the effects of potential fuel densification and poten-tial fuel rod bow.

1.

The 2.561.30 DNBR or the limit produced combination of the by a nuclear radial peak, power peaking axial peak, andfactor of Fg position o =f the axial peak that yields no less than a 1.30 DNBR.

2. The combination of radial and axial peak that causes central fuel melting at the hot spot. The limits are 20.4 kW/ft for batches 10, 2B, 4 and 5A and 20.5 kW/ft for batches 5B and 6. I Power peaking is not a directly observable quantity and therefore limits have been established on the basis of the reactor power imbalance produced by the power peaking.

The specified flow rates for curves 1 and 2 of Figure 2.1-2 correspond to the expected minimum flow rates with four pumps and three pumps, respectively.

The curve of Figure 2.1-1 is the most restrictive of all possible reactor coolant pump-maximum thermal power combinations shown in BASES Figure 2.1.

The curves of BASES Figure 2.1 represent the conditions at which a minimum DNBR of 1.30 is predicted at the maximum possible thermal power for the num-ber of reactor coolant pumps in operation or the local quality at the point of minimum DNBR is equal to +22%, whichever condition is more restrictive.

These curves include the potential effects of fuel rod bow and fuel densifica-l tion.

The DNBR as calculated by the B&W-2 DNB correlation continually increases from point of minimum DNBR, so that the exit DNBR is always higher. Extrapo-lation of the correlation beyou; :ts published quality range of +22% is justi-fied on the basis of experirk eta da t a.

B 2-2 8-3

2.2. LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1. REACTOR PROTECTION SYSTEM INSTRllMENTATION SETPOINTS l

The reactor protection system instrumentation trip setpoints specified in Table 2.2-1 are the values at which the reactor trips are set for each param-eter. The trip setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their safety limits.

The shutdown bypass provides for bypassing certain functions of the reactor protection system in order to permit control rod drive tests, zero power PHYS-ICS TESTS and certain startup and shutdown procedures. The purpose of the shutdown bypass high pressure trip is to prevent normal operation with shut-down bypass activated. This high pressure trip setpoint is lower than the normal low pressure trip setpoint so that the reactor must be tripped before the bypass is initiated. The high flux trip setpoint of 15.0% prevents any significant reactor power from being produced. Sufficient natural circula-tion would be available to remove 5.0% of RATED THERMAL POWER if none of the reactor coolant pumps were operating.

Manual Reactor Trip The manual reactor trip is redundant channel to the automatic reactor protec-tion system instrumentation channels and provides manual reactor trip capabil-i ty.

High Flux A high flux trip at high power level (neutron flux) provides reactor core pro-tection against reactivity excursions which are too rapid to be protected by temperature and pressure protective circuitry.

During normal station operation, reactor trip is initiated when the reactor power level reaches 104.94% of rated power. Due to transient overshoot, heat balance, and instrument errors, the maximum actual power at which a trip would be actuated could be 112%, which was used in the safety analysis.

i l

B 2-4 8-4

LIMITING SAFETY SYSTEM SETTINGS BASES RC High Temperature The RC high temperature trip <618 F prevents the reactor outlet temperature from exceeding the design limTts and acts as a backup trip for all power ex-cursion transients.

t Flux -- AFlux/ Flow .

4 The power level trip setpoint produced by the reactor coolant system flow is based on a flux-to-flow ratio which has been established to accomrodate flow decreasing transients from high power where protection is not provided by the high flux / number of reactor coolant pumps on trips.

The power level trip setpoint produced by the power-to-flow ratio provides both high power level and low flow protection in the event the reactor power level increases or the reactor coolant flow rate decreases. The power level setpoint produced by the power-to-flow ratio provides overpower DNB protec-tion for all modes of pump operation. For every flow rate there is a maximum permissible power level, and for every power level there is a minimum permis-sible low flow rate. Examples of typical power level and low flow rate com-binations for the pump situations of Table 2.2-1 that would result in a trip are as follows:

1. Trip would occur when four reactor coolant pumps are operating if power is 106.9% and reactor coolant flow rate is 100% of full flow rate, or flow rate is 93.5% of full flow rate and power level is 100%.

2. Trip would occur when three reactor coolant pumps are operating if power is 79.9% and reactor coolant fl ow rate is 74.7% of full fl ow rate, or flow rate is 70.2% of full ficw rate and power is 75%.

For safety calculations the instrumentation errors for the power level were used. Full flow rate in the above two examples is defined as the flow calcu-lated by the heat balance at 100% power.

B 2-5 8-5

LIMITING SAFETY SYSTEM SETTINGS BASES The AXIAL POWER IMBALANCE boundaries are established in order to prevent reac-tor thermal limits from being exceeded. These thermal limits are either power peaking kW/ft limits or DNBR limits. The AXIAL POWER IMBALANCE reduces the power level trip produced by a flux-to-flow ratio such that the bounda-ries of Figure 2.2-1 are produced.

_R_C Pressure - Low, High, and Pressure Temperature l The high and low trips are provided to limit the pressure range in which reac-I tor operation is permitted.

During a slow reactivity insertion startup accident from low power or a slow reactivity insertion from high power, the RC high pressure setpoint is reached before the high flux trip setpoint. The trip setpoint for RC high pressure, 2300 psig, has been established to maintain the systen pressure be-low the safety limit, 2750 psig, for any design transient. The RC high pres-sure trip is backed up by the pressurizer code safety valves for RCS over pressure protection, and is therefore set lower than the set pressure for these valves, 2435 psig. The RC high pressure trip also backs up the high flux trip. .

The RC low pressure ,1983.4 psig, and RC pressure-temp.rature (12.60 tout -

5662.2) psig, trip setpoints have been established to maintain the DNB ratio greater than or equal to 1.30 for those design accidents that result in a pressure reduction. It also prevents reactor operation at pressures below the valid range of DNB correlation limits, protecting against DNB.

High Flux / Number of Reactor Coolant Pumps On In conjunction with the flux - Aflux/ flow trip the high flux / number of reac-tor coolant pumps on trip prevents the minimum core DNBR from decreasing below 1.30 by tripping the reactor due to the loss of reactor coolant pump (s) . The pump monitors also restrict the power level for the number of pumps in operation.

B 2-6 i

8-6

i REACTIVITY CONTROL SYSTEMS REGULATING R00 INSERTION LIMITS LIMITING CONDITION FOR OPERATION

.t 3.1.3.6 The regulating rod. groups shall be limited in physical insertion as shown on Figures 3.1-2a, 3.1-2b, 3.1-2c, 3.1-3a, 3.1-3b and 3.1-3c.

APPLICABILITY: MODES 1* and 2*#.

ACTION:

With the regulating rod groups inserted beyond the above insertion limits (in a region other than acceptable operation), or with any group sequence or over-lap outside the specified limits, except for surveillance testing pursuant to Specification 4.1.3.1.2, either:

a. Restore the regulating groups to within the limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or

b. Reduce THERMAL POWER to less than or equal to that fraction of RATED THER-MAL POWER which is allowed by the rod group position using the above fig-ures within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or

c. Be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

NOTE: If in unacceptable region, also see Section 3/4.1.1.1.

• See Special Test Exceptions 3.10.1 and 3.10.2.

1. With k eff 3 1.0.

DAVIS-BESSE, UNIT 1 3/4 1-26 8-7

REACTIVITY CONTROL SYSTEMS AXIAL POWER SHAPING R0D INSERTION LIMITS LIMITING CONDITION FOR OPERATION

-1 3.1.3.9 The axial power shaping rod group shall be limited in physical in-sertion as shown on Figures 3.1-Sa, 3.1-5b, 3.1-5c, 3.1-5d, 3.1-Se and 3.1-5f.

APPLICABILITY: MODES 1 and 2*.

> ACTION:

With the axial power shaping rod group outside the above insertion limits, either:

a. Restore the axial power shaping rod group to within the limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or

b. Reduce THERMAL POWER to less than or equal to that fraction of RATED -

THERMAL POWER which is allowed by the rod group position using the above figures within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or

c. Be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.1.3.9 The position of the axial power shaping rod group shall be deter-mined to be within the insertion limits at least once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> except when the axial power shaping rod insertion limit alarm is inoperable, then verify the group to be within the insertion limit at least once every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

• With k eff 2 1.0.

DAVIS-BESSE, UNIT 1 - 3/4 1-34 8-8

3 /4. 2 '. POWER DISTRIBUTION LIMITS AXIAL POWER IMBALANCE LIMITING CONDITION FOR OPERATION 3.2.1 AXIAL POWER IMBALANCE shall be maintained within the limits shown on Figures 3.2-la, 3.2-1b, 3.2-Ic, 3.2-2a, 3.2-2b and 3.2-2c.

APPLICABILITY: MODE 1 above 40% of RATED THERMAL POWER.*

ACTION:

With AXIAL POWER IMBALANCE exceeding the limits specified above, either:

a. Restore the AXIAL POWER IMBALANCE to within its limits within 15 minutes, or I

b. Within one hour reduce power until imbalance limits are met or to 40% of RATED THERMAL POWER or less.

SURVEILLANCE REQUIREMENTS 4.2.1. The AXIAL POWER IMBALANCE shall be determined to be within limits at least once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when above 40% of RATED THERMAL POWER except when the AXIAL POWER IMBALANCE alarm is inoperable, then calculate the AXIAL POWER IMBALANCE at least once per hour.

1

(

• See Special Test Exception 3.10.1.

I DAVIS-BESSE, UNIT 1 3/4 2-1 l 8-9 l .

POWER DISTRIBUTION LIMITS' BASES FN Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio AH of the integral of linear power along the rod on which minimun DPER occurs to the average rod power.'

It has been detennined by extensive analysis of possible operating power shapes that the design limits on nuclear power peaking and on minimum DNBR at '

full power are met, provided:

Fn < 2.93; FN < 1.71 AH -

Power peaking is not a directly observable quantity and therefore limits have been established on the bases of the AXIAL POWER IMBALANCE produced by the power peaking. It has been detemined that the above hot channel factor lim-its will be met provided the following conditions are maintained.

1. Control rods in a single group move together with no individual rod in-sertion differing by more than 6.5% (indicated position) from the group average height.

2. Regulating rod groups are sequenced with overlapping groups as required in Specification 3.1.3.6.

3. The regulating rod insertion limits of Specification 3.1.3.6 are main-tained.

4. AXIAL POWER IMBALANCE limits are maintained. The AXIAL POWER IMBALANCE is a measure of the difference in power between the top and bottom halves of the core. Calculations of core average axial peaking factors for cany plants and measurements from operating plants under a variety of operat-ing conditions have been correlated with AXIAL POWER IMBALANCE. The cor-relation shows that the design power shape is not exceeded if the AXIAL POWER IMBALANCE is maintained between the limits specified in Specifica-

, tion 3.2.1.

l l The design limit power peaking factors are the most restrictive calculated at full power for the range from all control rods fully withdrawn to minimum al-lowable control rod insertion and are the core DFBR design basis. Therefo re, for. operation at a fraction of RATED THERMAL POWER, the design limits are met. When using incore detectors to make power distribution maps to deter-N.

mine FO and F AH

a. The measurement of total peaking factor Fheas , shall be increased by 1.4 percent to account for manufacturing tolerances and further increased by 7.5 percent to account for measurement error.

B 3/4 2-2 l 8-10 l

Figure 8-1. Reactor Core Safety Limit 1 (Tech. Spec. Figure 2.1-2) 1

% RATED THERMAL POWER L

, 4 PUMP . .l20 (33,112)

(-31,112) LIMIT

( -46,100) .

100 (40,100)

(-31,89.24) (33,89.24) 3 PUMP I

- 80

(-46,77.24) (

ACCEPTABLE

- 60 UNACCEPTABLE OPERATION UNACCEPTABLE OPERATION FOR SPECIFIED OPERATION RC PUMP COMBINATION 40 20 t t i I I

-60 -40 -20 0 20 40 60 Axial Power imbalance, %

PUMPS OPERATING REACTOR COOLANT FLOW, GPM 4 387,200 3 290,l00 8-11

Figure 8-2. Trip Setpoint for Flux - AFlux/ Flow (Tech. Spec Figure 2.2-1)

Curve shows trip setpoint for a 25%

flow reduction for three pump operation (290,l00 gpm). The actual setpoint wiil be directly proportional to the actual flow with three pumps.

% RATED THERMAL POWER UNACCEPTABLE

- 10 OPERATION UNACCEPTABLE TION (-14,106.9) (17,106 9) l M2 =-l .8576 Mg =0.9288 - 100 l i PUMP LIMIT l

(-32,90 182) (26,90.182) l I(-l 4,79.85) __

(17,7985)

I

! 3 PUMP l l LIMIT (26,63.132)

(-32,63.132) - 60 l I

!ACCEPTABLEOPERATIONFORl l

l SPECIFIED RC PUMP COMB-

_ _ go l lINATION I I I I

I I J I

N * -

- 20 l

- l

?l l

="l i

hlja l "w>

li I" , ,

' i l i l

-60 -40 -20 0 20 40 60 Axial Power imbalance, %

8-12

Figure 8-2a. Reactor Core Safety Limit (Tech Spec Figure 2.1-1) 2600 - .

2400 -

RC HIGH PRESSURE TRIP 618,2300 RC HIGH TEMPERATURE S 2200 -

TRIP 618,2124.6 I ACCEPfABLE OPERATION RC PRESSURE TEMPERATURE TRIP 2000 - 606.79, 1983.4 _ SAFETY LIMIT RC LOW PRESSURE TRIP 1800 I i I I I I 590 600 610 620 630 640 Reactor Outlet Tergerature *F 8-12a

This page intentionally left blank 8-12b

~

l Figura 8-3. R gulating Group Position Liaits, O to 24+10/-0

' EFPD, Four RC Pumps - Davis-Basse 1. Cycle 4 (Tech. Spec. Figure 3.1-2a)

(246,102 (275,1'02)

(300,102)

_ Power Level 00 Cutoff = 100%

- SHUTDOWN MARGIN (268,92)

-l LIMIT l { (265,80) 80 -

l [

a

i. f g .

5 S

e-60 -

UNACCEPTABLE OPERATION

[*

I (185,50) @ (22550)

? -

o

= 90 -

ACCEPTABLE

! y OPERATION E

s.

y 20 -

2 (il2,15) .

)

,2.4) (iOO,0) , ,

0<

1 03 200 300 F RodIndex(f. Withdrawn) l ,g i !g GR 5 .

0 75 100

{ {f jgf Get 63 i y _y 0 25 75 100 GR 71 l .i i

0 25 100 I

Figure 8-4. Regulating Group Position Limits, 24+10/-0 to 150110 EFPD, Four RC Ptsnps - Davis-Besse 1, Cycle 4 (Tech.

Spec. Figure 3.1-2b)

(270,102)

( 2146, 1 0 2)

O(300, IO2) iOO - Power Level Cutoff = 100%

(262,92)

SHUTDOWN 2 MARGIN W LIMIT E 80 - (26i,80) a 4

i 5 UNACCEPTABLE ly l

H OPERATION

8 60 -

' D a

(185,50) (225,50) m o f 180 b

i

n. ACCEPTABLE OPERATION i {e 4 $ 20 -

(112,i5)

,2.10 (iOO,0)

! 0( , ,

0 iOO 200 i B' 300 g Rod Index (7. Wi thdrawn) j Eo GR 5 i ,

gg 0 i

p 75 100 j$

$ GR 6 i , ,

$g i ax 0 25 75 iOO GR 7 p g ,

O 25 joo 4

Figure 8-5. Regulating Group Position Limits After 150 i10 EFPD, Four RC Pumps - Davis-Besse lo Cycle 4 (Tech. Spec. Figure 3.1-2c)

(270, iO2)

. (300,102) 100 - Power Level Cutoff = 100% (262,92)

SHUTDOWN g MARGIN W LIMIT I(255,80) 2 80 -

d E UNACCEPTABLE OPERATION RESTRICTED OPERATION h

g 60 -

U (215,50) (225,50) p ,"

~

o j 40 -

2 5 ACCEPTABLE OPERATION

, 20 -

" (l'48,i5) 100,0) , i 0( ,2.3) 30 0 m 16D 200 0

g Rod 'index (7. Withdrawn) 8 i

{Sr GR 5 g 0

i 75 100

$h

= g ,g GR 6 g g j=8 0 25 75 100 l l GR 7 l 0 25 100

Figure 8-6. Regulating Group Position Limits, 0 to 24+10/-0 EFPD, Three RC Pumps - Davis-Besse 1, Cycle 4 (Tech. Spec. Figure 3.1-3a) 100 -

iE

!W

' (246,77) (275,77) 80 -

O '

lI SHUTDOWN MARGIN (268,69.5) l] LIMIT 8

, y s i(265,60.5) 60 - UNACCEPTABLE

!oI OPERATION 4 a:

o j 40 (185,38 (225,38) p s e E'

s_

I 20 -

(l12,Il 75' ACCEPTABLE OPERATION

,.3) (100,0) , ,

[ 0( 300 0 100 200 Rod Index (% Wi thdrawn) m -

$. GR 5 i i i f

, ;g 0 75 100

] EE- GR 6 ;  ; g g

!s.

0 25 75 l'00

• GR 71 i l 0 25 100

Figure 8-7. Regulating Group Position Limits, 24+10/-0 to 150 10 EFPD, Three RC Pumps - Davis-Besse 1, Cycle 4 (Tech.

Spec. Figure 3.1-3b) 100 -

$ (246,77) (270,77) g 80 - 0 (300,77)

o. SHUTDOWN a MARGIN (262,69.5)

LIMIT l

( 61,60.5)

, 60 -

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Figure 3.1-Sa)

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Spec. Figure 3.1-Se) 120 -

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9. STARTUP PROGRAM - PHYSICS TESTING

• The planned startup test program associated with core performance is outlined bel ow. These tests verify that core performance is within the assumptions of the safety analysis and provide confirmation for continued safe operation of the unit. ~

9.1. Precritical Tests f 9.1.1. Control Rod Trip 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 1.66 seconds at the conditions stated above.

It should be noted that safety analysis calculations are based on a rod drop time of 1.40 seconds from fully withdrawn to two-thirds inserted. Since the most accurate position indication is obtained from the zone reference switch at the 75% inserted position, this position is used for data gathering in-stead of the two-thirds inserted position. The acceptance criterion of 1.40 seconds corrected to a 75% inserted position (by rod insertion versus time correlation) is 1.66 seconds.

9.1.2. Reactor Coolant Flow Reactor coolant (RC) flow with four reactor coolant pumps (RCPs) running will be measured at HZP steady-state conditions. Acceptance criteria require that the measured flow be within allowable limits.

9-1

n __

n 9.2. Zero Power Physics Tests 9.2.1. Critical Boron Concentration Criticality is obtained by deboration at a constant dilution rate. Once crit-icality is achieved, equilibrium boron is obtained and the critical boron con- _

centration determined. The critical boron concentration is calculated by cor-recting for any rod withdrawal required in achieving equilibrium boron. The ]

acceptance criterion placed on critical boron concentration is that the actu-al boron concentration must be within 100 ppm boron of the predicted value.

9.2.2. Temperature Reactivity Coefficient The isothermal temperature coefficient is reasured at approximately the all-rods-out configuration and at the HZP rod insertion limit. The average cool- -

ant temperature is varied by first increasing and then decreasing the tempera- '

ture by 5 F. During the change in temperature, reactivity feedback is compen-sated by a discrete change in rod motion; the change is then calculated by the summation of reactivity (obtained from a reactivity calculation on a strip chart recorder) associated with the temperature change. Acceptance cri-teria state that the measured value shall not differ from the predicted value by more than 0.4 x 10-4 (ak/k)/*F (predicted value obtained from Physics Test Manual curves).

The moderator coefficient of reactivity is calculated in conjunction with the temperature coefficisat measurement. After the temperature coefficient has been measured, a predicted value of the fuel Doppler coefficient of reactiv-ity is added to obtain the moderator coefficient. Tbis value must not be in excess of the acceptance criteria limit of +0.9 x 10-4 (ak/k)/ F.

9.2.3. Control Rod Group Reactivity Worth Control bank group reactivity worths (groups 5, 6, and 7) are measured at HZP conditions using the boron / rod swap method. The boron / rod swap method con-sists, of establishing a deboration rate in the RC system and compensating for the reactivity changes of this deboration by inserting control rod groups 7, 6, and 5 in incremental steps. The reactivity changes that occur during these measurements are calculated based on reactineter data, and differential rod worths are obtained fran the measured reactivity worth versus the change 9-2

in rod group position. The differential rod worths of each of the control-ling groups are then summed to obtain integral rod group worths. The accept-ance criteria for the control bank group worths are as follows:

1. Individual bank 5, 6, 7 worth:

predicted value - measured value x 100 < 15.

measured value --

2. Sun of groups 5, 6, and 7:

predicted value - measured value x 100 --< 10.

measured value 9.2.4. Ejected Control Rod Reactivity Worth After CRA groups 7, 6, and 5 have been positioned near the minimum rod inser-tion limit, the ejected rod is borated to 100% withdrawn and the worth ob-tained by adding the incremental changes in reactivity by boration.

Af ter the ejected rod has been borated to 100% withdrawn and equilibrium baron established, the ejected rod is then swapped with the controlling rod group and the worth determined by the change in the previously calibrated con-trolling rod group position. The boron and rod swap values are averaged and error-adjusted to determine ejected rod worth. Acceptance criteria for the ejected rod worth test are as follows:

1. predicted value - measured value x100 < 2 0.

measured value -

2. Measured value (error adjusted) i 1.0% ak/k.

The predicted ejected rod worth is given in the Physics Test Manual.

9.3. Power Escalation Tests

9. 3.1. Core Power Distribution Verification at s40, s75, and $100% FP With Nominal Control Rod Position Core power distribution tests are performed at s40, s75, and s100%FP. The test at 40% FP is essentially a check on power distribution in the core to identify any abnormalities before escalating to the 75% FP plateau. Rod in-dex is established at a naninal FP rod configuration at which the core power distribution was calculated. APSR position is established to provide a core power imbalance corresponding to the imbalance at which the core power distri-bution calculations were performed.

9-3

The following acceptance criteria are placed on the 40% FP test:

1. The worst-case maximum linear heat rate must be less than the LOCA limit.

2. The minimum DtBR must be greater than 1.30

3. The value obtained from the extrapolation of the minimun DfBR to the next power plateau overpower trip setpoint must be greater than 1.30 or the ex-trapolated value of imbalance must fall outside the reactor protector sys-

~

tem (RPS) power / imbalance / flow trip envelope.

The value obtained from the extrapolation of the worst-case maximum LHR 4.

to the next power plateau overpower trip setpoint must be less than the .

fuel melt limit or the extrapolated value of imbalance must fall outside the RPS power / imbalance / flow trip envelope.

5. The quadrant power tilt shall not exceed the limits specified in the Tech-nical Specifications.

6. The highest measured and predicted radial peaks shall be within the fol-lowing limits:

predicted value - measured value x 100 < 8.

measured value

7. The highest measured and predicted total peaks shall be within the follow-ing limits:

predicted value - measured value x 100 < 12.

measured value Items 1, 2, 5, 6, and 7 above are established to verify core nuclear and ther-mal calculational models, thereby verifying the acceptability of data from these models for input to safety evaluations.

Items 3 and 4 establish the criteria whereby escalation to the next power pla- '

teau may be accomplished without exceeding the safety limits specified by the safety analysis with regard to DtBR and LHR.

The power distribution tests perfomed at 75 and 100% FP are identical to the 40% FP test except that core equilibrium xenon is established before the 75 and 100% FP tests. Accordingly, the 75 and 100% FP measured peak acceptance criteria are as follows:

9-4

1. The highest measured and predicted radial peaks shall be within the fol-lowing limits:

predicted value - measured value x 100 < 5.

measured value

2. The highest measured and predicted total peaks shall be within the follow-ing limits:

predicted value - measured value x 100 -

< 7.5.

measured value 9.3.2. Incore Versus Excore Detector Imbalance Correlation Verification at s40% FP Imbalances are set up in the core by control rod positioning. Various imbal-ances are read simultaneously on the incore detectors and excore power range detectors. The excore versus incore detector offset slopes must be at least 1.15. If the excore versus incore detector offset slope criterion is not met, gain amplifiers on the excore detector signal processing equipment are .

adjusted to provide the required gain.

9.3.3. Temperature Reactivity Coefficient at s100% FP The average RC temperature is decreased and then increased by about SF at con-stant reactor power. The reactivity associated with each temperature change is obtained from the change in the controlling rod group position. Control-ling rod group worth is measured by the fast insert / withdraw method. The tem-perature reactivity coefficient is calculated from the measured changes in re-activity and temperature.

Acceptance criteria state that the moderator temperature coefficient shall not be positive above 95% FP.

9.3.4. Power Doppler Reactivity Coefficient at s100% FP Reactor power is decreased and then increased by about 5% FP. The reactivity change is obtained from the change in controlling rod group position. Con-trol rod group worth is measured using the fast insert / withdraw method. Re-activity corrections are made for changes in xenon and RC temperature that occur during the measurement. The power Doppler reactivity coefficient is calculated from the measured reactivity change, which is adjusted as stated above, and the measured power change.

9-5

{

The predicted value of the power Doppler reactivity coefficient is given in -

the Physics Test Manual. Acceptance criteria state that the measured value shall be more negative than -0.55 x 10-4 (ak/k)/*. FP.

9. 4. Procedure for Use When Acceptance Criteria Are Not Met If acceptance criteria for any test are not met, an evaluation is performed with participation by B&W technicai personnel as required. Further specific actions depend on the evaluation results. These actions can include repeat-ing the tests with more detailed attention to test prerequisites, added tests to search for anomalies, or design personnel performing detailed analyses of potential safety problems because of parameter deviation. Power is not esca-lated until the evaluation shows that plant safety will not be compromised by such escalation.

s 9-6 mmii -

l l

e REFERENCES i

1 BPRA Retainer Design Report, BAW-1496, Babcock & Wilcox, Lynchburg, Vi rginia, May 1978.

2 Davis-Besse Unit 1, Final Safety Analysis Report, 'acket No. 50-346.

3 J. H. Taylor (B&W) to P. S. Check (NRC), Letter, September 15, 1978.

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

! 5 Porgram to Detennine In-Reactor Perfonnance of B&W Fuels - Cladding Creep Collapse, BAW-10084PA, Rev 2, Babcock & Wilcox, Lynchburg, Virginia, Cecember 1978.

I 6 TACO-2 Fuel Pin Perfonnance Analysis, BAW-10141P, Babcock & Wilcox, Lynchburg, Virginia, January 1979.

l 7 J. H. Taylor (B&W) to J. S. Berggren (NRC), Letter, "B&W's Responses to TAC 02 Questions," April 8,1982.

8 TAFY - Fuel Pin Temperature and Gas Pressure Analysis, BAW-10044, Babcock

& Wilcox, Lynchburg, Virginia, May 1972.

9 B&W Version of P0Q07 Code, BAW-10117A, Babcock & Wilcox, Lynchburg, Vi rginia, January 1977.

10 Core Calculational Techniques and Procedures, BAW-10118A, Babcock &

Wilcox, Lynchburg, Virginia, December 1979.

11 Assembly Calculations and Fitted Nuclear Data, BAW-10116A, Babcock 8 Wilcox, Lynchburg, Vi rginia, May 1977.

12 Davis-Besse Nuclear Power Station Unit 1, Cycle 3 Reload Report, BAW-1707, Rev.1, March 1982.

A-1

13 Davis-Besse tJnit 1 Fuel Densification Report, BAW-1401, Babcock & Wilcox, Lynchburg, Virginia, April 1975.

14 Attachment 1 to Application to Amend Operating License for Removal of Burnable Poison Rod and Orifice Rod Assemblies, BAW-1489, Rev.1, Babcock

& Wilcox, Lynchburg, Virginia, May 1978.

15 Fuel Rod Bowing in Babcock & Wilcox Fuel Designs, BAW-10147P, Babcock &

Wilcox, Lynchburg, Virginia, April 1981.

16 ECCS Evaluation of B&W's 177-F A Raised-loop NSS, BAW-10105, Rev.1, Babcock & Wilcox, Lynchburg, Virginia, July 1975.

A-2

- _ _ _ _ _ _