Text

Rev 1 to Davis-Besse Nuclear Power Station Unit 1,Cycle 4 - Reload Rept
	ML20080S485
Person / Time
Site:	Davis Besse
Issue date:	10/31/1983
From:	; BABCOCK & WILCOX CO.
To:	;
Shared Package
ML20080S445	List: ML20080S444; ML20080S450; ML20080S485;
References
	BAW-1783, BAW-1783-R01, BAW-1783-R1, TAC-54261, NUDOCS 8402290132
	Download: ML20080S485 (78)
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BAW-1783, Rev. 1

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Ib DAVIS-BESSE NUCLEAR POWER STATION UNIT 1, CYCLE 4 - RELOAD REPORT 1

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{fL October 1983 DAVIS-BESSE NUCLEAR POWER STATION

[~ ' UNIT 1, CYCLE .4 - RELOAD REPORT

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BABC0CK & WILC0X Utility Power Generation Division P. O. Box 1260 Lynchburg, Virginia 24505 (i '

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1 CONTENTS

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1. INTRODUCTION AND

SUMMARY

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2. OPERATING HISTDRY ........................ 2-1

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

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4. r uE t SYS TEM 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 Collapse ................. 4-1 4.2.2. Cladding Stress .................. 4-1

{ 4.2.3. Cladding Strain .................. 4-2 4.3. Th e mal De s ig n . . . . . . . . . . . . . . . . . . . . . . . 4-2 4.4. Materi al C ompa ti b il i ty . . . . . . . . . . . . . . . . . . . 4-2 4.5. Ope rati ng Expe ri ence . . . . . . . . . . . . . . . . . . . . 4-2

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

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

7. iCCIDENT 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 k 9.1. Precritical Tests ..................... 9-1 9.1.1. Control Rod Trip Test ............... 9-1 9.1. 2. Reactor Coolant Fl ow . . . . . . . . . . . . . . . . 9-1 9-2

(- 9. 2. Zero Power Physics Tests . . . . . . . . . . . .

9.2.1. Critical Boron Concentration . . . . . .

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

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CONTENTS (Cont'd)

Page 9.3. Powe r E sc al a ti o n Te s t s . . . . . . . . . . . . . . . . . . . 9-3 9.3.1. Core Power Distribution Verification at 40, 75, and 100 FP With Nominal Control Rod Position . . . . 9-3 9.3.2. Incore Versus Excore Detector Imbalance 1 Correlation Vcrification at 40% FP ........ 9-5 j

.9.3.3. Temperature Reactivity Coefficient at 100% F P . . . 9-5 9.3.4. Power Doppler Reactivity Coefficient at 100% FP . 9-5

-9.4. Procedure for Use When Acceptance Criteria Are Not Met . . . 9-6

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R EF E R E NC E S . . . . . . . . . . ' . . . . . . . . . . . . . . . . . . A-1

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List of Tables

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'4-1. F uel De s i g n Pa rame te rs . . . . . . . . . . . . . . . . . . . . . 4-4 4-2. Fuel Thermal Analysi s Parameters . . . . . . . . . . . . . . . . 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 Themal-Hydraulic Des.ign Conditions ....................... 6-2 7-1. Comparison ~of Key Parameters for Accident Analysis . . . . . . . 7-3

2. Bounding Values for Allowable LOCA Peak Linear Heat Rates ... 7-3

=8-1. . Reactor Protection Sy' sten instrumentation Trip Setpoints . . . . 8-2 8-2. Quadrant Power Til t - Limits . . . . . . . . . . . . . . . . . . . 8-10a List of Figures Figure 3-1. Davis-Besse Cycle 4 Full Core Loading Diagram ......... 3-3 3-2. Enrichment and Burnup Distribution for Davis-Besse Ur.it 1, Cycle 4 ............................. .

3-4

3. . Control Rod Locations for Davis-Besse Unit 1, Cycle 4 ..... 3-5 5- 1.~ B0C (4 EFPD), Cycle 4 Two-Dimensional Relative Power Distribution - Full Power, Equilibrium Xenon,' APSRs Inserted . . 5-6

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8-1. Reacto r Co re Sa fety - Limi t . . . . . . . . . . . . . . . . . . . . 8-11 8-2. Trip Setpoint for Flux - Fl ux/Fl ow .............. 8-12 8-2a. Reactor Core Sa fety Limi ts . . . . . . . . . . . . . . . . . . . 8-12a 8-3. Regulating Group Position Limits, O to 24+10/-0 EFPD, Four RC Pumps Davis-Besse 1, Cycle 4 ............... 8-13 3-4. . Regulating Group Position Limits, 24+10/-0 to 150i10 EFPD, Four RC Pumps - Davi s-Besse 1, Cycl e 4 . . . . . . . . . . . . . 8-14

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Figures (Cont'd)

Figure Page 8-5. Regulating Group Position Limits,150 t 10 to 2001 10

' EFPD, Four RC Pumps - Davis-Besse 1, Cycle 4 ........ 8-15 8-Sa, Regulating Group Position Limits, 200.t 10 to 240 t 10 EFPD, Four RC Pumps - Davis-3 esse 1, Cycle 4 ......... 8-15a

( 8-Sb. Regulating Group Position Limits, 200 i 10 to 280 i 10 EFPD, 1 L Four RC Pumps, APSRs Withdrawn - Davis-Besse 1, Cycle 4 . . . . 8-15b 8-6. Regulating Group Position Limits, O to 24+10/-0 EFPD, Three ' RC Pumps - Davi s-Besse 1, Cycle 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,150t10 to 200110 EFPD, Three RC Pumps - Davis-Besse 1. Cycle 4 . .. . . . . . . . . . . 8-18

{ 8-8a. Regulating Group Position Limits, 200 t 10 to 240110 EFPD, Three RC Pumps - Davis- Besse 1 Cycle 4 ........ 8-18a i r 8-8b. Regulating Group Position Limits, 200 t 10 to 280 t 10 L. EFPD, Three RC Pumps, APSRs Withdrawn - Davis-Besse 1, Cycle 4 ........................... 8-18b 8-9.. APSR Position Limits, O to 24+10/-0 EFPD, Four RC Pumps -

(' Davis-Bdsse 1, Cycle 4 .................... 8-19 8-10 APSR Position Limits, 24+10/-0 to 150t10 EFPD, Four RC Pumps - Davis-Besse 1, Cycle 4 . . . . . . .......... 8-20 8-11. APSR Position Limits 150t10 to 200t10 EFPD, Four RC Pumps -

Davis-Besse 1 Cycle 4 .................... 8-21 8-11a. APSR Position Limits, 200 t 10 to 240 i 10 EFPD, Four RC Pumps - Davis-Besse 1 Cycle 4 . . .......... 8-21a 1 8-11b. APSR Position Limits, 200 t 10 to 280 t 10 EFPD, Three or Four RC Pumps APSRs Withdrawn - Davis-Besse 1, Cyc l e 4 . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-21b

12. APSR Position Limits, O to 24+10/-0 EFPD, Three RC Pumps -

Davis-Besse 1, Cycle 4 .................... 8-22 l 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,150t10 to 200110 EFPD, Three RC Pumps - Davis-Besse 1, Cycle 4 . . . . . . .......... 8-24 8-14a. APSR Position Limits, 200 t 10 to 240 i 10 EFPD, Three RC Pumps - Davi s-Besse 1 Cycle 4 . . . . . . . . . . . . 8-24a 7 8-15. Axial Power Imbalance Limits, O to 24+10/-0 EFPD, Four RC Pumps - Davi s-Besse 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, 150110 to 200t10 EFPD, Four RC Pumps - Davis-Besse 1, Cycle 4 . . .......... 8-27 8-17a. Axial Power Imhalance Limits, 200 t 10 to 240110 EFPD, Four RC Pumps - Davis-Besse 1, Cycle 4 ......... 8-27a

' 8-17b. Axial Power Imbalance Limits, 200 t 10 to 280110 1 EFPD, Four RC Pumps, APSRs Withdrawn - Davis-Besse 1, Cycle 4 . . . . . ... . . . . . . . . . . . . . . . . . . . . . 8-27b 8-18. Axial Power Imbalance Limits, O to 24+10/-0 EFPD, Three RC Pumps - Davi s-Besse 1, Cycl e 4 . . . . . . . . . . . . 8-28 8-19. Axial Power Imbalarce Limits, 24+10/-0 to 150110 EFPD, Three RC Pumps - Davis-Besse 1, Cycle 4 . . . . . . . . . . . . 8-29 -

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Figures (Cont'd) u Figure Page 8-20 Axial Power Imbalance Limits,150t10 to 200t10 EFPD, Three RC Pumps - Davi s-Besse 1, Cycl e 4 . . . . . . . . . . . . . . . 8-30 8-20a. Axial Power Imbalance Limits, 200 t 10 to 240 t 10 EFPD, Three RC Pumps - Davis-Besse 1, Cycle 4 . . . . . . . . . 8-30a 1 8-20b. Axial Power Imbalance Limits, 200 t 10 to 280 t 10 EFPD, Three RC Pumps, APSRs Withdrawn - Davis-Besse 1, Cy c l e 4 . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-30b 8-21. Control Rod Core locations and Group Assignments -

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

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1. INTRODUCTION AND

SUMMARY

8 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.

Retainersi 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

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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 September 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 the' base cycle'4 are 268 and 240 effective full power days (EFPD), respectively. An alternate cycle 4 design of 280 EFPD, based on axial power shaping rod (APSR) I

{ withdrawal at 200 EFPD and power coastdown, is also documented in this repo rt.

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 shuffled. The cycle 4 shuffle pattern is discussed in section 3.

The APSRs wer'e pulled at 200 EFPD to increase the lifetime of cycle 3. The APSR pull coupled with a power coastdown resulted in a cycle 3 length of approximately 268 EFPD. The alternate cycle 4 design also includes an APSR 1

(. pull and power coastdown.

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[ 3. GENERAL DESCRIPTION

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The Davis-Besse Unit 1 reactor core is described in detail in chapter 4 of

( the 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 r cd guide tubes, and one 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 ID and 2B have a 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-ence 3 provides guidelines for a fuel shuffle method that reduces the number of assemblies that are cross-core shuffled. The cycle 4 design expanded upcn this method so that only eight fuel assemblies are cross-core shuf fled. This will minimize any carryover effects from tilts in previous cycles.

Figure 3-1 is the core loading diagram for Davis-Besse Uni.t 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 assenblies will be shuffled to their cycle 4 locations. Batches 4 and SA have an initial uranium-235 enrichment

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of 3.04-wt %. Batch 5B has an initial enrichment of 2.99 wt %. Seventeen batch 10 assemblies with an initial enrichment of 1.98 wt % and 20 batch 2B 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 core. Figure 3-2 is a quarter-core map showing each assembly's burnup at the beginning of cycle (B0C) 4 and its initial enrichment.

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Cycle 4 is operated in a feed-and-bleed modes The core reactivity control is supplieo mainly by soluble boron and supplemented by 53 full-length Ag-In-Cd

- control rod assemblies (CRAs). In addition to the full-length control rods, eight axial power shaping rods (APSRs) are provided for additional control of 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 )

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

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L Figure 3-1. Davis-Besse Cycle 4 Full Core Loading Diagram

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FUEL TRANSFER CANAL I X

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6 6 6 28 58 2B 6 6 6 g L4 RB L12 Cy2 Cy2

~6 58 4 10 SB 4 5B 10 4 5B 6 C M2 K3 08 N3 F10 N13 re K13 M14 ryl Cv1 6 58 5B 4 28 4 58 4 23 4 58 SB 6 0 P11 A7 L5 C6 05 DB D11 CID Lil G15 BS Cy2 Cy?

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6 4 09 4

E10 5B 61 4

B6 5A L1 4

L8 5A L15 4

B10 58 A9 4

E6 4

C7 6

6 6 ID 28 4 '4 ID 5B 10 4 4 28 ID 6 6

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H6 - F3 FZ C1 ES B12 G9 K9 F14 F13 H12 Cy1 Cy2 Cy1 Cy1 Cy2 Cy1 6 28 58 4 5A 10 SB 2B 5B ID 5A 4 5B 23 6 g D10 C12 E4 A10 G7 84 C11 D14 E11 A6 E12 C4 D6 Cy2 Cy1 Cy2 Cy1 Cy2 6 58 4 58 4 SB " 28 1D 28 58 4 58 4 58 6 W-H HIS F6 H4 H10 D2 E3 L6 M13 N14 H6 H12 L10 H1 Cy2 Cy1 Cy2 6 2B 58 4 SA 10 58 28 5B ID SA 4 SB 2i: 6 K N10 012 M4 R10 M5 N2 05 P12 K9 R6 M12 04 NG Cy2 Cy1 Cy2 Cyl Cy2 6 6 10 2B 4 4 10 58 1D 4 4 2B ID 6 6 L , H4 L3 L2 G7 K7 P4 M11 K7 L14 L13 H10

, Cy1 Cy2 Cyl Cyl Cy2 Cy1 6 4 4 58 4 5A 4 5A 4 58 4 4 6 M- 09 M10 R7 P6 F1 F8 FIS P10 K15 M6 07 6 5B 58 4 2B 4 58 4 28 4 58 58 6

(' N P11 K1 F5 06 N5 N8 N11 010 R9 Fil P5 Cy2

( Cy2 6 58 4 10 58 4 bB -10 4 58 6 0 E2 G3 L8 D3 L3 D13 NB G13 E14

y. Cyl Cy1 6 6 6 28 5B 28 6 6 6 s P- F4 AB F12 Cy2 Cy2 6 6 6 6 6 k

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4 5 6 7 8 9 10 11 12 13 14 15 1 2 3 Batch N$N y r$5kN t d cycle 1 Cy2 = reinserted fras cy61e 2 3-3 Babcock &'Wilcox a MCDermott Company a

l Figure 3-2. Enrichnent and Burnup Distribution for j Davis-Besse. Unit 1, Cycle 4 )

8 9 10 11 .12 13 14 15 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 g

2.63 2.99 1.98 3.04 3.04 '2.99 2.63 2.99

?4,006 6,586 13,254 6,533 19,646 10,164 ' 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 m

19,828 6,527 .16,237 8,012 17,950 21,080 0 y 2.99 3.04 '2.63 3.04 2,99 2.99 2.99 11,568 19,617 23,501 17,930 8,007 8,906 0 3.04 2.99 1.98 3.04 2.99 2.99 19,764 10,141 13,560 21,087 8,889 0 2.99 2.63 2.99 2.99 + 2.99 8,33S 23,007 0 0 <

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

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1 A l 8 4 7 4 C I 8 8 1 0 7 8 5. 8 7 E 'i 1 5 5 1 F 4 8 3 7 3 8 4

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2 2 6 L 4 - 8 ~3' 7 3 8 4 M 1 5' ,

5 1 N 7 8 5 8 7 0 1 6 6 I

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l 2 3 4 5 6 7 8 9 10 11 12 13 14 15 b GROUP NO. OF ROOS FUNCTIONS X- - GROUP NUMBER 1 8 SAFETY 2 4 SAFETY 3 5 SAFETY

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5 8 CONTROL

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7 12 CONTROL

" ' 8 8 APSRs i

TOTAL # 61 3-5 Babcock & Wilcox y a McDermott company

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4. FUEL SYSTEM DESIGN

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4.1. Fuel : Assembly Mechanical Design The types bf 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 ir, 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

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4.2. Fuel Rod Design The fuel rod ' design :nd mech:nical evaluation are discussed below.

~4.2.1. Cladding Collaps_e_

, Due to its previous incore exposure time, the fuel of bacch 2B is more limit- -

ing than batches 10, 4, 5A, SB, and 6. The batch 2B assembly power histories

were analyzed to detemine the most limiting three-cycle power history for cieep collapse. -This power history was compared to a generic analysis to en-( sure that creep ovalization will not affect the fuel perfomance during '

Davis-Besse Unit 1, cycle 4 The generic analysis was based on reference 5 and is applicable to the batch 28 design.

The creep collapse analysis (Table 41) predicts a collapse time longcr than

( 35,000 effective full power hours (EFPH), which is longer than the expected residence time of 22,800 EFPH. l1

(' 4.2.2. Cladding Stress The Davis-Besse Unit 1, cycle 4 stress parameters are enveloped by a conserva-f 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 g rate. The design values are higher than the worst-case values the Davis- I 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 dianeter (10).

4.3. Thermal Desian All fuel in the cycle 4 cord is themally similar. The cycle 4 themal analy-ses represent a change in ar,alytical method. The analyses for the incoming batch 6 fuel have been perfcmed with the TAC 026 code using the analysis I

methodology described in reference 7. This methodology uses nominal undensi-fied input parameters provided in Table 4-2. Densification effects are ac-counted for in the TAC 02 densification model. 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 from batches 10, 2B, 4, and 5A were evaluated using g

TAFY38 analyses perfomed for prior cycles. I The thermal design evsluation 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 perform?a 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 heer- evaluated for the highest Surnap fuel rod and is predicted to he less than the nominal reactor coolant system pressure of 2200 psia. The naximum burnup of any fuel rod during cycle 4 is less than 42,000 mwd /ntU.

4.4 Material Compatibility The campatibi?ity of all possible fuel cladding - cool 3nt assembly interac-tions for hatch 6 FAs is identical to that of present fuel.

o 4.5. Operating 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-cunulated for tight Babcock & Wilcox (B&W) 177-FA plants using the Mark-B FA:

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L Max FA burnup,I#) mwd /mtU Cumulative net b Current electric Reactor cycle incore Discharged output,(b) MWh Oconee 1 7 48,010 40,000 41,241,515

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Oconee 2 6 27,240 36,800 36,475,957 h Oconee 3 7 22,975 35,450 37,0?2,597 Three Mile 5 25,000 32,400 23,840,053

( Island 1 Arkansas Nuclear 6 23,160 36,540 34,949,454 Unit 1

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Raacho Seco 5 37,883 37,730 29,933,40?

b Crystal River 3' 4 28,110 29,900 22,081,044

. Davis-Besse 3 28,820 25,326 12,898,260 (a)As of February 28, 1983.

(b)As of October 31, 1982.

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Table 4-1. Fuel Desir, Parameters

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Batch 10 2B 4 5A SB 6 FA type Mk-B4A Mk-B4A MS-B4A Mk-B4A Mk-B4A Mk-B4A l Number of assemblies 17 20 44 8 40 48 Fuel rod OD, in. 0.430 0.430 0.430 0.430 0.430 0.430 )

Fuel rod 10. in. 0.377 0.377 0.377 0.377 0.777 0.377 Flexible spacer type Spring Spring Spring Spring Spring Spring ]

Rigid. spacer type Zr-4 Z r-4 Zr-4 Z r-4 Zr-4 Zr-4 Undensified active fuel length, in. 143.5 143.5 143.44 143.44 143.20 143.20 Fuel pellet-(mean) dia., in. 0.3675 0.3675 0.3697 0.3697 0.3686 3.3686 Fuel pellet initial density, % TD mean 96 96 94 '?% 95 95 Initial fuel nent, wt % 23knrich- 1.98 2.63 3.04 3.04 2.99 2.99 Estimated residence 'l time ,- EF PH 15,6 % 22,800 20,064 12,950 12,960 6,720 l1 J Cladding collapse time , EF PH >3 5,000 >3 5,000 >35,000 >35,000 >35,000 >35,000

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r Table 4-2. Fuel Thermal Analysis Parameters

, , Batch 4/5A 6B 6 1"f28 Number 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

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Nominal stack height, in. 143.5 143.44 143.2 143.2

(. Enrichnent, wt % 235U 1.90/2.63 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

{

(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 TAC 02 analyses.

(c)BOL, TAC 02 code.

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5. NUCLEAR DESIGN

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5.1. Physics Characteristics "

(. Table 5-1 compares the core physics parameters of cycle 3 with those of both '

the base and alternate designs for cycle 4. These values were generated using

(' PDQ07 9 -ll f or both cycles. Since the core has not yet reached an equilibrium ,

cycle,' differences between the cycles in core physics parameters are to be expected. Figure 5-1 illustrates a representative relative power distribution for the B0C at full power (FP) with equilibrium xenon and group

( 8 inserted.

Due to the difference in design cycle lengths, the critical boron concentra-tions 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 section 8. The adequacy of the shutdown margia 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).

Babcock & Wilcox 5-1 . = co u....,

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5.2. Changes in Nuclear Design There are no significant core design changes between the reference cycle and t

I 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

- pa rameters. No significant operational or procedural changes exist with re- ]

gard to axial or radiel power shap:t, xenon, or tilt control. The alternate 1

design includes an APSR piill at 200 EFPD and a planned power coastdown to end 1

)

of cycle' (EOC) at 280 EFPD. The staisility and control of the core with APSR's withdrawn have been analyzed. The calculated stability index without

]

A" R's is - 0.0548h-1, which demonstrates the axial stability of the core.

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Revision 1 (10/14/83)

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Table 5-1. Davis-Besse Unit 1, Cycle 4 Physics Parameters Cycle 4 Base Alternate

(' Cycle 3 design (a) design

[ Cycle length, EFPD Cycle burnup, mwd /mtU 8,929 268 240 8,014 280 9,350 Average core burnup - E0C, mwd /mtU 20,191 18,924 20,259

{ Initial core loading, mtU 83.2 83.0 83.0 Critic HZP1)bboron - BOC, No Xe, ppm 1,231 1,250 1,250 HFP b) Group 8 inserted 1,015 1,042 1,042 Critical boron - E0C, Eq. Xe, ppm N

Group 8 withdrawn H 2f0 4 (c)

( Control rod worths - HFP, 800, % ak/k ,

( Group 6 0.93 1.02 1.02 Group 7 1.52 1.73 1.73 Group 8 0.31 0.27 0.27 Control rod worths - HFP, E0C, % ak/k Group 7 1.53 1.74 1.68 1 Group 8 NA 0.35 NA Max ejected rod worth - HZP, % Ak/k (location) 00C Groups 5-8 inserted 0.78 0.85(d) 0.d5(d)

(N-12)

(_ E0C Groups 5-7 inserted, Group 8 0.72 (N-12)d) 0.85t (N-12)d)

'0.85t withdrawn (N-12) (N-12) (N-12)

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

L BOC 1,44 1.70 1.70 (N-12) (L-14) (L-14)

E0C 1.25 1.54 1.46

(~ Power deficit-HZP to HFP, Eq. Xe, % ak/k (L-14) (L-14) (N-12)

B0C f,4 EFPD) -1.79 -1.77 -1.77 E0C -2.34 -2.36 -2.33

{ Dopfler coeff - HFP,10-5 ak/k/ F B0C,- No Xe,1042 ppm, Group 8 inserted -1.46 -1.47 -1.47 E0C, Eq. Xe,10 ppm, Group 8 withdrawn -1.63 -1.58 -1.63

(. Moderator coeff - HFP,10-4 Ak/k/ F I B0C, No Xe,1042 ppm, Group 8 inserted -1.13 -1.00 -1.00 ECC, Eg Xe, 10 ppm, Group 8 withdrawn -2.89 -2.87 -2.76

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Boron worth - HFP, ppm /% Ak/k 80C (1042 ppm) 110 108 108 E0C (10 ppm) 97 97 96

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Table 5-1. (Cont'd }

Cycle 4 Base Alternate Cycle 3 desion(d) design Xenon worth - WP, % ak/k B0C (4 EFPD) 2.67 2.67 2.67 EOL (equilibrium) 2.73 2.74 2.14 Ef fective- delayed neutron fraction - HFP

' BOC 0.00595 0.00598 0.00598 1 EU, 0.00530 0.00530 0.00530 -

(a) Group 8 is irserted at EOC in the Base Cycle 4 Design.

(b)HZP denotes hot zero power (532F Tavg); HFP denotes hot full power ]

(584F Ta vg)-

(c) Power coastdown to E0C at 10 ppmb.

- (d) Ejected rod worth at the rod insertion limit. }

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- Table 5-2. Shutdown Margin Calculation for Davis-Besse, Cycle 4 L

E0C,% AK/K F Base L' design Alternate design J B0C 240 EFP0 204 EFP0 280 EFPD

% AK/K Bank 8 in Bank 8 out Bank 8 out e

Available Rod Wor:.h Total rod worth, HZP 7.69 7.89 7.72 7.85 Worth reduction due to burnup of -0.16 -0.19 -0.19 -0.20 poison material

{ Maximun stuck rod, HZr -1.54 -1.44 -1.46

-1.73 Net worth- 5.83 6.16 6.09 6.19 Less 10% uncertainty -0.58 -0.62 -0.61 -0.62 3 Total available worth 5.25 5.54 5.48 5.57

[ Required Rod Worth Power deficit, WP to HZP 1. ? 7 2.36 2.24 2.33

{~ Max allowable inserted rod worth 0.51 0.6a 0.74 0.74 Flux redistribution 0.73 1,15 1.12 1.19 Total required worth 3.01 4.15 4.10 4.25 Shutdown Margin Total available minus total 2.24 1.39 1.38 1.32 required Note: Required shutdown :crgin is 1.00% Ak/k

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Figure 5-1. BOC (4 EFPD), Cycle 4 Two-Dimensional Relative l'ower Distribution > Full Power, Ep.ilibrium Xenon, APSRs Inserted (a) 8 9 10 11 12 13 14 15 H 0.816 0.885 1.293 1.142 1.179 1.043 1.155 0.986 K 0.889 1.206 0.982 1.286 1.000 1.108 0.909 0.937 ]

L 1.?98 0.988 1.034 1.045 0.725 0.829 1.240 0.774 3

)

M 1.143 1.285 1.040 1.155 0.967 0.904 1.022

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ti 1.180 1.001 0.7N240.967 1.184 1.029 0.714

]

r 0 -1.004 1.109 0.829 0.911 1.035 0.811 P 1.15G 0.910 1.241 1.024 0.715 0.986 0.937 0.775 Inserted rod g. omber x.xx Relative power (a) Calculated results from two-dimensional pin-by-pin PDQ079 .

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6. THERMAL-HYDRAULIC DESIGN The fresh batch 6 fuel is hydraulically and geometrically similar to the

f. 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 identical 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 tcpical 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 productior capability of the FAs with irradiation.

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

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

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Design power -level, MWt 2772

' System pressure, psia 2200 ]

Re 3ctor. coolant flow, % design 110 Vescel 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 Enthalpy rise q 1.011 Heat flux (Fy)(F ). 1.014 )

i Flow area 0.98 Active fuel length See Table 4-2

}

Avg heat flux 100% power, Btu /h-ft2 1.89 x 10 (a)5 Max heat flux,100% power Btu /h-ft 2 4.85x10(a) 5

)

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

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

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7. ACCIDENT AND TRANSIENT ANALYSIS 7.1. General Safety Analys3 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-h sure.that thermal perfonnance 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 detemining 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 faedback 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-

f. ture as a function of linear heat rate (LHR) and the lifetime pin pressure g . 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

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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, thennal-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 i 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.

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Table 7-1. Comparison of Key Parameters F for Accident Analysis FSAR and densif'n Base Alternate report cycle 4 cycle 4 Parameter value value value BOL(a) Doppler coeff,10-5, ak/k/*F -1,28 -1.'4 7 -1.47

( E0L(b) Doppler coeff,10-6, ak/k/'F -1.45(c) -1.58 -1.63 BOL moderator coeff,10-4, ak/k/*F +0.13 -1.00 -1.00 1 E0L moderator coeff,10-4, Ak/k/'F -3.0 -2.87 -2.76

(

All rod bank worth (HZP), % ak/k 10.0 7.69 7.69 Boron reactivity worth (HFP), ppm /1% ak/k 100 108 108 Max ejected rod worth (HFP), % ak/k 0.65 0.46 0.46 Max dropped rod worth (HFP), % ok/k 0.65 0.20 0.20 Initial boron cone (HFP), ppm 1407 1042 1042

( (a)BOL denotes beginning of life.

(b)EOL denotes end of li fe.

(c)-1.77 x 10-5 ok/h/'F was used for steam line failure analysis.

[ Tabl e 7-2. Bounding Values for Allowable LOCA Peak Linear H, eat Rates Al1owable Al1owable Core peak LHR, peak LHR,

(. el evation, first 24 EFPD, balance of cycle, ft kW/ft kW/ft 2 15.5 16.5

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4 16.8 17.2

[. 6 18.0 18.4

, 9 17.5 17.5

< 10 17.0 17.0

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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 thennal design criteria will not be vio-lated.

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

1. Manual reactor trip Not applicable. Not applicable.

High flux <104.94% of RATED THERMAL POWER with <104.94% of RATED THERMAL F0WER with l

2. Tour pumps operating #

Tour puinps operating l

<79.85% of RATED THERMAL POWER with l

! <79.85% of RATED THERMAL POWER with l l three pumps operating three pumps operating # l RC high temperature <618*F <618*F#

3.

4. Flux -- afl ux/ fl ow(1) Trip setpoint not to exceed the lim- Allowable values not to exceed the it line of Figure 2.2-1 limit line of Figure 2,2-1#

5. RC 1ow pressure (l) 11983.4 psig 11983.4 psig* 11983.4 psig**

m m lo S 12300.0 psig* <2300.0 psig**

6. RC high pressure 12300_psig

7. RC pressure-temperature (l) 1(12.60 Tout F - 5662.2) psig 1(12.60 Tout F - 5662.2) psig# l

<55.1% of RATED THERMAL POWER with <55.1% of RATED THERMAL POWER with

8. High fluqtqumber of RC one pump operating in each loop #

pumps on i one pump operating in each loop

<0.0% of RATED THEPMAL POWER with <0.0% of RATED THERMAL POWER with two pumps operating in one loop and two pumps operating in one loop and no pumps operating in the other loop no pumps operating in the other loop f

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

erating

,g no <4 psigI S Er 9. Containment pressure high H psig 3"

lY W.

30

-< o o s ~ m m - m m m m_ m

~

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 or the DNBR limit produced combination of the by a nuclear radial pawerpeak, peak, axial peaking 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 SB and 6 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 L DER 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 DER is equal to +22%, whichever condition is more restrictive.

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

The DNBR as calculated by the B&W-2 DE correlation continually increases

( from point of minimum DER, so that the exit DER is always higher. Extrapo-lation of the correlation beyond its published quality range of +22% is justi-fied.on the basis of experimental data.

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I 2.2. LIMITINC SAFETY SYSTEM SETTINGS BASES __

2.2.1. REACTOR FROTECTION SYSTEM INSTR' J MENTATION SETP0!hiS 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 par 5m-eter. The trip setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceedino the9 safety limits.

The shutdown bypass provides for bypassing certain functions of the reactor lg protection system in order to pemit 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 nomal operation with shut-down bypass activated. This high pressure trip tetpoint is lower than the nomal 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 PATED 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 sy; tem instrumentation channels and provides manual reactor trip capabil-i ty.

High Flux 7 A high flux trip at high power level (neutron flux) provides reacter core pro-tection against reactivity excursions which are too rapid to be protected by temperature and pressure protective circuitry.

During nomal station operation, reactor trip is initiated when the reactor power level reaches 104.94% of rated power. Due to Leansient 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.

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s LIMITING-SAFETY SYSTEM SETTINGS-L BASES

( RC High Temperature The RC high temperature trip <616'F prevents the reactor out1ct temperature from exceeding the design limTts and acts as a backup trip for all power ex-cursion transients.

( Flux -- AFlux/ Flow The power level trip setpoint produced by the reactor coolant system flow is (n based on a flux-to-flow ratio which has been estaolished to accommodate flow decreasing transients from high power where protection is r60t provided by the L high flux / number of reactor coolant pumps on trips.

The power level trip setpoint produced.by the power-cc-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 lacreases. The power level setpoint produced by the power-to-flow ratio provides overpower 9NB 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 tiip are as follows:

1. Trip would occur wh'en 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. . irip would occur when three reactor coolant pumps are' operating if power is 79.9% and reactor coolant flow rate is 74.7%.of full fiaw rate, or flow rate is 70.2% of full flow rate and power is 75%.

{ _

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

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

RC Pressure - Low, High, and Pressure Temperature The high and low trips are provided to limit the pressure rance in which reac-tor operation is pemitted.

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 3 pressure, 2300 psig, has been estsblished to maintain the systen pressure N- l 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 prc5sure for these valves, 2435 psig. The RC high pressure trip also backs up the high flux trip.

t The RC low pressure ,1983.4 psig, and RC pressure-temperature (12.60 ot ut -

5662.2) psig, trip setpoints have been established to naintain the DNB ratio greater than or equal to 1.30 for those design accidents that result in a pressure reduction. It also prevents reactor operatico 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 cn trip prevents the minimum core DfER 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.

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~ PEACTIVITY CONTROL SYSTEMS REGULATING R00 INSERTION LIMITS LIMITING CONDITION FOR OPERATION

[

3.1.3.6 The regulating rod groups shall be limited in physical insertion as shown on Figures 3.1-2a, -2b, and -2c and 3.1-3a, -3b, and -3c for the first 200 10 EFPD of operation. If the axial power shaping rods are completely withdrawn at 200 110 EFPD for extension of cycle length, then the regulating rod groups shall be limited in physical insertion as shown on Figures 3.1-2e 1 and 3.1-3e for the remainder of the cycle. However, if the axial power shap-ing rods are not completely withdrawn at 200 10 EFPD, then the regulating rod groups shall be limited in physical insertion as shown on Figures 3.1-2d and 3.1-3d for the remainder of the cycle. A rod group overlap of 25 5%

shall be maintained between sequential withdrawn groups 5, 6 and 7.

ADPLICABILITY: MODES 1= n;d 2*#,

ACTION With the regulating rod groups inserted beyond the above insertion limits (in a region other than accepta'le operation), or with any group sequence or over-lap outside the specified 1.mits, 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 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , 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 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or

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

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

See Special Test Exceptions 3.10.1 and 3.10.2.

vWith k ggf 2 1.0.

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

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

3.1.3.9 The axial power shaping rod group shall be limited in physical in-sertion as shown oa Figures 3.1-Sa, -5b, -Sc, -5f, -5g and -Sh for the first ,

200 10 EFPD of operation. If this rod group is completely withdrawn at 200 J 10 EFPD for extension of cycle length, it shall not be reinserted in the core for remainder of the cycle and the limits of Figure 3.1-Se shall be app'icable. I However, if the rod group is not completely withdrawn at 200 110 EFPD, the

. group shall be limited in physical insertion as shown on Figures 3.1-5d and ]

-Si for the remainder of the cycle.

]

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 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , 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 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or )

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

]

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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 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 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

]

With k eff a 1.0.

]

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

8-8 Babcock & Wilcox e sco. ..n .....,

s Revision 1 (10/14/83)

F L

3/4.2. POWER DISTRIBUTION LIMITS AXIAL POWER IMBALANCE

[

LIMITING CONDITION FOR OPERATION

[

3.2.1 AXIAL POWER IMBALANCE shall be maintained mthin the limits shown on Figures 3.2-la, -)b and -Ic and 3.2-2a, -2b and -2c for the first 200 c10

( EFPD of operation. If the axial po.wer shaping rods are completely withdrawn at 200 10 EFPD for extension of cycle length, then the AXIAL POWER IMBALANCE shall be maintained within the limits shown on Figures 3.2-le and 3.2-2e for 1 the remainder of the cycle. However, if the axial power shapir.g rods are not

[ completely withdrawa at 200 10 EFPD, then the AXIAL POWER IMBALANCE shall be maintained within the limits shown on Figures 3.2-1d and 3.2-2d for the re-mainder of the cycle.

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

! 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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when above 40% of RATED THERMAL POWER except wnen 5 the AXIAL POWER IMBALANCE alarm is inoperable, then calculate the AXIAL POWER IMBALANCE at least once per hour.

[ l.

(

l

See Special Test Exception 3.10.1.

DAVIS-BESSE, UNIT 1 3/4 2-1 Babcock s,Wilcox 8-9 a "co"*" ""**" l m . . . . . . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

'i

'k POWER DISTRIBUTION LIMITS BASES :

= l FN Nuclear Enthaloy Rise Hot Channel Factor, is defined as the ratio AH of the integra'l of linear pwer along the rod on which minimun Df0R occurs to the average Nd power.

It has been determined by extensive analysis of possible operating power shapes that the design limits on nuclear power peaking aN. on miniiNm DiBR at full power are met, provided:

F g ~< 2. 93 ; FN < 1.71 l SH ~ t 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- . .y; s3.j;.p its will be met provided the following conditions are maintained.

1 Control rods in a single group move togethe:c with no individual rod in- J.4f w sertion differing by more than 16.5% (indicated position) from the group average height.

hi y;.,; .:

2. Regulating rod groups are sequenced with overlapping groups as required in Specification 3.1.3.6. %,y ~~ d

.o

3. The regulating rod insertion limits of Specification 3.1.3.6 are main- +.:NA tained. 1 A*

l.Q,D 4 4 AXIAL F0WER IMBALANCE limits are maintained. The AXIAL POWER IMBALANCE Of ;h '

is a measure of the difference in power between the top and bottom halves [M of the core, Calculations of core everage axial peaking f actors for many le g g 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 M(A d N.>

g4 .c' POWER IMBALANCE is maintained between the limits specified in Specifica-Eg tion 3.2.1.

The design limit power peaking factors are the most restrictive calculated at lI Q:

j ch Y%

full power for the range from all control rods fully withdrawn to minimum al- .

M bj 4 luable control rod insertion and are the core DNBR design basis. Therefo re, for operation at a fraction of RATED THERMAL P?.lER, the design limits are ls, : ..g met. When using incore detectors to make power distribution maps to deter-minc Fq and F N :g$f ' .

j,f"$: :

AH g..f

a. The neasurement of total peaking factor Fg , shall be increased by 1.4 yyt e, ,

percent to account for manufacturing tolerances and further increased by 4. s. $ J 7.5 percent to account for measurement error. ig y I%d B 3/4 2-2 QW#:

' l 4;.:7%

6 Revision 1 (10/14/83)' .

Table 8-2. Quadrant Power Tilt Limits m' (Tech. Spec. Table 3.2.2) e -

Steady state Transient Maximum ', ,.

limit limit limit .s

{ .-

L Measurement independent 4.92 11.07 20.0 4 QUADRANT POWER TILT E'

r ' '- %

QUA0 RANT POWER TILT as L

measured by:

Symmetrical incore 3.43 8.93 20.0

{ detector system y

i '

Power range channels -

1.96 6.96 20.0 Minimun incore detector system 1.90 4.40 20.0

[

o

[

[ .

[

[ :

N

[ .

[ c-

[

3/4 2-12

[ 8-10a Babcock s.Wilcox

.uco ..n ...m -

G

- Figure 8-1. Reactor Core Safety Limit .

(Tech. Spec. Figure 2.1-2) 5 RATED THERMAL POWER 4 PUMP . 120 LIMIT (33,i121

(-31,112) <

6 -

{

( -46,100) .

100 (%,100) 3

(-31,89.24) (33,89.24) ,

3 PUI _

LIMIT - 80

(-46,77.24) ; )(40,77.24)

ACCENABLE f-

- 60 UNACCEPTABLE l OPERATION _

UNACCEPTABLE OPERATION l FOR SPECIFIED OPERATION RC PUMP COMBINATION l -

40 g.

^

b

-20 r.

l P t i 1 - f I I 40 -40 -20 0 20 40 60 Axial Power imbalance, %

PUNPS OPERATING REACTOR COOLANT FLOW, GPM 4 387,200 I 3 290,100 .

.s

~

l ...

8-11 Babcock & Wilcox

. a.o ..nc..,-,

J.

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

Curve shows trip setpoint for a 25%

flow reductior, for three pump operation )

(290,l00 gpm). The actual setpoint wi.nl t,e directly proportional to the actual -

flow with- three pumps. -

% RATED THERNAL POWER UNACCEPTABLE

- - 120 OPERATION UNACCEPTABLE 1 J

OPERAT10N (-!4,1069) (17,1069)

M2 =-l .8576 Mg =0 9288 i pgMp - 100 ]

LIMIT I I (26,90.182)

(-32,90 182)

I i i(-l4,79 85) ___

(17,7985)

I i

! 3 PUMP 1 I IMIT (26,63.132)

(-32,63.132) - -

I i 1ACCEPTABLEOPERATIONFOR !

l ISPECIFIEDRCPUNPCOMB-

- - go l

]

llMATION I I I I

i i I l ]

- 20 l N

T T*l ]

i l lt: IM

! l l "m ! ,

I- , ,

, , I , i i ]

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

]

8-12 Babcock & Wilcox

.=co.-.n.... ,

F-L F Figure 8-24. Recctor Core Saf ety Limit (Tech Spec Figure 2.1-1) 2600 -

[

2400 -

RC HIGH FRESSURE TRIP 618,1300

(

m S

2200 - l' c / RC HIGH TEMPERATURE TRIP 8,2124.6 5' ACCEFTABLF

(

~

OPERATION RC PRESSURE TEMPERATURE TRIP 2000 - 606.79, 1983.4 / ! - SAFETY LIMIT RC LOW PRESSURE TRIP 1800 f i i 1 1 J

{ 590 600 610 620 630 640 Reactor Outlet Temperature. *F

[

(

l s

[..

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8-12a Babcock s.Wilcox

. uco.,..n c.....,

. _r- v r" _

w w et r- .

w, cw Figure 8-3. Regulating Group Fos) tion Limits, O to 24+10/-0 EFPD, Four RC Pumps - Davis-Besse 1. Cycle 4 (Tech. Spec. Figure 3.1-2a)

(246,102 (275,IO2) iOO

- Power Level (300,102)

Cutoff = 100% l SHUTDOWN MARGIN (268,92)

{ LIMIT y 80 -

(265,80)

Y $g e

5 UNACCEPTABLE 3

60 -

OPERA TIGN [&

(l85,50) @ (225,50)

? I

" 40 _

! 5 ACCEPTABLE

$ OPERATION l S '

$ 20 -

2 (i;2,I5) -

o< .2.4) (100.0) i 3 0 l00 cc 200 000

$ RodIndex(% Withdrawn)

EO GR 5 i i yE- 0 75 100 i s.

! GR 0: g i ,,

0 25 75

8x

100 GR 71 l l 0 25 100

Figure 8-4. Regulating Group Position 1.imits, 24+10/-0 to 150t10 EFPD, Four RC Pumps - Davis-Besse 1, Cycle 4 (Tech.

Spec. Figure 3.1-2b)

(270,102) o(300,iO2)

Power Level (246,102) 100 -

Cutoff = 100%

(262,92)

- SHUTDOWN E, MARGIN k 80 p(61',80) a 8

UNACCEF TABl.E 5 OPERATION 4@'

S 60 -

[

t2 (185,50) 4 (225,50)

[o 4 co

$ u0 -

E E ACCEPTA3LE OPERATION b

d 20 -

(112,15) o< 2.4) (i".0) i _

i m 0 100 200 300 E. Rod index (% Wi thdrawn) h$o GR 5 : I I 0 75 100

'((

!! GR 6 i i i .I

$8 3M 0 25 75 100 GR 7 l l l 0 25 100

--; t----, .u n - - - ==== . . . . _ mmet

v , _ . r- .- - v_ r, .. w r r-m r, w w- y Figure 8-5. Regulating Group Position ifmits,150 1 10 y to 200 1 10 EFPD, Four RC pumps - Davis- "

Besse 1, Cycle 4 (Tech. Spec. Figure 3.1-2c)

(270, IO2)

' (300,102)

Ic3 - Power Level Cutoff = 100% ,

(262,92)

SHUTDOWN 2 MARGIN si!

LIMIT >(255,80) 2 80 -

d E UNACCEPTABLE OPERATION RESTRICTED OPERATION o 60 -

t2 (215,50) (225,50) m [O e-*

'" e g 40 - '

E e'

S ACCEPTABLE OPERATION O

20 -

" (i48,i5) w a

0(

6

$ . /)

2.3 Id6 100,0) i 200 j

300 m

7

(

Rod 'Index (% Withdrawn) g -

i i

?w GR 5 i ;-

jg 0 75 100 R

g 0

E= GR 6 g g l .l 38 25 75 100 3 ,

- O y .

GR 7 I I I 5

0 25 100 2

h

_____________m_

1 Figure 8-Ca. Regulat.in g Group Position Limits, 200 1 10 9 1

to 240 1 10 EFPD, Four RC Pumps - Davis-Besse 1, Cycle 4 (Tech. Spec. Figure 3.1-2d) ;

(270,IO25 .

(266,l'02) '(300,iO2) l 100 -

Power Level

_ Cutoff = 100% (262,92) 2 SNUTDOWN N MARGIN 2 80 - LIMIT I(255,00) i d

b UNACCEPTABLE g 0PERAT'0N OPERATION RESTRICTED g 60 -

ti (2i5,50) (225,50)

?

l 0;

I U 160 -

8 b

e ,

s I 20 o

ACCEPTABLE OPERATION (i48,i5) 0t ,

IM,0) i i g 0 1(6 200 300 $

,g' Rod'Indax(% Withdrawn) -

gh GR 5 , i O

!, d 75 100 R 5s %

ig GR 6 i y y j b iR 0 25 75 100 GR 7 I I 7 0 25 100

; e_.; % > , % o s m

- r- ~ r v em. n r- u --- w Figure 8-5b. Regulating Group Positicn Limits, 200 1 10 to 280 i 10 EFPD, Four RC Pumps, APSRs Withdrawn -

Davis Besse 1, Cycle 4 (Tech. Spec. Figure 3.1-2e)

(262,102)

O(300,102) 100 POWER EVEL CUT 0FF = 100% ,,(256,92) 2 y SHUTDOWN OPERATION RESTRICTED l g 80 -

MARGIN (30,60) I l _, LIMIT l 1 g '

05

c UN ACCEPTABL E

$ 60 -

g OPERATION N (209,50)

? ;

1 G

" t W _

8 b

S.

ACCEPTABLE OPERAT10N l g 20 -

o.

(138,15) l (0,2.0) :o 1 0 (0r '

I I '

~

100 200 300 g Rod Index(% Withdrawn) [

pg GR 5 l l l c iw a e-0 75 i00 R

!$ l 4

GR 6l l

$nE ax O 25 75 100 $

GR 7 l l l 0 25 100

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

2 (246,77) (275,77)

o. 80 -

0 (300,77) a SHUTDOWN

@ MARGIN (268,69.5)

$ LIMIT 60 - UNACCEPTABLE g*' [ (265,60.5) k' OPERATION 4 D

p#

(225,38) 7 { 40 -

(185,38 5 $

e s.

y 20 -

j (li2,II.75'- ACCEPTABLE OPERATION

'N I i 0( . 3) 200 300 0 100 Rodindex(% Withdrawn)

E GR S i i i

.E O 75 100 GR 6 l l l l 75 100

$$ 0 25

%n ja GR 71 I I

'* 25 100 O

_m__s xn - _ _ m

v v .- m e- r-m ~ -x r ---- x _ r w Figure 8-7. Regulating Group Position Limits, 24+10/-0 to 150 10 EFPD, Three RC Ptsnps - Davis-Besse 1, Cycle 4 (Tech.

[

Spec. Figure 3.1-3b) 100 -

i e

l 5 (246,77) (270,77) 0 (300,77) l g 80 -

SHL'TDOWN g

, o_

j MARGIN (262,'69.5)

! g LlHIT s@d (61,60.5) 60 -

1 l

e @p h UNACCEPTABLE w OPERATION

)

co 40 -

(185,38 (225,38)

A b 20 -

3 o-(l12,II.75) ACCEPTABLE OPERATION O

,2.3) ,

( #'N I I 100 200 300 O

RodIndex(7. Withdrawn)

W GR 5 i l l E- 0 75 100

8 I I I

$E GR 61 0 25 75 800

?$ I EE GR 7 1 1 100 30 0 25

a. ._

J Figure 8-8. Regulating Group Position Limits, 150 1 10 to y 200 10 EFPD, Three RC Pumps - Davis-Besse 1, Cycle 4 (Tech. Spec. Figure 3.1-3c) 100 -

2 E

l I g 8o -

(266,77)(270,77)O(300,77) l a SHUTDOWN y MARGIN (262,69.5) y LlHIT H

60 -

(255,60.5)

S UNACCEPTACLE Q

OPERATION OPERATION RESTRICTED (215,38) (225,38)

{ 11 0 E

5 h 20 -

g 7(;gg,ig,7g) ACCEPTABLE OPERATION

.2 100.0) i i x 0< - o 0 100 200 300 5.

m Rod index (% Wi thd rawn) {-

h?- GR 5 l .i ~

8 0 7b 100 t9- GR 6 l 1 I l P_

3;

o 25 75 a00 %

2n h 32 GR 7 l l l 5 0 25 100 mm

v e - r .

v m u- x .r-y Figure 8-8a. Regulating Group Position Limits, 200 1 10 to 240 i 10 EFPD, Three RC Pumps - Davis Besse 1 Cycle 4 (Tech. Spec. Figure 3.1-3d) 100 -

2 m

3 n.

80 -

(266,77)(270,77)') (300,77) I a SHUTDOWN I MARGIN (262,69.5)

@ LIMIT

" (255,60.5) 60 -

@ UNACCEPTABLE

, g 0PERATION l

OPERATION RESTRICTED l O

$ U # -

(215,38) (225,38)

S $

E h 20 -

E (148,11.75) ACCEPTABLE OPERATION o< w .z. ._

100.0) i i j,'

O I00 200 300 2.

R Rod index (f. Wi thdrawn) S

.K GR 5 .i ~

28 0 75 100 O y x- o 3e N

=g GR 6 I I .I %

jg: 0 25 '75 100 g iN .

Gil 7 [ l l

0 25 100

Figure 8-8b. Regulating Group Position Limits, 200 1 10 to 280 10 EFPD, Three RC Pumps, APSRs Withdrawn -

Davis-Besse 1, Cycle 4 (Tech. Spec. Figure 3.1-3e) l00 -

1 p (262,77) (300,77) w 80 - MARGIN 3 LlHIT

[ (256,69.5)

_a 9 UNACLEPTABLE OPERATION RESTPICTED 60 OPERATION h

(2160,60.5 )

e e

a f*

(209,38) u b

a ACCEPTABLE OPE'1ATION Y 20 -

I f (138,11.75)

C I I I O

O 100 200 300 E Rod Index(f. Withdrawn) g-g GR S i 0

l 75 3

100 7

e -

E l O GR 6 l l l

$E 25 75 100 o

AP O D

:6 GR 7 I I I $

- O 25 100 5

-~

I 3 i N . .n_ m. h meg

s.

L y Figure 8-9. APSR Position Limits, 0 to 24+10/-0 EFPD, Four j RC Pumps - Davis-Besse 1, Cycle 4 (Tech. Spec.

Figure 3.1-Sa) -

(8,102) (38,IO2) 100 - C O

( (8,92) (38,92) RESTRICTED REGION 5

k 80( (0,80) (42,80) a f

c:

, N w

o 60 -

r # PERMISSIBLE ls I OPERATING REGION (l00,50) g 40 -

E 5.

s j 20 -

c.

0 i i

( i i i i i i i 1 0 10 20 30 40 50 60 70 80 90 iOO

( APSR Position (f. Wittdrawn)

I t

( 8-19 Babcock & Wilcox

. uco n c.... ,

Figure 8-10. APSR Position Limits, 24+10/-0 to 150 10 EFPD, Four RC Pumps - Davis-Besse 1, Cycle 4 (Tech. Spec. Figure 3.1-5b)

(8,102) (38,102) 100 O O f

(8,92) (38,92) e ]

$ I 2 80( (0,80) (42,80) RESTRICTED ;

g REGION x

e I

S 60 -

b !

Y (l'00.50)

{ u yo _ PERMISSIBLE OPERATING REGION

]

f 20 - -

1 0 t i i i i i i i i i 0 10 20 30 40 50 60 70 80 90 100 APSRPosition(7. Withdrawn) 1 J

i:

8-20 Babcock & Wilcox

. u.o....n c.....,

s Revision 1 (10/14/83) ,

F L

e Figure 8-11. APSR Position Limits, 150 i 10 to 200 10 1

L EFPD, Four RC Pumps - Davis-Besse 1, Cycle 4 (Tech. Spec. Figure 3.1-5c) f-L (8,102) (38,IO2) 100 O e

(8,92) (38,92) RESTRICTED REGION v- 2 o 80 ( (0,80) 48,80) c.

a :

E (100.70)!1

$ 60 -

e h

FER$41SSIBLE OPERATING REGION "o

( g 40 -

8

< b

( S h 20 -

E O i s i i i : i 3 0 10 20 30 40 50 60 70 80 90 100

(. APSRPosition(7eWithdrawn)

(

8-21 Babcock & Wilcox

. u o....n .... ..

J'

l Revision 1 (10/14/83)

Figure 8-11a. APSR Position Limits, 200 10 to 240 10 EFPD. Four RC Pumps - Davis-Besse 1, Cycle 4 1 (Tech. Spec. Figure 3.1-5d)

(8,iO2) (38,IO2) 0 100 ,.

(8,92) (38,92) 5 3 80 ( (0,'80) (48,80)

a. -

a RESTRICTED REGION I

5 60 -

h PERMISSIBLE OPERATING REGION o (iOO,50)

40 -

8 b

b s.

g 20 --

2 0 't i i i i i i i i i 0 10 20 2 40 50 60 70 80 90 100 c

APSR Position (% Withdrawn)

)

J l

8-21a Babcock & Wilcox

. w o.,..n c.....,

Revision 1 (10/14/83) k

- Figure 8-11b. APSR Position Limits, 200 10 to 280 10 EFPD, !

l Three or Four RC Pumps, APSRs Withdrawn -

Davis Besse 1, Cycle 4 (Tech. Spec. Figure 3.1-Se)

F' 110 -

l l

L i 100 -

[.

90 -

l l >

i

^

80 -

~

5

. E a.

g 70 -

=

APSR INSERTION NOT ALLOWED

( $ IN THIS TIME INTERVAL E 60 -

1

=

( W

& 50 -

3u 40 -

30 -

E E 20 -

(. 10 -

( 0 t i i i e i i i 1 90 100 0 10 20 30 40 50 60 70 80

( APSR Positian (5 Witnarawn)

I r

N

[

8-21b Babcock & Wilcox

. = co ..n c.. . ,

)

i Revision 1 (10/14/83)

Figure 8-12. APSR Position Limits, O to 24+10/-0 EFPD, Three RC Pumps - Davis-Besse 1, Cycle 4 (Tech. Spec. Figure 3.1-5f) j1 120 -

g iOO -

E I 2 80 -.(8.77) -

(38,77) l (8,69. 5) (38,69.5) 60( (0,60. 5) (42,60.5) RESTRICTED REGION 5

8 40 - PERMISSIBLE l

$ OPERATING REGION (iOO,38) l s.

% l e 20 - l l

0 I i i i t i t i i 1 0' 10 20 30 40 50 60 70 80 90 !00 APSRPosition(f. Withdrawn) 1

(

8-22 Babcock & Wilcox

. =co....n e . ,

Revision 1 (10/14/83) i L

y Figure 8-13. APSR Position Limits, 24+10/-0 to 150 10 EFPD, L. Three RC Pumps - Davis-Besse 1 Cycle 4 (Tech.

Spec. Figure 3.1-59) I1

( 120 -

g iOO ,

80 -(S77) (38,77)

( $ RESTRICTED

$ (8,69.5) (38,69.5) REGION f .

60([(0,50.5) (42,60.5) o Ue

{

,E % -

a.

{ PERMISSIBLE (100,38) g OPERATING REGION E 20 -

0 I I ' I I I I I i 1 0 10 20 30 % 50 60 70 80 90 100 APSR Position ('4 Withdrawn)

{

3 8-23 Babcock & Wilcox

. uco....n . ...,

Revision 1 (10/14/83)

~

i Figure 8-14. APSR Position Limits, 150 10 to 200 10 EFPD, Three RC Pumps - Davis-Besse 1, Cycle 4 1 Tech. Spec. Figure 3.1-5h) 120 E iOO -

s 2

(8,77) (38,77) RESTRICTED REGION 80 -

E ]

$ (8,69.5) (38,69.5)

W

$ 60 ( (0,60.5) _

48.60.5) (ggg,gg)

I C l 8

b 40 -

c.

[ PERMISSIBLE I OPERATING REGION 20 -

0 t ' I ' i ' ' ' ' '

0 10 20 30 % 50 60 70 80 90 100 APSR Posi tion (f. Wi thdrawn) 1 J

8-24 Babcock & Wilcox

. uco ..n c.... ,

Revision 1 (10/14/83)

~.

k

-- Figure 8-14a. APSR Position Limits, 200 10 to 240 t 10 1

L EFPD, Three RC Pumps - Davis-Besse 1, Cycle 4 (Tech. Spec. Figure 3.1-51) 120 E iOO -

w 2

[. >

- (8,77) (38,77) h 80 . RESTRICTED REGION W'

{' (38,69.5) w (8,69.5) h 60 t (0,60.5) Hi,60. 5)

{

t

{ T.

n.

$ 140 -

(100,38)

! PERMISSI BLE g OPERATING REGION 20 -

[

0 t ' ' ' ' ' ' ' '

0 10 20 30 14 0 50 60 70 80 90 100 APSR Position (% Withdrawn)

( ,

(

[

8-24a Babcock & Wilcox

l

{

L

/

Figure 8-15. Axial Power Imbalance Limits, 0 to 24+10/-0 EFPD, Four RC Pumps - Davis-Besse 1, Cycle 4 (Tech. Spec. Cigure 3.2-la)

(-17,102) (20,l'02)

( -

- 100

(-22,32) -

- 90 (25,92)

{

(-25,80) @-- 80 (30,80) 8 c.

(. g-- 70 th 5

( PERMISSIBLE @-- 60 OPERATING Q RESTRICTED REGION "

REG 0N % 50 Y

8

( h t #

( a.

30

{ -- 20 10 1 I I I l l

-30 -20 -10 0 10 20 30 f Axial Power imbalance (f.)

[

8-25 Babcock s.Wilcox

1fll;1l;;

>x.-m m ,oxo,. h msnro "f { tcM. NM+eONsO 1

+

M eEO aL.m s s +eo **UOfWO mmoO. ,ocn ;xn S 5nM i O g . MIa9 2

= 4

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(-26,iO2) (20,IO2)

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(-31,92) (2s,92) 3

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1, Cycle 4 (Tech. Spec. Figure 3.2-2e)

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-- 80 (17.25,77)

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1 8 4 7 4

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C 1 6 6 1 0 7 8 5 8 7 E 1 5 5 1

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_ 1 8 SAFETY X GROUP NUMBER 2 4 S t."ETY 3 5 SAFETY s 4 8 SAFETY 5 8 CONTROL 6 8 CONTROL

( 7 12 CONTROL 8

1 APSRs TOTAL

61 8-31 Babcock & Wilcox

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9. STARTUP PR0(RAW - PHYSICS TESTING

(

The planned startup test program associated with core performance is outlined

[. bel ow. . These tests verify that core perfomance is within the assumptions of the safety analysis and provide confirmation for continued safe operation of

(.: the unit.

9.1. Precritical Tests

[~ 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 f ully 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 f seconds corrected to a 75% inserted position (by rod insertica 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 f- be measured at HZP steady-state conditions. Acceptance criteria require that the measured flow be within allowable limits.

(

l 9-1 Babcock s.Wilcox

. uco....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 isothernal temperature coefficient is measured 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 ccmpen-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 coefficient 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. This 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 bani. group reactivity worths (groups 6, 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 measurenents are cal culated based on reactimeter data, and differential rod worths are obtained from the measured reactivity worth versus the change

]

9-2 Babcock & Wilcox a lacDermott company

E e

in rod group position. The differential rod worths of each of the control-L 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.

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

h 1. predicted value - measured value x100 -< 20.

measured value

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

< The credicted 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 Roo Position Core power distribution tests are performed at 40, S75, and $100%FP. The

{

test at 40% FP is essentially a check on power distribution in tne core to identify any abnormalities before escalating to the 75% FP plateau. Rod in-dex is established at a nominal 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 Babcock & Wilcox r . =c o.-.n c.. ..~

j

]

1 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 DER must be greater than 1.30. ]

3. The value obtained from the extrapolation of the minimum DER 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.

4. The value obtained from the extrapolation of the worst-case maximum LHR to the next power plateau overpower trin 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 t'e 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-i ng -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 cstablish 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 DER and LHR.

The power distribution tests performed 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:

1 Babcock & Wilcox 9-4 .mo n .,

c _

r L.

F L- 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 highen w asured 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 Inbalance 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.

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

9.3.4 Power Doppler Reactivity Coefficient at $100% FP Reactor power is decreased and then increased by about 5% FP. The reactivity change is obtained from the change in controllin9 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 f 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.

f Babcock & Wilcox 5-5 . m o.... m .....,

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 technical 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.

]

)

1

)

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Babcock & Wilcox 9-6 , m o ,, ,,,, ,,,,..,

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REFERENCES

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

2 Davis-Besse Unit 1, Final Safety Analysis Report, Docket 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-(

tion," January 14, 1980.

(- 5 .Porgram to Determine In-Reactor Perfomance of B&W Fuels - Cladding Creep Collapse, BAW-10084PA, Rev 2, Babcock & Wilcox, Lynchburg, Virginia, December 1978.

{

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

(

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 Pic Temperature and Gas Pressure Analysis, BAW-10044, Babcock

( & Wilcox, Lynchburg, Virginia, May 1972.

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

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

f Wilcox, Lynchburg, Virginia, December 1979.

11 Assembly Calculations and Fitted Nuclear Data, BAW-10116A, Eabcock &

Wilcox, Lynchburg, Virginia, May 1977.

12 Davis-Besse Nuclear Power Station Unit 1, Cycle 3 Reload Report,

. BAW-1707, Rev.1, March 1982.

[

{ Babcock & Wilcox A-1

.um..n. .m

l1

]

13 Davis-Besse Unit 1 Fuel Densification Report, BAW-1401, Babcock & Wilcox, Lynchburg, Vi rginia, 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,- Vi rginia, 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-FA Raised-loop NSS, BAW-10105, Rev.1,

. Babcock & Wilcox, Lynchburg, Virginia, July 1975

]

)

.)

l A-2 Babcock & Wilcox

_ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _

v t e Letter Sequence Other
TAC:54261, (Open)
Initiation Request, Request Acceptance... Administration Questions Answers Results Approval... Other: ML20080S485, ML20087M897
MONTHYEAR1984-02-17 [Table View]

ML20080S485

Text

SUMMARY

SUMMARY

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