ML20041G383

Cycle 3 - Reload Rept

ML20041G383
Person / Time
Site:	Davis Besse
Issue date:	02/28/1982
From:	BABCOCK & WILCOX CO.
To:
Shared Package
ML20041G380	List: ML20041G379 ML20041G381 ML20041G382 ML20041G383
References
BAW-1707, TAC-48044, NUDOCS 8203220179
Download: ML20041G383 (94)

Text

I BAW-1707 February 1982 l

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DAVIS-BESSE NUCLEAR POWER STATION UNIT 1, CYCLE 3 - RELOAD REPORT I

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I Babcock &Wilcox 8203220179 820305 i

PDR ADOCK 05000:14 6 PDP p

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.a BAW-1707 j

February 1982 iE i

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!l DAVIS-BESSE NUCLEAR POWER STATION UNIT 1, CYCLE 3 - RELOAD REPORT l

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I BABCOCK & WILCOX I

Nuclear Power Group Nuclear Power Generation Division P. O. Box 1260 i

Lynchburg, Virginia 24505 Babcock & Wilcox

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I CONTENTS I

Page 1.

INTRODUCTION AND SINMARY.

1-1 2.

OPERATING HISTORY 2-1 8

t 3.

GENERAL DESCRIPTION 3-1 4.

FUEL SYSTEM DESIGN.

4-1 a

4.1.

Fuel Assembly Mechanical Design 4-1 4.2.

Fuel Rod Design 4-1 4.2.1.

Cladding Collapse 4-1 I

4.2.2.

Cladding Stress 4-1 4.2.3.

Cladding Strain 4-1 4.3.

Thermal Design.

4-2 I

4.4.

Material Compatibility.

4-2 4.5.

Operating Experience.

4-2 5.

NUCLEAR DESIGN.

5-1 5.1.

Physics Characteristics 5-1 5.2.

Changes in Nuclear Design 5-2 8

6.

THERMAL-HYDRAULIC DESIGN.

6-1 7.

ACCIDENT AND TRANSIENT ANALYSIS 7-1 7.1.

General Safety Analysis 7-1 7.2.

Accident Evaluation 7-1 8.

PROPOSED MODIFICATIONS 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.

Reactor Coolant Flow.

9-1 9.1.3.

RC Flow Coastdown 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 I

9.2.3.

Control Rod Group Reactivity Worth.

9-2 9.2.4.

Ejected Control Rod Reactivity Worth..

9-3 I

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CONTENTS (Cont'd)

Page 9.3.

Power Escalation Tests.

9-3 9.3.1.

Core Power Distribution Verification at 440, s75, and s100% FP With Nominal Control Rod Position.

9-3 9.3.2.

Incore Vs Excore Detector Imbalance g

Correlation Verification at $40% FP 9-5

'E 9.3.3.

Temperature Reactivity Coef ficient at s100% FP 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 REFERENCES.

A-1 I

List of Tables Table 4-1.

Fuel Design Parameters.

4-3 4-2.

Fuel Thermal Analysis Parameters.

4-4 5-1.

Davis Besse 1 Cycle 3 Physics Parameters.

5-3 5-2.

Shutdown Margin Calculation for Davis Besse 1 Cycle 3 5-4 6-1.

Davis Besse Cycles 2 and 3 Thermal-Hydraulic Design Conditions 6-2 6-2.

DNBR Rod Bow Penalty, Davis Besse Cycle 3 6-2 7-1.

Comparison of Key Parameters for Accident Analysis.

7-2 7-2.

Bounding Values for Allowable LOCA Peak Linear Heat Rates.

7-3 I

List of Figures Figure 3-1.

Davis Besse 1, Cycle 3 Shuffle 3-2 3-2.

Enrichment and Burnup Distribution for Davis Besse 1, Cycle 3..

3-3 3-3.

Control Rod Locations for Davis Besse 1, Cycle 3 3-4 5-1.

BOC, Cycle 3 Two-Dimensional Relative Power Distribution -

Full Power, Equilibrium Xenon, APSRs Inserted.

5-5 g

8-1.

Reactor Core Safety Limit 8-25 g

8-2.

Reactor Core Safety Limit 8-26 8-3.

Trip Setpoint for Flux-A Flux / Flow 8-27 8-3a.

Allowable Value for Flux-AFlux/ Flow.

8-27a

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I Figures (Cont'd)

Figure Page 8-4.

Pressure / Temperature Limits at Maximum Allowable I

Power for Minimum DNBR 8-28 8-5.

Regulating Group Position Limits 0 to 60 EFPD, Four RC Pumps -- Davis Besse 1 Cycle 3 8-29 8-6.

Regulating Group Position Limits, 50 to 150 10 EFPD, I

Four RC Pumps - Davis Besse 1. Cycle 3 8-30 8-7.

Regulating Group Position Limits,150 ! 10 to 200 ! 10 EFPD, Four RC Pumps - Davis Besse 1, Cycle 3 8-31 I

8-8.

Regulating Group Position Limits, 200 ! 10 to 230 10 EFPD, Four RC Pumps - Davis Besse 1, Cycle 3 8-32 8-9.

Regulating Group Position Limits, 200 ! 10 to 268 ! 10 1

EFPD, Four RC Pumps, APSRs Withdrawn - Davis Besse 1, Cycle 3.

8-33 8-10.

Regulating Group Position Limits, O to 60 EFPD, Three RC Pumps - Davis Besse 1, Cycle 3.

8-34 I

8-11.

Regulating Group Position Limits, 50 to 150 ! 10 EFPD, Three RC Pumps - Davis Besse 1, Cycle 3............

8-35 8-12.

Regulating Group Position Limits, 150 ! 10 to 200 10 g

EFPD, Three RC Pumps -- Davis Besse 1, Cycle 3.....

8-36 3

8-13.

Regulating Group Position Limits, 200 2 10 to 230 2 10 SFPD, Three RC Pumps -- Davis Besse 1, Cycle 3.........

8-37 8-14.

Regulating Group Position Limits, 200 10 to 268 2 10 I

EFPD, Three RC Pumps, APSRs Withdrawn -- Davis Besse 1, Cycle 3.

8-38 8-15.

APSR Position Limits, O to 60 EFPD, Four RC Pumps -

0 Davis Besse 1, Cycle 3 8-39 8-16.

APSR Position Limits, 50 to 150 ! 10 EFPD, Four RC Pumps -- Davis Besse 1, Cycle 3 8-40 8-17.

APSR Position Limits, 150 10 to 200 ! 10 EFPD, I

Four RC Pumps -- Davis Besse 1, Cycle 3 8-41 8-18.

APSR Position Limits, 200 ! 10 to 230 ! 10 EFPD, Four RC Pumps -- Davis Besse 1, Cycle 3 8-42 I

8-19.

APSR Position Limits, 200 ! 10 to 268 ! 10 EFPD, Three or Four RC Pumps APSRs Withdrawn - Davis Besse 1, Cycle 3..

8-43 g

8-20.

APSR Positic Limits, O to 60 EFPD, Three RC Pumps -

3 Davis Besse 1, Cycle 3 8-44 8-21.

APSR Position Limits, 50 to 150 10 EFPD, Three RC Pumps -- Davis Besse 1, Cycle 3 8-45 I

8-22.

APSR Position Limits, 150 10 to 200 10 EFPD, Three RC Pumps -- Davis Besse 1, Cycle 3............

8-46 8-23.

APSR Position Limits, 200 ! 10 to 230 10 EFPD, l

Three RC Pumps - Davis Besse 1, Cycle 3..

8-47 m

8-24.

Axial Power Imbalance Limits, O to 60 EFPD, Four RC Pumps -- Davis Besse 1, Cycle 3 8-48 8-25.

Axial Power Imbalance Limits, 50 to 150 ! 10 EFPD, I

Four RC Pumps -- Davis Besse 1, Cycle 3 8-49 8-26.

Axial Power Imbalance Limits, 150 10 to 200 2 10 EFPD, Four RC Pumps - Davis Besse 1, Cycle 3 8-50 I

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Figures (Cont'd) pf Page 8-27.

Axial Power Imbalance Limits, 200 ! 10 to 230 10 EFPD, Four RC Pumps - Davis Besse 1, Cycle 3 8-51 8-28.

Axial Power Imbalance Limits, 200 10 to 268 ! 10 EFPD, Four RC Pumps, APSRs Withdrawn - Davis Besse 1, Cycle 3....

8-52 8-29.

Axial Power Imbalance Limits, O to 60 EFPD, Three RC E

Pumps - Davis Besse 1, Cycle 3 8-53 g

8-30.

Axial Power Imbalance Limits, 50 to 150 ! 10 EFPD, Three RC Pumps - Davis Besse 1, Cycle 3.....

8-54 8-31.

Axial Power Imbalance Limits,150 ! 10 to 200 10 E7PD, Three RC Pumps - Davis Besse 1, Cycle 3......

8-55 8-32.

Axial Power Imbalance Limits, 200 ! 10 to 230 ! 10 EFPD, l

Three RC Pumps - Davis Besse 1, Cycle 3...

8-56 8-33.

Axial Power Imbalance Limits, 200 10 to 268 ! 10 EFPD, 5

Three RC Pumps, APSRs Uithdrawn - Davis Besse 1 Cycle 3 8-57 8-34.

Control Rod Core Locations and Group Assignments -

Davis Besse 1, Cycle 3 8-58 I

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1.

INTRODUCTION AND

SUMMARY

I This report justifies operation of the Davis Besse Nuclear Power Station Unit 1 at the rated core power of 2772 MWt for cycle 3.

The required analyses are I

included as outlined in the USNRC document, " Guidance for Proposed License Amendments Relating to Refueling," June 1975. This report utilizes the analyt-ical techniques and design bases documented in several reports that have been submitted to the USNRC and approved by that agency.

I Cycle 3 reactor and fuel parameters related to power capability are summarized in.this report and compared to cycle 2.

All accidents analyzed in the Davis

.I Besse FSAR have been reviewed for cycle 3 operation and in cases where cycle 3 characteristics were conservative compared to cycle 1, no new analyses were performed.

l Retainers will be installed on the two fuel assemblies (FAs) containing neu-tron sources. The retainers will provide positive retention during reactor operation. The effects on continued operation without orifice rod assemblies I

(ORAs) and.the addition of' retainers have been accounted for in the analysis performed for cycle 3.

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The Technical Specifications, have been reviewed and modified where required for cycle 3 operation. ~ Based on the' analyses performed, taking into account the ECCS Final Acceptance Criteria and postulated fuel densification effects, it is concluded that Davis Besse Unit 1, cycle 3 can be operated safely at its licensed. core power level of 2772 MWt.

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OPEttATIEG h1 STORY 4

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The reference cycle for the nuclear and thermal-hydraulic analyses of Davis Besse Unit 1 is the currently operating cycle 2, which achieved criticality on l

November 1, 1980. Power escalation began on November 3, 1980, and full power l '

(2'172 MWt) was reached on December 3, 1980. No operating anomalies occurred l.

during cycle 2 ot.etation that would adversely affect fuel performance during l

cycle 3.

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3.

GENERAL DESCRIPTION I

The Davis Besse Unit I reactor core is described in detail in Chapter 4 of the Final Safety Analysis Report for the unit.2 The cycle 3 core consists of 177 FAs, each of which is a 15 by 15 array containing 208 fuel rods, 16 control. rod guide tubes, and one incore instrument guide tube.

All FAs in oatches 4, 5A, I

and SB have a constant nominal fuel loading of 468.25 kg of uranium.

Batches IC and 3 have a fuel loading of 472.24 kg of uranium. The fuel consists of dished-end cylindrical pellets of uranium dioxide clad in cold-worked Zircaloy-4.

The undensified 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.

Figure 3-1 is the core loading diagram for Davis Besse 1 cycle 3.

Twelve batch IB assemies and 61 batch 2 assemblies will be discharged at the end of cycle 2.

All batch 3 and 4 assemblies will be shuffled to new locations. Batches 3 and 4 have initial uranium-235 enrichments of 2.96 and 3.04 wt %, respective-ly.

Twenty-five batch IC assemblies with an initial enrichment of 1.98 wt %

I will be reinserted in cycle 3.

A split-enrichment feed batch consisting of 8 batch SA and 40 batch 5B assemblies, with uranium enrichments of 3.04 and 2.99 wt %, respectively, will be inserted in cycle 3 and will primarily occupy the periphery of the core. Figure 3-2 is an eighth-core map shc, wing each assembly's burnup at the beginning of cycle 3 (BOC-3) and its initial enrichment.

Cycle 3 will be operated in a feed-and-bleed mode. The core reactivity control will be supplied 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 con-trol of the axial power distribution. The cycle 3 locations of the 61 control rods and the group designations are indicated in Figure 3-3.

Although the rod I

group designations differ, the core locations of the 61 control rods for cycle 3 are identical to those of the reference cycle 2.

3-1 Babcock & Wilcox

I Figure 3-1.

Davis Besse 1, Cycle 3 Shuffle FUEL TRANSFER CANAL

=

X l

5A

$8

$8 58 SA A

F F

F F

F 58 58 4

3 3

3 4

58 58 F

F LI 09 03 07 Ll5 F

F IC 58 3

3 4

IC 4

3 3

58 IC C

Nao F

MS Pio N3 E8 N13 P6 M7 F

L4 (evl)

(cyl)

(cyll 58 58 3

4 BC IC 58 IC IC 4

3 58 58 0

F F

Hl0 4

09 M9 F

M7 07 Min L8 F

F (cyll (cyl)

(cyl) (cyl) 58 3

=

3 4

3 3

3 4

3 4

3 56

[

F Xil Bil E9 C14 Mll NB 45 D2 K7 85 E5 F

SA 4

3 lC 4

4 3

4 3

4 4

IC 3

4 SA W

F AIO Lin El3 P4 R9 MIO R8 N6 El P12 K3 L2 A6 F

F (cyi)

(cyi) 58 3

4 BC 3

3 4

3 4

3 3

IC 4

3 58 F

E13 Cl2 Kit Ml2 Lil KI5 P8 R7 L5 M4 55 C4 K3 F

G (cyi)

(cyi) 58 3

IC 58 3

4 3

IC 3

4 3

58 sc 3

58 W-F 013 M9 F

Ml2 HIS M14 LIO H2 Hi H4 F

M7 C3 F

- Y H

(cyi)

(cyi; (cyi) 58 3

4 BC 3

3 4

3 4

3 3

6C 4

3 58 F

Gl3 012 Gil E12 Fil A9 88 G1 F5 E4 G5 04 G3 F

E (cyl)

(cyl)

SA 4

3 IC 4

4 3

4 3

4 4

IC 3

SA F

RIO Fl4 G13 84 GIS E10 48 E6 A7 512 G3 F2

1. 6 F

b (cyl)

(cyl) 58 3

4 3

4 3

3 3

4 3

4 3

58 M

F Gil Pal G9 NI4 Dit 08 05 N2 G7 P5 G5 F

58 58 3

4 lC IC 58 BC IC 4

3 58 58 N

F F

F8 E2 C9 E9 F

E7 C7 Els n6 F

F (evi (evt) f evt )

(evt)

IC 58 3

3 4

sc 4

3 3

58 IC 0

F12 F

E9 810 D3 G8 013 86 E7 F

D6 (evt1 (e vt I fevti 58 58 4

3 3

3 4

58 58 P

F F

F1 C9 Cia C7 Fi$

r F

SA 58 58 58 5A F

F F

F F

I 1

2 3

4 5

6 7

8 9

10 11 12 13 14 15 BATCH X-~ PREVIOUS CORE LOCATION NOTE: Cy 1 - REINSERTED FROM CYCLE 1 I

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Figure 3-2.

Enrichment and Burnup Distribution for Davis Besse 1, Cycle 3 8

9 10 11 12 13 14 15 1.98 2.96 3.04 2.96 2.99 1.98 2.96 2.99 I

H 12644 18960 9866 25564 0

12376 14508 0

3.04 2.96 2.96 1.98 3.04 2.96 2.99 9221 23117 23621 13364 12000 19260 0

3.04 3.04 1.98 2.96 3.04 3.04 L

9222 7673 13170 20258 7141 0

2.96 3.04 2.96 2.99 y

20103 10045 21260 0

2.96 2.99 2.99 5

N 21926 0

0

.98 0

ii m P

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x.xx INITIAL ENRICHMENT

...,,, -e0c eu or.

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1 Ii Figure 3-3.

Control Rod Locations for Davis Besse 1, Cycle 3 t

A 4

8 4

7 4

C 1

6 6

1 0

7 8

5 8

7 E

1 5

5 1

F 4

8 3

7 3

8 4

G 6

2 2

6 HW-7 5

7 3

7 5

7

-Y K

6 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 1

P 4

7 4

I 1

2 3

4 5

6 7

8 9

10 11 12 13 14 15 GROUP NO. OF RODS FUNCTIONS X- - GROUP NUMBER 1

8 SAFETV 2

4 SAFETY 3

5 SAFETY 4

8 SAFETY 5

8 CONTROL E

6 8

CONTROL g

7 12 CONTROL 8

8 APSRs TOTAL #5 I

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4.

FUEL SYSTEM DESIGN I

4.1.

Fuel Assembly Mechanical Design The types of FAs and pertinent fuel parameters for Davis Besse 1 cycle 3 are listed in Table 4-1.

All Mark B FAs are identical in concept and are mechani-cally interchangeable.

Retainer assemblies will be used on the two FAs that contain the regenerative neutron sources. The justification for the design and use of the retainer assemblies is described in references 1 and 3.

4.2.

Fuel Rod Design The fuel rod design and mechanical evaluation are discussed below.

4.2.1.

Cladding Collapse The fuel of batch 3 is more limiting than that of batches 1, 4, and 5 because of its previous incore exposure time.

The batch 3 assembly power histories were analyzed to determine 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 1 cycle 3.

The generic analysis was based on reference 4 and is applica-ble to the batch 3 design.

The creep collapse analysis predicts a collapse time longer than 30,000 EFPH, which is longer than the expected residence time of 22,080 EFPH (Table 4-1).

4.2.2.

Cladding Stress The Davis Besse 1 cycle 3 stress parameters are enveloped by a conservative fuel rod stress analysis. No method was used for analysis of cycle 3 that had not been used on the previous cycle.

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 cladding strain is less than 1% at the design local pellet burnup and heat generation 4-1

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The design values are higher than the worst-case values the Davis Besse 1 cycle 3 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.

Thermal Design All fuel assemblies in the Davis Besse cycle 3 core are thermally similar E

Fuel dimensions, linear heat rate (LHR) capabilities, and average fuel tempera-W tures for each batch for cycle 3 are summarized in Table 4-2.

LHR capabili-ties are based on centerline fuel melt and are established with the TAFY-3 5

code using the methods and models described in reference 6.

The maximum in-ternal pin pressure at EOL is less than the nominal system pressure of 2200 psia based on a maximum fuel pin burnup of 36,597 mwd /mtU.

4.4.

Material Compatibility The compatibility of all possible fuel cladding-coolant assembly interactions for batch 5 FAs is identical to that of present fuel.

4.5.

Operating Experience Operating experience with the Mark B 15 by 15 fuel assembly has verified the adequacy of its design. As of December 31, 1981, the following experience has been accumulated for eight B&W 177-fuel assembly plants using the Mark B fuel assembly:

Max FA burnup(a)

Cumulative Current net electric Reactor cycle Incore Discharged output, MWh(b)

Oconee 1 7

40,000 40,000 37,363,143 Oconee 2 5

36,800 33,700 33,544,897 Oconee 3 6

33,000 32,060 33,504,857 TMI-l 5

25,000 32,400 23,840,053 ANO-1 5

29,780 33,220 29,642,809 l

Rancho Seco 5

31.160 37,730 26,159,475 Crystal River 3 4

17,400 29,900 17,534,439 Davis Besse 2

23,360 13,250 11,375,238 I

(")As of December 31, 1981.

)As of September 30, 1981.

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Fuel Design Parameters Batch IC Batch 3 Batch 4 Batch SA Batch 5B Fuel assembly type Mark B4A Mark B4A Mark B4A Mark B4A Mark B4A Number of assemblies 25 60 44 8

40 Fuel rod OD, in.

0.430 0.430 0.430 0.430 0.430 Fuel rod ID, in.

0.377 0.377 0.377 0.377 0.377 Flexible spacer type Spring Spring Spring Spring Spring Rigid spacer type Zirc-4 Zirc-4 Zirc-4 Zirc-4 Zirc-4 Undensified active fuel length, in.

143.5 143.5 143.44 143.44 143.20 Fuel pellet (mean) dia., in.

0.3675 0.3675 0.3697 0.3697 0.3686 Fuel pellet initial density,

% TD mean 96 96 94 94 95 t,.

i" Initial fuel enrichment, 2ss wt %

U l.98 2.96 3.04 3.04 2.99 Estimated residence time, EFPli 21,936 22,080 20,304 20,160 20,160 Cladding collapse time. EFPil

>30,000

>30,000

>30,000

>30,000

>30,000

?

8x 90 n-

I Table 4-2.

Fuel Thermal Analysis Parameters Batches IC/3 Batches 4/5A Batch SB No. of assemblies 25/60 44/8 40 Nominal pellet density, % TD 96 94 95 Pellet diameter, in.

0.3675 0.3697 0.3686 Stack height, in.

143.5 143.44 143.20 Densified Fuel Parameters Pellet diameter, in.("}

0.3651 0.3648 0.3649 Stack height, in. ( }

143.14 141.65 142.13 Nominal LilR @ 2772 MWt, kW/ft 6.14 6.21 6.19 Avg fuel temp @ nominal LHR, F 1340 1355 1370 LHR to fuel melt, kW/ft 20.4(')

20.4 20.3 Note: Core average densified LHR: 6.17 kW/ft.

I

(^ Densified from lower tolerance limit on density to 96.5% TD.

(

Densified from nominal density to 96.5% TD.

Five batch 3 fuel assemblies have au LHR to fuel melt of 20.35 kW/ft.

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5.

NUCLEAR DESIGN I

5.1.

Physics Characteristics Table 5-1 compares the core physics parameters of cycle 2 with those of both the base and alternate designs for cycle 3.

These values were generated using PDQ07 for both cycles.7-S Since the core has not yet reached an equilibrium cycle, differences in core physics parameters are to be expected between the cycles. Figure 5-1 illustrates a representative relative power distribution I

for BOC-3 at full power (FP) with equilibrium xenon and group 8 inserted.

The critical boron concentrations for cycle 3 differ from those of the refer-I ence cycle 2 due to the difference in design cycle lengths.10'll The hot full power (HFP) control rod worths are different because in cycle 3 the bank loca-tions and designations have changed from those of the reference cycle. Control rod worths are sufficient to maintain the required shutdown margin, as indicated in Table 5-2.

The ejected rod worths in Table 5-1 are the maximum calculated values.

It is difficult to compare maximum ejected rod worths between cycles since neither the rod patterns from which the rod is assumed to be ejected nor the isotopic distributions are identical.

Calculated ejecte.d rod worths and their adherence to criteria are considered at all times in life and at all power I

levels in the development of the rod position limits presented in section 8.

Tia maximum stuck rod worths for cycle 3 are also different than those for cycle I

2.

The adequacy of the shutdown margin with cycle 3 stuck rod worths is shown in Table 5-2.

The following conservatisms were applied for the shutdown calcu-lations:

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 cal-culated using a two-dimensional model. The cycle 3 power deficits from HZP I

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to HFP are more negative than those for cycle 2 due to the more negative moder-ator coefficients in cycle 3.

The differential boron and xenon worths are similar in both cycles.

The effective delayed neutron fractions for cycle 3 show a decrease with burnup (also reported for reference cycle 2).

5.2.

Changes in Nuclear Design There are no significant core design changes between the reference cycle and cycle 3.

The same calculational methods and design information were used to obtain the important nuclear design parameters. No significant operational or procedural changes exist for the base design with regard to axial or radial power shape, xenon, or tilt control. The alternate design includes an APSR pull at 200 EFPD and a planned power coastdown to EOC at 268 EFPD. The stabil-ity and control of the core with APSRs withdrawn have been analyzed. The cal-culated stability index without APSRs is -0.0314 h-1, which demonstrates the g

axial stability of the core.

The operational and RPS limits (Technicti Speci-5 fication changes) for cycle 3 are presented in section 8.

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Table 5-1.

Davis Besse 1 Cycle 3 Physics Parameters Cycle 3 Base Alternate Cycle 2 desien(b) design Cycle length. EFPD 306 230 268 Cycle burnup. mwd /mtU 10.167 7.663 8.929 Average core burnup - EOC. mwd /mtU 19.340 18.925 20.191 Initial core loading, mtc 83.4 83.2 83.2 Critical boron - BOC. no Xe. ppm HZP(8){ gr up 8 inserted 1.379 1.231 1.231 HFP.

J 1.197 1.015 1.015 I

Critical boron - EOC eq Xe. ppm HFP[ group 8 withdrawn 10 49 10 I

Control rod worths - HTP. BOC % ak/k Group 6 0.98 0.93 0.93 Group 7 1.47 1.52 1.52 Group 8 0.34 0.31 0.31 I

Control rod worths - HTP, EOC % ak/k Group 7 1.54 1.58 1.53 Group 8 NA 0.39 NA Max ejected rod worth - HZP. % ik/k BOC. groups 5-8 inserted 0.77 0.78 0.78 I

IC EOC. groups 5-8 inserted 0.77 0.76 0.76 Max stuck rod worth - HZP. % ak/k BOC 1.14 1.44 1.44 EOC l.06 1.43 1.43 I

Powe r de fic it. HZP to HFP. eq Xe % ak/k BOC (4 EFPD)

-1.51

-1.79

-1.79 EOC

-2.49

-2.36

-2.34 I

Doppler coeff - HTP. 10 ak/k/*F BOC. no Xe

-1.44

-1.46

-1.46 EOC. eq Xe

-1.58

-1.58

-1.63 Modarator coeff - HFP. 10-' t.k/k/*F I

BOC. no Xe. 1015 ppm. group 8 inserted

-0.56

-1.13

-1.13 EOC eq Xe. 10 ppm. group 8 withdrawn

-2.95

-3.00

-2.89 Boron worth - HFP. ppm /% ak/k BOC (1015 ppmb) 111 110 110 EOC (10 ppm) 96 98 97 I

Xenon worth - HFP. % ak/k BOC (4 EFPD) 2.67 2.67 2.67 EOC (equilibrium) 2.73 2.74 2.73 I

Effective delayed neutron fraction - HFP BOC 0.00594 0.00595 0.00595 EOC 0.00526 0.00537 0.00530 I*}HZP denotes hot zero power (532F Tavg); HFP denotes hot full power (584F core Tavg)*

Croup 8 is inserted at EOC in the base cycle 3 design.

I (c)For the alternate cycle 3 design. stuck and ejected rod worths are for 200 EFPD with group 8 inserted (EOC values are smaller).

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5-3 Babcock & Wilcox

I l

lIl Table 5-2.

Shutdown Margin Calculation for Davis Besse 1 Cycle 3 EOC, % Ak/k Alternate design BOC Base 200 EFPD, 268 EFPD, l

% Ak/k, design bank 8 in bank 8 out 5

Available Rod Worth Total rod worth, HZP(")

7.46 7.78 7.74 7.78 Worth reduction due to burnup of poison material

-0.10

-0.11

-0.11

-0.11 Maximum stuck rod, HZP

-1.44

-1.43

-1.43

-1.25 Net worth 5.92 6.24 6.20 6.42 Less 10% uncertainty

-0.59

-0.62

-0.62

-0.64 Total available worth 5.33 5.62 5.58 5.78 Required Rod Worth Power deficit, HFP to HZP 1.79 2.36 2.29 2.34 Max allowable inserted rod worth 0.38 0.58 0.58 0.58 g

Flux redistribution 0.77 1.15 1.13 1.18 E

Total required worth 2.94 4.09 4.00 4.10 Shutdown Margin Total available minus total required 2.39 1.53 1.58 1.68 Note: Required shutdown margin is 1.00% Ak/k.

(")HZP: hot zero power (532F Tayg), HFP: hot full power (584F core Tavg)-

t l

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5-4 Babcock & Wilcox

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Figure 5-1.

BOC (4 EFPD), Cycle 3 Two-Dimensional Relative Power Distribution - Full Power, Equilibrium I

Xenon APSRs Inserted W 8

9 10 11 12 13 14 15 l

1.00 1.14 1.21 0.94 1.36 0.93 1.02 0.96 K

1.28 1.04 0.94 0.87 1.07 0.96 0.92 I

8 l

1.28 1.22 0.77 0.92 1.08 0.74 I

M 1.05 1.13 0.97 1.04 N

1.03 1.20 0.76 0

0.55 I

P I

R R

I INSERTED R00 GROUP NUMBER x.x x RELATIVE POWER DENSITY (a) CALCULATED RESULTS FROM TWD-OlMENSIONAL PIN BY-PIN P0007.7 I

l 5-5 Babcock & Wilcox

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I 6.

THERMAL-HYDRAULIC DESIGN I

The fresh Batch 5 fuel is hydraulically and geometrically similar to the other fuel loaded into the cycle 3 core. The thermal-hydraulic design evaluation I

supporting cycle 3 operation is based on the methods and models described in references 6 and 12.

The cycle 3 thermal-hydraulic design is identical to that of cycle 2.

The thermal-hydraulic design conditions for cycles 2 and 3 are summarized in Table 6-1.

The rod bow penalty for cycle 3 is based on the procedure approved by reference 13.

A rod bow penalty was calculated for each fuel batch based on the highest assembly burnup in that batch. DNBR analyses were then performed based on the highest pin peak for each batch during cycle 3.

Results summarized in Table 6-2 show that the fresh batch 5 fuel is the most restrictive on DNBR even when rod bow penalties are included. Therefore, the cycle 3 rod bow penalty is based on the highest predicted batch 5 assembly burnup of 11,986 mwd /mtU. The I

resulting rod bow penalty is 0.0% when 1% credit is taken for the use of a flow area reduction hot channel factor in DNBR calculations.

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6-1 Babcock & Wilcox

I Table 6-1.

Davis Besse Cycles 2 and 3 Thermal-Hydraulic Design Conditions Design power level, MWt 2772 System pressure, psia 220 E

Reactor coolant flow, % design 110 W

Vessel inlet / outlet coolant temp.,

100% power, F 557.7/606.3 Ref design radial-local power peaking factor 1.71 g

Ref design axial flux shape 1.5 cosine 5

with tails Hot channel factors Enthalpy rise (F )

1.011 9

Heat flux (F")

1.014 9

Flow area 0.98 g

Active fuel length See Table 4-2 W

Avg heat flux, 100% power, 10 (a) 5 Btu /h-ft2 1.89 x Max heat flux, 100% power, 10 (a) 5 Btu /h-ft2 4.85 x CHF correlation BAW-2 Minimum DNBR, (% power) 1.79 I

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

I Table 6-2.

DNBR Rod Eow Penalty, Davis Besse Cycle 3 l

DNBR less l

Max BU Rod Bow Max pin

DNBR, rod bow ^

l Batch mwd /mtU penalty, %

peak BAW-2 penalty IC 21,360 1.8 1.05 3.58 3.56 3

33,902 4.6 1.22 2.90 2.80 g

4 21,743 1.8 1.42 2.21 2.19

=

5 11,986

<0.5 1.48 2.03 2.03 I

(" Includes 1% credit for flow area reduction hot channel factor.

l 6-2 Babcock & Wilcox l

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I 7.

ACCIDENT AND TRANSlENT ANALYSIS 7.1.

General Safety Analysis 2

Each FSAR accident analysis has been examined with respect to changes in the cycle 3 parameters to determine the effects of the cycle 3 reload and to ensure that thermal performance during hypothetical transients is not degraded. The effects of fuel densification on the FSAR accident results have been evaluated and are reported in reference 6.

I Improved fuel utilization and the inherent increase in core average burnup examined in cycle 3 have resulted in a higher plutonium-to-uranium fission ratio than that used in the FSAR. A comparison of the cycle 3 iodine and noble gas inventories with those used in the FSAR Chapter 15 analyses showed that the I

thyroid and while body doses were well below 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, I

(2) thermal-hydraulic, and (3) kinetics parameters including the reactivity feedback coefficients and control rod worths.

Fuel thermal analysis parameters from each batch in cycle 3 are given in Table 4-2.

A comparison of the cycle 3 thermal-hydraulic maximum design conditions to the previous cycle values is presented in Table 6-1.

A comparison of the key kinetics parameters from the FSAR and cycle 3 is provided in Table 7-1.

A generic LOCA analysis for B&W 177-FA raised-loop NSS has been performed using the Final Acceptance Criteria ECCS Evaluation Model."

The combination of aver-age fuel temperature as a function of LHR and the lifetime pin pressure data used in the LOCA limits analysis is conservative compared to those calculated for this reload. Thus, the analysis and the LOCA limits reported in reference 14 provide conservative results for the operation of Davis Besse 1 cycle 3 fuel. A tabulation showing the bounding values for allowable LOCA peak linear I

7-1 Babcock & Wilcox

I heat rates for Davis Besse 1 cycle 3 fuel are provided in Table 7-2.

The basis for two sets of LOCA limits is provided in reference 15.

It is concluded by examination of cycle 3 core thermal, thermal-hydraulic, and kinetics properties with respect to acceptable previous cycle values that this g

core reload will not adversely affect the ability to safely operate the Davis W

)

Besse 1 plant during cycle 3.

Considering the previously accepted design basis used in the FSAR and subsequent cycles, the transient evaluation of cycle 3 is considered to be bounded by previously accepted analyses. The initial condi-tions of the transients in cycle 3 are bounded by the FSAR and/or the fuel densification report.

I Table 7-1.

Comparison of Key Parameters for Accident Analysis FSAR and densif'n Base Alternate um report cycle 3 cycle 3 Parameter va'ue value value BOL Doppler coeff, 10-5, Ak/k/ F

-1.28

-1.46

-1.46 EOL Doppler coeff, 10-5, Ak/k/*F

-1.45(^)

-1.58

-1.63 BOL moderator coeff, 10-", Ak/k/ F

+0.13

-1.13

-1.13 EOL moderator coeff, 10-", Ak/k/*F

-3.0

-3.0

-2.89 All rod bank worth (HZP), % Ak/k 10.0 7.46 7.46 Boron reactivity worth (HFP), ppm /1% Ak/k 100 110 110 Max ejected rod worth (HFP), % Ak/k 0.65 0.38 0.34 Max dropped rod worth (HFP), % Ak/k 0.65 0.20 0.20 Initial boron cone (HFP), ppm 1407 1015 1015

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

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I 7-2 Babcock & Wilcox I

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Table 7-2.

Bounding Values for Allowable LOCA Peak Linear Heat Rates Allowable Allowable

' g Core peak LHR, peak LHR, 3

elevation, first 50 EFPD, balance of cycle, ft kW/ft kW/ft 2

16.0 16.5 4

16.8 17.2 6

18.0 18.4 8

17.5 17.5 10 17.0 17.0 I

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I 7-3 Babcock & Wilcox

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8.

PROPOSED MODIFICATIONS TO TECHNICAL SPECIFICATIONS I

The Technical Specifications have been revised for cycle 3 operation to ac-count for changes in power peaking and control rod worths. In addition, changes were the result of the following:

1.

The RPS trip setpoints were calculated with the effects of revised instru-I mentation errors included. The revised errors include a component for drift and do not break it out as a separate term. Consequently, the al-lowable values and the setpoints in Table 2.2-1 are equal and Tech Spec Figure 2.2-2 has been deleted.

2.

Protection has been provided for an overpower condition that could occur during an overcooling transient.

3.

The power level cutoff at 92% full power has been eliminated.

These proposed Technical Specifications include two sets of operating limits for the period af ter 200 EFPD. This will allow operating flexibility during the latter portion of the fuel cycle.

Based on the Technical Specifications derived from the analyses presented in this report, the Final Acceptance Criteria ECCS limits will not be exceeded, nor will the thermal design criteria be violated.

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8-1 Babcock & Wilcox

Table 2..-l.

Reactor Protection System Instrumentation Trip Setpoints Functional unit Trip setpoint Allowable values 1.

Fbnual reactor trip Not applicable.

Not applicable.

2.

High flux 1104.97. of RATED THERMAL POWER with

$104.9% of RATED THERMAL POWER with four pumps operating four pumps operating #

179.5% of RATED THERMAL POWER with 179.5% of RATED THERMAL POWER with three pumps operating three pumps operatingF 3.

RC high temperature 1618 F 5618*F 4.

Flux - A flux / flow ( }

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 8 5.

RC low pressure (

21985 psig 21985.0 psig*

21985.0 psig**

w 6.

RC high pressure s2300 psig s2300.0 psig*

s2300.0 psig**

w RC pressure-temperature (

2(12.60 T

• F - 5600) psig 2(12.60 T

• F - 5660) psig#

7.

g g

8.

High flux / number of RC

$55.0% of RATED THERMAL POWER with 555.0% of RATED TilERMAL POWER with pumps on(l) one pump operating in each loop one pump operating in each loop #

s0.0% of RATED THERMAL POWER with 50.0% 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 pump operating in the other loop #

s0.07. of RATED TilERMAL POWER with no s0.0% of RATED THERMAL POWER with no pumps operating or only one pump op-pumps operating or only one pump op-m erating erating#

mcr g

g 9.

Containment pressure 54 psig s4 psig gp high e

!E m

e e

e m

e e

m m

Eis e

m e

m e

m e

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

m m

m W

W M

M M

M M

M M

M M

M m

W W

W Table 2.2-1.

(Cont'd)

(

Trip may be manually bypassed when RCS pressure s1820 psig by actuating shutdown bypass provided that:

I a.

The high flux trip setpoint is s5% of RATED TilERMAL POWER.

b.

The shutdown bypase high pressure trip setpoint of 41820 psig is imposed.

I c.

The shutdown bypass is remcved when RCS pressure >1820 psig, i

I

• Allowable value for CHANNEL FUNCTIONAL TEST.

• • Allowable value for CHANNEL CALIBRATION.

1

1. Allowabl e value for CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION.

i e

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}

fu 1

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8x i

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I 2.1.

SAFETY LIMITS BASES I

2.1.1 and 2.1.2 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel cladding and possible cladding perforation which would result in the release of fission products to the reactor coolant.

Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime would re-sult in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat trans-g fer coefficient. DNB is not a directly measurable parameter during operation g

and therefore THERR\\L POWER and reactor coolant temperature and pressure have been related to DNB through the B&W-2 DNB correlation.

The DNB correlation has been developed to predict the DNB flux and the location of DNB for axially l

uniform and non-uniform heat flux distributions. The local DNB heat flux 5

ratio, D'iBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.

The minimum value of the DNBR during steady state operation, normal operation-al transients, and anticipated transients is limited to 1.30.

This value cor-responds to a 95 percent probability at a 95 percent confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions.

The curve presented in Figure 2.1-1 represents the conditions at which a mini-mum DNBR of 1.30 is predicted for the maximum possible thermal power 112% when g

the reactor coolant flow is 387, 200 gpm, which is 110% of design flow rate g

for four operating reactor coolant pumps. This curve is based on the follow-ing hot channel factors with potential fuel densification and fuel rod bowing effects:

[Z 1.50 F = 2.56; F

= 1.71;

=

Q AH 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 withdrawal, and form the core DNBR design basis.

DAVIS-BESSE, UNIT 1 B 2-1 8-4

I S'.FETY LIMITS I

BASES I

The reactor trip envelope appears to approach the safety limit more closely than it actually does because the reactor trip pressures are measured at a I

location where the indicated pressure is about 30 psi less than core outlet pressure, providing a more conservative margin to the safety limit.

I 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 1.30 DNBR limit produced by a nuclear power peaking factor of Fq =

2.56 or the combination of the radial peak, axial peak, and position of 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 IC, 4, and 55 I

batch 3 assemblies; 20.35 for the five remaining batch 3 assemblies; and 20.3 for batch 5.

Power peaking is not a directly observable quantity and therefore limits have I

been established on the basis of the reactor power imbalance produced by the power peaking.

I 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 I

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

DNBR of 1.30 is predicted at the maximum possible thermal power for the number I

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 ef fects of fuel rod bow and f uel densification.

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. Extrapolation of the correlation beyond its published quality range of +22% is justified on I

the basis of experimental data.

I lI DAVIS-BESSE, UNIT 1 B 2-2 I

l l

lI 8-5

I SAFETY LIMITS BASES I

For the curves of BASES Figure 2.1, a pressure-temperature point above and to l

the left of the curve would result in a DNBR greater than 1.30 or a local g

quality at the point of minimum DNBR less than +22% for that particular reac-g tor coolant pump situation. The 1.30 DNBR curve for three pump operation is l

more restrictive than any other reactor coolant pump situation because any pressure / temperature point above and to the left of the three pump curve will l

be above and to the left of the four pump curve.

W 2.1.3.

REACTOR COOLANT SYSTDI PRESSURE The restriction of this safety limit protects the integrity of the reactor cool-ant system from overpressurization and thereby prevents the release of radio-nuclides contained in the reactor coolant from reaching the containment atmos-phere.

The reactor pressure vessel and pressurizer are designed to Section III of the g

ASME Boiler and Pressure Vessel Code which permits a maximum transient pres-g sure of 110%, 2750 psig, of design pressure. The reactor coolant system pip--

ing, valves and fittings, are designed to ANSI B 31.7, 1968 Edition, which g

permits a maximum transient pressure of 110%, 2750 psig, of component design g

pressure.

The safety licit of 2750 poig is therefore consistant with the de-sign criteria and associated code requirements.

The entire reactor coolant system is hydrotested at 3122 psig, 125% of design pressure, to demonstrate integrity prior to initial operation.

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I DAVIS-BESSE, UNIT 1 B 2-3 l

8-6

J 2.2.

LIMITING SAFETY SYSTEM SETTINGS I

BASES 2.2.1.

REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS I

The reactor protection system instrumentation trip setpaint specified in Table 2.2-1 are the values at which the reactor trips are set for each parameter.

g The trip setpoints have been selected to ensure that the reactor core and re-

,3 actor coolant system are prevented from exceeding their safety limits.

'g The shutdown bypass provides for bypassing certain functions of the reactor g

protection system in order to permit control rod drive tests, zero power PHYSICS TESTS and certain startup and shutdown procedures. The purpose of the shutdown bypass high pressure trip is to prevent normal operation with shut-j down bypass activated. This high pressure trip setpoint is lower than the 5

normal low pressure trip setpoint so that the reactor must be tripped before the bypass is initiated. The high flux trip setpoint of s5.0% prevents any g

significant reactor power from being produced.

Sufficient natural circulation g

would be available to remove 5.0% of RATED THERMAL POWER if none of the reac-tor coolant pumps were operating.

I Manual Reactor Trip I

The manual reactor trip is a redundant channel to the automatic reactor pro-tection system instrumentation channels and provides manual reactor trip capa-bility.

I High Flux I

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 l

power level reaches 104.9% of rated power. Due to calibration 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 I

I

" ^ ' ' " - " " " " " " * " '

8-7

LIMITING SAFETY SYSTEM SETTINGS II BASES I

o RC High Temperature The RC high temperature trip 5618 F prevents the reactor outlet temperature l

from exceeding the design limits and acts as a backup trip for all power ex-cursion transients.

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

i 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 protection for all modes of pump operation. For every flow rate there is a maximum per-missible power level, and for every power level there is a minimum permissible low flow rate.

Examples of typical power level and low flow rate combinations 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 106.4% and reactor coolant flow rate is 100% of full flow rate, or flow I

rate is 93.9% of full flow rate and power level is 100%.

g 2.

Trip would occur when three reactor coolant pumps are operating if power is 79.5% and reactor coolant flow rate is 74.7% of full flow rate, or flow rate is 70.4% of full flow rate and power is 75%.

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

I I

I DAVIS-BESSE, UNIT 1 B 2-5 8-8

'I

~

LIMITING SAFETY SYSTEM SETTINGS BASES _

I The AXIAL POWER IMBALANCE boundaries are established in order to prevent re-actor. thermal limits frort being exceeded.

These thermal limits are either I

power peaking kW/ft limits or DNBR limits. The AXIAL POWER IMBALANCE reduces the power level trip produced by a flux-to-flow ratio sr.n that the boundaries 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 range in which re-actor 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 tne high flux trip setpoint. The trip setpoint for RC high pressure, I

2.~

) psig, has been established to maintcin the system pressure below the safe-ty limit, 2750 psig, for any design transient.

The RC high pressure trip is backed up by the pressurizer code safety valves for RCS over pressure protec-tion knd is therefore set l'ower than the set pressure for these valves, 2435 I

psig. The RC high pressure trip also backs up the high flux trip.

The RC low pressure, 1985 psig, and RC pressure-temperature (12.60 Tout -

I 5660) 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 -- f. flux / 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).

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l DAVIS-BESSE, UNIT 1 B 2-6 g

e-e

I1

' REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION 3.1.1.2 The moderator temperature coefficient (MTC) shall be:

a.

Less positive than 0.9 x 10-" /.k/k/*F whenever TilERMAL POWER is < 95% of RATED THERMAL POWER, b.

Less peaitive than 0.0 x 10~" t.k/k/*F whenever THERMAL POWER is 2 95% of E

RATED THERMAL POWER, and g

c.

Equal to or less negative than -3.0 x 10" Lk/k/*F at RATED THERMAL POWEL.

APPLICABILITY: MODES 1 and 2*#.

ACTION:

With the moderator temperature coefficient outside any of the above limits, be f.n 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 I

4.1.1.?.1 The MTC shall be determined to be within its limits by confirmatory measurements. MTC measured values shall be extrapolated and/or compensated to permit direct comparison with the above limits.

4.1.1.3.2 The !frC shall be determined at the following frequencies and THERMAL POWER conditions during each fuel cycle:

(

a.

Prior to initial operation above 5% of RATFn THERMAL POWER, af ter each fuel loading, b.

At any THERMAL POWER, within 7 days after reaching a RATED THERMAL POWER equilibrium boron concentration of 300 ppm.

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• With k 2 1.0.

df

1. See Special Test Exception 3.10.2.

DAVIS-BESSE, UNIT 1 3/4 1-4 8-9a l

I REACTIVITY CONTROL SYSTEMS REGULATING ROD INSERTION LIMITS I

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 I

200 !!0 EFPD of operation.

If the axial power shaping rods are complete'y withdrawn at 200 210 EFPD for extension of cycle length, then the regulating rod groups shall be limited in physical insertion as shown on Figures 3.1-2e I

and 3.1-3e for the remainder of the cycle. Ilowever, if the axial power shap-ing rods are not completely withdrawn at 200 !!O EFPD, then the regulating rod groups shall be limited in physical insertion as shown on Figures 3.1-2d I

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.

APPLICABILITY: MODES 1* and 2*#.

ACTION:

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

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 TilERMAL 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.

l

• See Special Test Exceptions 3.10.1 and 3.10.2.

1. With k 2 1.0.

eff l

l l

DAVIS-BESSE, UNIT 1 3/4 1-26 8-10

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

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

200 !!0 EFPD of operation.

If this rod group is completely withdrawn at 200 110 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 pplicable.

I However, if the rod group is not completely withdrawn at 200 !10 EFPD, the group shall be limited in physical insertion as shown on Figures 3.1-5d and

-Si for the remainder of the cycle.

I 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 THERE\\L 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 />.

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 2 1.0.

gf DAVIS-BESSE, UNIT 1 3/4 1-34 8-11

I 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, -lb and -Ic and 3.2-2a, -2b and -2c for the first 200 210 EFPD of operation.

If the axial power shaping rods are completely withdrawn at 200 !10 EFPD for extension of cycle len,th, then the AXIAL POWER IMBALANCE e

shall be maintained within the limits shown on Figures 3.2-le and 3.2-2e for the remainder of the cycle. However, if the axial power shaping rods are not g

completely withdrawn at 200 !!0 EFPD, then the AXIAL POWER IMBALANCE shall be g

maintained within the limits shown on Figures 3.2-Id 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 F0WER or less.

I l

SURVEILLANCE REQUIRDfENTS 1

l 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.

I

• See Special Test Exception 3.10.1.

DAVIS-BESSE, UNIT 1 3/4 2-1 8-12

I I

POWER DISTRIBUTION LIMITS NUCLEAR HEAT FLUX HOT CHANNEL FACTOR - F LIMITING CONDITION FOR OPERATION 3.2.2 F shall be limited by the following relationships:

q I

,2.93 FQ~

p where P = RATED THERMAL POWER and P < l.0.

I APPLICABILITY: MODE 1 ACTICN:

With F exceeding its limit:

I a.

Reduce THERMAL POWER at least 1% for each 1% FQ exceeds the limit within 15 minutes and similarly reduce the high flux trip setpoint and flux-a flux-flow trip setpoint within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, b.

Demonstrate through incore mapping that F is within its limit within 24 q

hours af ter exceeding the limit or reduce THERMAL POWER to less than 5%

of RATED THERMAL POWER within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

c.

Identify and correct the cause of the out of limit condition prior to in-creasing THERMAL POWER above the reduced limit required by a or b, above; subsequent POWER OPERATION may proceed provided that Fq is demonstrated through incore mapping to be within its limit at a nominal 50% of RATED THERMAL POWER prior to exceeding this THERMAL POWER, at a nominal 75% of RATED THERMAL POWER prior to exceeding this THERMAL POWER and within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after attaining 95% or greater RATED THERMAL POWER.

SURVEILLANCE REQUIREMENTS I

l 4.2.2.1 F shall be determined to be within its limit by using the incore 0

5 detectors td obtain a power distribution map.

8-13 l

1

Table 3.2-2.

Quadrant Power Tilt Limits Steady state Transient Maximum limit limit limit Measurement Independent j

QUADRANT P0k'ER TILT 4.92 11.07 20.0 QUADRANT PO'n'ER TILT as measured by:

1 l

Symmetrical Incore Detector System 3.03 8.53 20.0 l

Power Range Channels 1.96 6.96 20.0 i

l Minimum Incore Detector System 1.90 4.40 20.0 l

I Il iI L

l I-I DAVIS-BESSE, UNIT 1 3/4 2-12 i

Babcock & Wilcox 8-14

I

[UNDER NRC REVIEW}

POWER DISTRIBUTION LIMITS CHANGES PREVIOUSLY PROPOSED BY LETTER I

DNB PARAMETERS Seria!No.

73l ogeyjotl 1

LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB related parameters shall be maintained within the limits shown on Table 3.2-1:

a.

Reactor coolant hot leg temperature.

b.

Reactor coolant pressure.

c.

Reactor coolant flow rate.

I APPLICABILITY: MODE 1.

ACTION:

If parameter a or b above exceeds its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

If parameter c exceeds its limit, either:

1.

Restore the parameter to witN.n its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or 2.

Limit THERMAL POWER at least 2% below RATED THERMAL POWER for each 1%

parameter c is outside its limit for four pump operation within the I

next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or limit THERMAL POWER at least 2% below 75% of RATED THERMAL POWER for each 1% parameter c is outside its limit for 3 pump operation within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.2.5.1 Each of the parameters of Table 3.2-1 shall be verified to be with-in their limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

I 4.2.5.2 The reactor coolant system total flow rate shall be determined to be within its limit by measurement at least once per 18 months.

I DAVIS-BESSE, UNIT 1 3/4 2-13 8-15 UNDER 3RC RD!EW

iI.

Table 3.2-1.

DNB Margin Limits Four RC Three RC g

pumps pumps E

Parameter operating operating RC hot leg temp.,

T, *F c610 1610(*)

H 22062.7 22058.7(*)

RC pressure, psig( }

RC flow rate, gpm(C} 2396,880 2297,340 (a) Applicable to the loop with two RC pumps operating.

Limit not applicable during either a THER-MAL POWER ramp increase in excess of 5% of RATED THERMAL POWER per minute or a THERMAL g

j POWER step increase of greater than 10% of g

RATED THERMAL POWER.

c)These flows include a flow rate uncertainty of 2.5%.

l l

I I

I I

DAVIS-BESSE, UNIT 1 3/4 2-14 8-16

I l

3/4.4.

REACTOR COOLANT SYSTEM I

3/4.4.1.

C00LAhT LOOPS AND COOLANT CIRCULATION I

STARTUP AND POWER OPERATION LIMITING CONDITION FOR OPERATION 3.4.1.1 Both reactor coolant loops and both reactor coolant pumps in each loop shall be in operation.

APPLICABILITY: MODES 1 and 2*.

ACTION:

a.

With one reactor coolant pur not in operation, STARTUP and POWER OPERATION may be initiated and may proceed provided THERMAL POWER is restricted to less than 79.5% of RATED THERMAL POWER and within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> the setpoints l

for the following trips have been reduced to the values specified in Specification 2.2.1 for operation with three reactor coolant pumps operat-ing:

1.

High Flux 2.

Flux-AFlux-Flow SURVEILLANCE REQUIREMENTS o

4.4.1.1 The above required reactor coolant loops shall be verified to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

4.4.1.2 The reactor rrotective instrumentation channels specified in the ap-plicable ACTION _catement above shall be verified to have had their trip set-points changed to the values specified in Specification 2.2.1 for the appli-cable number of reactor coolant pumps operating either:

a.

Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after switching to a different pump combination if the switch is made while operating, or b.

Prior to reactor criticality if the switch is made while shutdown.

I

• See Special Test Exception 3.10.3.

DAVIS-BESSE, UNIT 1 3/4 4-1 8-17 L

I 3/4.10.

SPECIAL TEST EXCEPTIONS CROUP llEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION 3.10.1.

The group height, insertion and power distribution limits of Speci-fications 3.1.3.1, 3.1.3.2, 3.1.3.5, 3.1.3.6, 3.1.3.7, 3.1.3.9, 3.2.1 and 3.2.4 may be suspended during the performance of PHYSICS TESTS provided:

a.

The THERMAL POWER is maintained s 85% of RATED THERMAL P0kTR, b.

The High Flux Trip Setpoint is 5 10% of RATED THERMAL POWER higher than the THERMAL POWER at which the test is performed, with a maximum setting of 90% of RATED THERMAL POWER, and, c.

The limits of Specifications 3.2.2 and 3.2.3 are maintained and determined at the frequencies specified in 4.10.1.2 below.

APPLICABILITY: MODE 1.

ACTION:

With any of the limits of Specifications 3.2.2 or 3.2.3 being exceeded while the requirements of Specifications 3.1.3.1, 3.1.3.2, 3.1.3.5, 3.1.3.6, 3.1.3.7, 3.1.3.9, 3.2.1 or 3.2.4 are suspended, either:

a.

Reduce THERMAL POWER sufficiently to satisfy the ACTION requirements of Specifications 3.2.2 and 3.2.3, or b.

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 4.10.1.1.

The High Flux Trip Setpoint shall be determined to be set within g

the limits specified within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> prior to the initiation of and at least 3

once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> during PHYSICS TESTS.

4.10.1.2.

The Surveillance Requirements of Specifications 4.2.2 and 4.2.3 shall be performed at least once per two hours during PHYSICS TESTS.

I DAVIS-BESSE, UNIT 1 3/4 10-1 8-18

SPECIAL TEST EXCEPTIONS PHYSICS TEST I

LIMITING CONDITION FOR OPERATION 3.10.2.

The limitations of Specifications 3.1.1.3, 3.1.3.1, 3.1.3.2, 3.1.3.5, 3.1.3.6, 3.1.3.7, and 3.1.3.9 may be suspended during the performance of PHYSICS TESTS provided:

a.

The THERMAL P0kTR does not exceed 5% of RATED THERMAL POWER, and b.

The reactor trip setpoints on the OPERABLE High Flux Channels are set at 5 25% of RATED THERMAL POWER.

c.

The nuclear instrumentation Source Range and Intermediate Range high startup rate control rod withdrawal inhibit are OPERABLE.

APPLICABILITY: MODE 2.

ACTION:

With the THERMAL POWER > 5% of RATED THERMAL POWER, immediately open the con-I trol rod drive trip breakers.

SURVEILLANCE REQUIREMENTS 4.10.2.1.

The THERMAL POWER shall be determined to be s 5% of RATED THERMAL POWER at least once per hour during PHYSICS TESTS.

4.10.2.2.

Each Source and Intermediate Range and High Flux Channel shall be subjected to a CHANNEL FUNCTIONAL TEST within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> prior to initiating PHYSICS TESTS.

I DAVIS-BESSE UNIT 1 3/4 10-2 8-19

REACTIVITY CONTROL SYSTEMS BASES 3 /4.1. 4.

MINDfUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical with the reactor coolant system average temperature less than 525"F.

This limitation is required to ensure (1) the moderator temperature coefficient is within its analyzed temperature range, (2) the protective instrumentation is within its normal operating range, (3) the pressurizer is capable of being in an OPERABLE status with a steam bubble, and (4) the reactor pressure vessel is above its minimum RTNDT temperature.

3/4.1.2.

BORATION SYSTEMS l

The boron injection system ensures that negative reactivity control is avail-able during each mode of facility operation.

The components required to per-form this function include (1) borated water sources, (2) makeup or DHR pumps, (3) separate flow paths, (4) boric acid pumps, (5) associated heat tracing systems, and (6) an emergency power supply from operable emergency busses.

With the RCS average temperature abuve 200*F, a minimum of two separate and g

redundant boron injection systems are provided to ensure single functional g

capability in the event an assumed failure renders one of the systems inop-erable. Allowable out-of-service periods ensure that minor component repair or corrective action may be completed without undue risk to overall facility safety from injection system failures during the repair period.

The boration capability of either system is sufficient to provide a SHUTDOWN g

MARGIN from all operating conditions of 1.0% Ak/k after xenon decay and cool-3 down to 200 F.

The maximum boration capability requirement occurs from full power equilibrium xenon conditions and requires the equivalent of either 7373 gallons of 8742 ppm borated water from the boric acid storage tanks or 52,726 gallons of 1800 ppm borated water from the borated water storage tank.

The requirements for a minimum contained volume of 434,650 gallons of borated l

water in the borated water storage tank ensures the capability for borating 5

the RCS to the desired level. The specified quantity of borated water is con-sistent with the ECCS requirements of Specification 3.5.4, therefore, the larger volume of borated water is specified.

With the RCS temperature below 200*F, one injection system is acceptable with-out single failure consideration on the basis of the I

DAVIS-BESSE, UNIT 1 B 3/4 1-2 8-20

REACTIVITY CONTROL SYSTEMS I

BASES 3/4.1,3.

MOVABLE CONTROL ASSEMBLIES (Continued)

The maximum rod drop time permitted is consistent with the assumed rod drop time used in the safety analyses. Measurement with Tavg 2 525*F and with re-actor coolant pumps operating ensures that the measured drop times will be representative of insertion times experienced during a reactor trip at operat-ing conditions.

I Cntrol rod positions and OPERABILITY of the rod position indicators are re-quired to be verified on a nominal basis of once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> with frequent verifications required if an automatic monitoring channel is inoperable. These I

verification frequencies are adequate for assuring that the applicable LCO's are satisfied.

Technical Specification 3.1.3.8 provides the ability to prevent excessive I

power peaking by transient xenon at RATED THERMAL POWER. Operating restric-tions resulting from transient xenon power peaking, including xenon-free startup, are inherently included in the limits of Sections 3.1.3.6 (Regulat-I ing Rod Insertion Limits), 3.1.3.9 (Axial Power Shaping Rod Insertion Limits),

and 3.2.1 (Axial Power Imbalance) for transient peaking behavior bounded by the following factors.

For the period of cycle operation where regulating rod groups 6 and 7 are allowed to be inserted at RATED THERMAL POWER, an 8%

I peaking increase is applied at or above 92% FP.

An 18% increase is applied below 92% FP.

For operation where only regulating rod group 7 is allowed to be inserted at RATED THERMAL POWER, a 5% peaking increase is applied at or above 92% FP and a 13% FP increase is applied below 92% FP.

If these values, checked every cycle, conservatively bound the peaking effects I

of all transient xenon, then the need for any hold at a power level cutoff be-low RATED THERMAL POWER is precluded. If not, either the power level at which the requirements of Section 3.1.3.8 must be satisfied or the above-listed fac-tors will be suitably adjusted to preserve the LOCA linear heat rate limits.

The limitation on axial power shaping rod insertion is necessary to ensure that power peaking limits are not exceeded.

I DAVIS-BESSE, UNIT 1 B 3/4 1-4 8-21

I 3/4.2.

POWER DISTRIBUTION LIMITS BASES he specifications of this section provide assurance of fuel integrity during Condition I (normal operation) and II (incidents of moderate f requency) events T

1.30 during normal opera-by: (a) maintaining the minimum DNBR in the core 2 l

term transients, (b) maintaining the peak linear power 5

tion and during shortduring normal operation, and (c) maintaining the peak density : 18.4 kW/ft power density less than the limits given in the bases to specification 2.1 in In addition, the above criteria must be met term transients.

during short accidents.

order to meet the assumptions used for the loss-of-coolant The power imbalance envelope defined in Figures 3.2-1 and 3.2-2 and the curves, Figures 3.1-2 and 3.1-3 are based on LOCA analyses insertion limit rate such that the maximum clad which have defined the maximum linear heat temperature will not exceed the Final Acceptance Criteria of 2200*F following l

Operation outside of the power imbalance envelope alone does not con-E a LOCA.

stitute a situation that would cause the Final Acceptance Criteria to be ex-The power imbalance envelope represents the bound-ceeded should a LOCA occur.

g ary of operation limited by the Final Acceptance Criteria only if the control rods are at the insertion limits, as defined by Figures 3.1-2 and 3.1-3 and if g

Additional conservatism is the steady-state If.mit QUADRANT POWER TILT exists.

introduced by application of:

I Nuclear uncertainty factors.

a.

b.

Thermal calibration uncertainty.

Fuel densification effects.

c.

d.

Hot rod manufacturing tolerance factors.

Potential fuel rod bow effects.

c.

E limited variations from the basic require-Thr ACTION statements which permit the orig-l meets are accompanied by additional restrictiens which ensures that inai criteria are met.

The definitions of the design limit nuclear power peaking factors as used in these specifications are as follows:

Nuclear heat flux hot channel factor, is defined as the maximum local fuel l

rod linear power density divided by the average fuel rod linear power den-l F0 sity, assuming nominal fuel pellet and rod dimensions.

I B 3/4 2-1 I

DAVIS-BESSE, UNIT 1 8-22 I

I POWER DISTRIBUTION LIMITS I

BASES Themeasurementofenthalpyrisehotchannelfactor,FfH,shallbein-b.

creased by 5 percent to account for measurement error.

For Condition II events, the core is protected from exceeding the values given in the bases to specification 2.1 locally, and from going below a minimum DNBR of 1.30, by automatic protection on power, AXIAL POWER IMBALANCE pressure and temperature. Only conditions 1 through 3, above, are mandatory since the AXIAL POWER IMBALANCE is an explicit input to the reactor protection system.

I The QUADRANT POWER TILT limit assures that the radial power distribution sat-isfies the design values used in the power capability analysis. Radial power I

distribution measurements are made during startup testing and periodically dur-ing power operation.

The QUADRANT POWER TILT limit at which corrective action is required provides I

DNB and linear heat generation rate protection with x-y plane power tilts.

In the event the tilt is not corrected, the margin for uncertainty on Fq is rein-stated by reducing the power by 2 percent for each percent of tilt in excess of the limit.

3/4.2.5.

DNB PARAMETERS The limits on the DNB related parameters assure that each of the parameters are maintained within the normal steady state envelope of operation assumed in the transient and accident analyses.

The limits are consistent with the FSAR I

initial assumptions and have been analytically demonstrated adequate to main-tain a minimum DNBR of 1.30 throughout each analyzed transient.

I The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodic surveillance of these parameters through instrument read-out is sufficient to ensure that the parameters are restored within their lim-its following load changes and other expected transient operation. The 18 month periodic measurement of the RCS total flow rate using delta P instrumen-I tation is adequate to detect flow degradation and ensure correlation of the flow indication channels with measured flow such that the indicated percent flow will provide sufficient verification of flow rate on a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> basis.

I DAVIS-BESSE, UNIT 1 B 3/4 2-3 8-23

3/4.4.

REACTOR COOLANT SYSTEM BASES e

3/4.4.1.

REACTOR COOLANT LOOPS The plant is designed to operate with both reactor coolant loops in operation, and maintain DNBR above 1.30 during all normal operations and anticipated transients. With one reactor coolant pump not in operation in one loop, g

THERMAL POWER is restricted by the Nuclear Overpower based on RCS flow and E

AX1AL POWER IMBALANCE, ensuring that the DNBR will be maintained above 1.30 at the maximum possible THERMAL POWER for the number of reactor coolant pumps in operation or the local quality at the point of minimum DNBR equal to 22%,

whichever is more restrictive.

In MODES 3, 4 and 5, a single reactor coolant loop or DHR loop provides suf-g ficient heat removal capability for removing decay heat; but single failure l

considerations require that at least two loops be OPERABIE. Thus, if the re-actor coolant loops are not OPERABLE, this specification requires two DRR loops to be OPERABLE.

Natural circulation flow or the operation of one DHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity l

changes during boron concentration reductions in the reactor coolant system.

5 The reactivity change rate associated with boron reduction will, therefore, be within the capacity of operator recognition and control.

3/4.4.2 and 3/4.4.3 EAFETY VALVES The pressurizer code safety valves operate to prevent the RCS from being pres-l surized above its safety limit of 2750 psig.

Each safety valve is designed to a

relieve 336,000 lbs per hour of saturated steam at the valve's setpoint.

The relief capacity of a single safety valve is adequate to relieve any over-pressure condition which could occur during shutdown.

In the event that no safety valves are OPERABLE, an operating DHR loop, connected to the RCS, pro-vides overpressure relief capability and will prevent RCS overpressurization.

During operation, all pressurizer code safety valves must be OPERABLE to pre-vent the RCS from being pressurized above its safety limit of 2750 psig. The g

combined relief capacity of all of these valves is greater than the maximum g

surge rate resulting from any transient.

Demonstration of the safety valves' lift settings will occur only during shut-down and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Code.

I DAVIS-BESSE, UNIT 1 B 3/4 4-1 6-24

I Figure 8-1.

Reactor Core Safety Limit (Tech Spec Figure 2.1-1) 2600 2400 RC HIGH PRESSURE TRIP 618,2300 RC HIGH TEMPERATURE

=

S 2200 TRIP I

ACCEPTABLE 618,2127 O

OPERATION RC PRESSURE TEMPERATURE TRIP

/!

2000 606.7. 1985 SAFETY LIMIT RC LOW PRESSURE TRIP 1800 E

I I

I f

f I

590 600 610 620 630 640 Reactor Outlet Temperature, F I

I I

I I

I

~"

~"

I I

Figure 8-2.

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

% RATED TMERMAL power dl

-- 120

(-30.0.182)

.. 100

(-50.0. 9 4.0 )

(22.4.89.3)

(-30.89.3 3 PUMP LlulT I

80

(-50.0.71.3) (

- 60 ACCEPTABLE OPERATton FOR SPECIFIED RC PUMP COM8inATION

-- 40 UNACCEPTA8LE UNACCEPTABLE OPERATION 0

OPERATION l

'd s

t i

1 I

f

-60

-40

-20 0

20 40 60 A11AL POWER. %

I PUMPS OPERATING REACTOR COOLANT FLOW. (GPM) 4 387.200 3

290.l00 I

8-26 Babcock & Wilcox

I I

Figure 8-3.

Trip Setpoint for Flux-A Flux / Flow i

(Tech Spec Figure 2.2-1) e

% RATED THERMAL POWER I

Curve shows trip setpoint for a 25%

flow reduction for three-pump operation (290, 100 gpm). The actual setpoint will be directly proportional to the

- 120 I

actual flow with three pumps.

8.O,106.4 I

-13.5.106.4 4 PUMP

-100 M,

1.282 LIMIT M : -l.656

=

I LINES 2

I I

I I

I

-33.0.81.4

- 80

-13.5.79.5 l

8.0,79.5 I

l LIMIT l

3 PUMP l LINES l

33.0,65.0

- 60

-33.0.54.5

(

UNACCEPTABLE ACCEPTABLE OPERATION FOR OPERATION SPECIFIED RC PUMP COMBINATION I

- 40 l

) 33.0,38.1 UNACCEPTABLE l

OPERAT10N m 1

- 201 o 2

I l

l=

Il 11 11 2

m" Il m

l I

e I

e i

I l

t I

I i

' -60

-40

-20 0

20 40 60 I

AXIAL POWER IMBALANCE, %

I l

I l

8-27 Babcock & Wilcox

I Figure 6-3a.

Allowable Value for Flux-A Flux / Flow

% RATED THERMAL POWER I

DELETED I

I AXIAL POWER IMBALANCE, %

(Tech Spec Figure 2.2-2)

I I

I I

I I

I I

I I

8-27a Babcock & Wilcox

l i

4 Figure 8-4.

Pressure / Temperature Limits at Maximum Allowable Power for Minimum DNBR (Tech Spec Figure 2.1) i l 2400 3 PUMP CURVE 2300

.I 3

4 PUMP CURVE 2200 1

C.

l 2100 0;

iE 5

'5 2000 g

I 1900 1800 1700 t

l t

t f

f 1

580 590 600 610 620 630 640 Reactor Outlet Temperature, (*F)

I PUMPS FLOW (GPM)

POWER 4

387,200

112r, 3

290.100 89,3'i I

I 8-28 Babcock & Wilcox

W m

W W

W M

M M

M M

M M

M M

M M

M W

W Figure 8-5.

Regulating Group Position L'.mits 0 to 60 EFPD, Four RC Pumps - Davis Besse 1. Cycle 3 (Tech Spec Figure 3.1-2a)

OPERAT10N RESIRlCTED 1

(290,102)

(223,102)

(271,102)c

'l POWER LEVEL 100 CUT 0FF = 1005 (211.92)

(300,92) 5 80 SHUIDOIN (261,80) 2 MARGIN Lluli E

UNACCEPTABLE 60 OPERAil0N g

OPERAil0N RESTRICIED cn

{

(168,50)

(225.5,50) o f

40 2

ACCEP T ABL E 5

OPERAT10N E

20 (96,15)

(0.2.5) 03 0

8 E

O 100 200 300 Q

Rod index ($ Withdrawn) o GR 5 0

15 100 GR 6.

0 25 75 100 g

ca 7 'O 15 ido x

Figure 8-6.

Regulating Group Position Limits, 50 to 150 2 10 EFPD, Four RC Pumps - Davis Besse 1, Cycle 3 (Tech Spec Figure 3.1-2b)

(271.102)

(300,102)

(247.3,102) c O P0BER LEVEL 100 CUTOFF = 1005 (271,92)

SHUT 00tN MARGIN LIMIT y

80 p

(252,80) o m

a 5

60 8

[

(188,50)

(225.5.50) o O

cs:

f 40 ACCEPTABLE h

UNACCEPTABLE OPERATION 2

0PERAll0N t

20

[

(113,15)

(0,2.5' 0

m S

0 100 200 300 g

Roo inoet (% Estnarann) o GR 5 Q.

0 75 100

'" 8

'O 15 15 ib0 f

GR 7 i i

i g

0 25 100 x

g g

W M

E

\\

7 Figure 8-7.

Regulating Group Position Limits, 150 ! 10 to 200 ! 10 EFPD, Four RC Pumps - Davis Besse 1, Cycle 3 (Tech Spec Figure 3.1-2c)

(271,102) (300,102)

(251.2.102)

O POWER LEVEL 100 CU10FF = 100%

(261.92) 2 80 SHU1009N g

(252,80)

MARGIN LIMil 4/

i E

b 60 C

h g%

a

?

(200,50)

(225.5.50)

UNACCEPTABLE

~

OPERAi!GN b

ACCEPTABLE Ot OPERA 110N

-f 20 (115.15)

(0,2.5) 0 i

U3 0

100 Roa innen (% Witnarann) 200 300 GR 5,

o j$

l'00

n-GR 6 i e

0 25 15 100 GR 7 i

8-j5 100 X

Figure 8-8.

Regulating Group Position Limits, 200 10 to 230 t 10 EFPD, Four RC Pumps - Davis Besse 1. Cycle 3 (Tech Spec Figure 3.1-2d)

(300,102)

(257.2,102)

{271.102)O POIER LEVEL 100 I

CUT 0FF = 1005

' (267.92)

SHU1001N 80 MARGIN LIMIT (252,80)

^

5 E

g y

60 c

m S

g 1

O UNACCEPTABLE OPERAil0N (200,50) 4 (225.5.50) to a:

o 40 5

t; ACCEPTABLE t

OPERATION 20 (115,15)

(0,2.5)

W D

I b

0 100 200 300 g

Ron Index (5 Withdrann) d

[5 100 g

'O i5 ido 25 s-cR 7 S

0 25 100 aus e

sus mai set sum aus sum m

mun amm ama sus sum gas amm num an uns

m m

m m

m M

M M

M M

M M

m eeW W

m Figure 8-9.

Regulating Group Position Limits, 200 10 to 268 10 EFPD, Four RC Pumps APSRs Withdrawn - Davis Besse 1. Cycle 3 (Tech Spec Figure 3.1-2e)

(250.102) (274.102) (300.102) 100 POWER LEVEL CUiOFF = 1005 (265.92) n 80 (255.80)

EHUIDutN NARGIN LIMli a

C y

60 UNACCEPTABLE OPERAil0N 0

g (200.50)

(225.50)

Y

~

40 ACCEPTABLE OPERAil0N 5

ta f

20 (118.15)

(0.2.5) 0 0

100 Roa inaen (5 Witnarawn)200 300 En y

GR 5 i

i i

n 0

75 100 t

O GR 6 i i

i

=

{

E-0 25 75 100 i

e GR 7 8 j

g 0

25 100 i

E O

d Figur_ 3-10.

Regulating Group Position Limits, O to 60 EFPD, Three RC Pumps - Davis Besse 1 Cycle 3 1

(Tech Spec Figure 3.1-3a) 100 5

80 (223.77)

(271.77) c 1290.77)

E

(

l.69.5)

(300,69.5) 60 SHUT 000N MARGIN LIMli 3

OPERAll0N m

RESTRICIED 6

e 5

40 UNACCEPTABLE (168.38)

(225.5.38)

{

t OPERAil0N O

ACCEPTABLE E

20 OPERAil0N (96.11.75)

(0.2.5 i

i 0

100 300 Roo innen ($ Eitnaraen) a1 GR 5.

0 75 100 n

GR 6 0

25 15 th0 x-GR 7.

[

P O

25 100 l

E l

W R

i

Figure 8-11.

Regulating Group Position Limits, 50 to 150 10 EFPD, Three RC Pumps - Davis Besse 1, Cycle 3 (Tech Spec Figure 3.1-3b) 100 S

80 (271,77) g (247.3,77) 3 (300,77) a_

y (271.69.5) a:

E

@q (252,60.5)

[

60 SHUIDOWN UNACCEPTABLE MARGIN LIMIT 4

OPERATION v

y 40 (188,38)

(225.5,38)

S t

5 20 ACCEPTABLE OPERATION (113.11.75)

(0,2.

0 100 Rod index (5 Witnaraen) 200 300 to

'O h5 l'00 7

o GR 6 E-0 25 75 100 P

CR 7 g

0 25 100 g

Figure 8-12.

Regulating Group Position T4aits, 150 ! 10 to 200 10 EFPD, Three RC Pumps - Davis Besse 1, Cycle 3 (Tech Spec Figure 3.1-3c) 100

^

a:

(271,77)

,E 80 (257.2.77) 3 (300,77) a-d E

(267,69.5)

E S

60 8 (252.60.5)

O SHUT 00RN

[

UNACCEPTABLE MARGIN LIMil OPERAll0N 6

4o (200,38)

(225.5.38)

{

5 26 ACCEPTABLE OPERAil0N (115.11.75)

(0,2 u; ~

un i

i G

c+.

0 100 Roa index (5 Withdrawn)200 300 GR 5 g

0 75 100 l

GR 6 i

e i

i o

Q-0 25 75 100 l

GR 7 i

i i

p 0

25 100 l

g W

O t

Figure 8-13.

Regulating Group Position Limits, 200 ! 10 to 230 ! 10 EFPD, Three RC Pumps - Davis Besse 1, Cycle 3 (Tech Spec Figure 3.1-3d) 100 (271,77)

(300,77)

S 80 (257.2,77) o (267,69.5)

N 5

(252*60'5) 60 SHUT 00$N 5

y g

MARGIN LIMIT s

?

O b

40 UNACCEPTABLE (200,38 (225.5.38)

S OPERAll0N 5

ACCEPTABLE 3

OPERAll0N E

20 (115.11.75) i e

a 0

300 Rod index (5 Witndrawn) f E

GR 5 i i

i 0

75 100 o

E-GR 6 i

i i

i 0

25 15 100 e

GR 1 i i

i Ei 0

25 100 g

0

I l

l Figure 8-14.

Regulating Group Position Limits, 200 ! 10 to 268 1 10 EFPD Three RC Pumps, APSRs Withdrawn - Davis Besse 1. Cycle 3 (Tech Spec Figure 3.1-3e) i 100 (274,77)

(300,77) 80 (250,77) o g

a_

a 4

(265,69.5) g w5 60 SHUTOOWN (255.60.5) o MARGIN LIMli s

p[

r 4

40 UNACCEPTABLE (200.38)

(225,38) e 5

2 OPERAil0N ACCEPTABLE

[

20 OPERAll0N 118.11.75)

(0.2.5)

I I

e u

100 200 300 Roa Index ($ 3etnarann) g GR 5 b

75 100 g

o GR 6 I i

i i

0 25 15 100 l

Q.

GR 7 g.

0 23 400

$w M

M WM BEEE m

e m

ser mas aus sus sus seu aus aus aus a

sus

I

~

I Figure 8-15.

APSR Position Limits, O to 60 EFPD, Four RC Pumps - Davis Besse 1. Cycle 3 (Tech Spec Figure 3.1-Sa)

I (8,102)

(38,102) 100 (6,92)

(40,92)

I

^=W RESTRICTED E

80 (4,80)

(42,80)

REGION I

!g

< (0.70)

O 60 m

I

=

g PERMISSIBLE (100,50)

E OPERATING REGION I

40

=

E E

a.

20 I

o I

O i

i i

i I

O 20 40 60 80 100 APSR Position (% Witnarawn)

I I

I I

8-39 Babcock & Wilcox

I Figure 8-16.

APSR Position Limits, 50 to 150 10 EFPD, Four RC Pumps - Davis Besse 1 Cycle 3 (Tech Spec Figure 3.1-5b)

(8.102)

(38.102)

(6.92)

(40,92)

RESTRICTED i

REGION x

(4.80)

(45.80) 5 80 a.

i g

"m

(

(0,70) 5 60

=

PERMISSIBLE OPERATING REGION (100,50)

E Z

g 5

40 0

E a.

20 I

E 0

D 20 40 60 80 100 g

l APSR Position (% Withdrawn)

I I

I I

I l

8--40 Babcock & Wilcox I

~

i

B Figure 8-17.

APSR Position Limits, 150 ! 10 to 200 ! 10 EFPD, Four RC Pumps - Davis Besse 1 Cycle 3 (Tech Spec Figure 3.1-Sc) i (6,102)

(38,102) e n

100 (6,92)

(40,92)

RESTRICTE0 80 (4,80)

(45,80) g w

I e

('

)

(100,70) l d

E l

[x 60 PERMISSIBLE

=

OPERATING REGION j

b O

i 40 i

8 I

20 lI I

0 0

20 40 60 80 100 APSR Position (% Witndrawn)

I I

I I

8-41 Babcock & Wilcox

~

l Figure 8-18.

APSR Position I.imits, 200 2 10 to 230 2 10 EFPD, Four RC Pumps - Davis Besse 1, Cycle 3 (Tech Spec Figure 3.1-5d)

(G,102)

(38,102) 100 C

l (40,92) eq(6.92) 80 (4,80)

(45,80)

RESTRICTE0 REGION cx:

(0.70) 60 E

W 5

(100,50)

PERMISSIBLE E

OPERATING REGION l

(

40 E

=

e E

l t

20 h

3 0

0 20 40 60 80 100 APSR Position (t Withdrawn)

I I

I I

8-42 Babcock & Wilcox

i I

Figure 8-19.

APSR Position Limits, 200 2 10 to 268 t 10 EFPD,

,l Three or Four RC Pumps, APSRs Withdrawn -

W Davis Besse 1, Cycle 3 (Tech Spec Figure 3.1-Se)

I 110 100 90 I

80

^m

1. 2 0

l f

APSR INSERTION NOT ALLOWE0 5

IN THIS TIME INTERVAL 5

60 I

E 50 3

40 30 I

3 20 I

10 0

e i

i i

i 0

10 20 30 40 50 60 70 80 90 100 APSR Position (% Withdrawn)

'I I

!I l

8-43 Babcock & Wilcox

I I

Figure 8-20.

APSR Position Limits, O to 60 EFPD, Three RC Pumps - Davis Besse 1, Cycle 3 (Tech Spec Figure 3.1-5f)

I 100 I

g 80

- (8,77)

(38,77)

E (6,69.5)

(40,69.5) a 3

b 3

y 60 (4,60.5)

(42,60.5)

RESTRICTE0

(

(0,53)

E REGION l

40 c.

PERMISSIBLE (100,38) a OPERATING REGION

'o E

a 20 I

0 e

i i

i i

0 20 40 60 80 100 APSR Position (fs Witndrawn)

I I

I I

I 8-44 Babcock & Wilcox

5 Figure 8-21.

APSR Position Limits, 50 to 150 10 EFPD, Three RC Pumps - Davis Besse 1, Cycle 3 (Tech Spec Figure 3.1.'g) 100 I

e 80 - (8 77)

(38,77)

I Y

E g

(6,69.5)

(40,69.5)

I E

(45,60.5)

RESTRICTE0 60 (4,60.5)

REGION I

o i (0.53) a:

I O

g 40 e

I

{

(100,38) a.

PERMISSIBLE f

20 OPERATING REGION l

I 0

i i

i i

0 20 40 60 80 100 APSR Position ($ Withdrawn)

. I I

I I

8-45 Babcock & Wilcox

I Figure 8-22.

APSR Position Limits, 150 ! 10 to 200 10 EFPD, Three RC Pumps - Davis Besse 1. Cycle 3 (Tech Spec Figure 3.1-5h) 100 I

5 80 - ( 6. 77)

(38,77) l

=

I c

a I

( 6,69. 5)

(40,69.5)

E RESTRICTE0 (45.60.5)

REGION E

60 - (4,60.5) oW l

5

(

(0,53)

(100,53) o E

g 40 E

?

I g

PERMISSIBLE 20 OPERATING REGION I

O I

I I

0 20 40 60 80 100 l

APSR Position (% Witndrawn)

I l

l I

1 I

I I

8-46 Babcock & Wilcox

~. -..

I I

Figure 8-23.

APSR Position Limits, 200 2 10 to 230 2 10 EFPD, Three RC Pumps - Davis Besse 1, Cycle 3 (Tech Spec Figure 3.1-51) l 100 I

h 80 _ (6,77)

(3g,77) rI e

e 3

(6.69.5)

(40,69.5)

I (45,60.5) 60 (4,60.5) a

( (0,53)

REGION I

=

g 40 (100,38) g PERMISSIBLE f

OPERATING REGION I

I o

0 20 40 60 80 100 APSR Position (S Tithdrawn)

I I

I I

8-47 Babcock & Wilcox

I Figure 8-24.

Axial Power Imbalance Limits, O to 60 EFPD, 1

Four RC Pumps - Davis Besse 1. Cycle 3 (Tech Spec Figure 3.2-la)

( - 17,10 2) r, 100

( - 17,9 2 )

( + 13,92 )

I

=

(-25,80) j 80

( + 11,80 )

a.

i I

E RESTRICTED 60 o

REGION 5

=

( 30,50)

PERMISSIBLE

( +2 5,50 )

OPERATING g

g i

REGION g

40 g

5 20 E

I I

I I

I I

40

-30

-20

-10 0

10 20 30 Ansal Poner imcalance (%)

I I

I I

I I

I 8-48 Babcock & Wilcox

I l

I Figure 8-25.

Axial Power Imbalance Limits, 50 to 150 10 EFPD, Four RC Pumps - Davis Besse 1, Cycle 3 (Tech Spec Figure 3.2-lb)

(+12.102)

( 20,102)

(-23.92)

(+14,92) 2

( -3 2. 80 )

$-- 80

( + 11,80)

RESTRICTED y

REGION I

S

- - 60 PERMISSIBLE OPERATING REGION

(+25.50)

(-35,50)

I

=5- - 40 b

I J'

-- 20 I

I I

I I

I I

I 40

-30

-20

-10 0

+10

+20

+30 Axial Poner Imnalance (S)

I I

I I

I I

l 8-49 Babcock & Wilcox

l I

l Figure 8-26.

Axial Power Imbalance Limits, 150 2 10 to 200 10 EFPD, Four RC Pumps - Davis Besse 1, Cycle 3 (Tech Spec Figure 3.2-Ic) l

( 25.102)

- - 100

( +15,102 )

I

( - 26. 5,92 )

(+18,92) l

(-33,80)'

f- - 80

(+21,80) l 2

i I

[-- 60 PERill SSIBL E g

(-35,50)h, OPERATING REGION Y

3 - - 40 I

j RESTRICTED REGION

~

-- 20 I

l l

l l

l l

l

-40 30

-20

-10 0

10 20 30 Axial Poner imoalance (%)

I I

I I

t I

I I

8-50 Babcock & Wilcox I

I I

Figure 8-27.

Axial Power Imbalance Limits, 200 1 10 to 230 2 10 EFPD, Four RC Pumps - Davis Besse 1. Cycle 3 I

(Tech Spec Figure 3.2-Id)

I

( 25.102) -

- (+15,102)

( 26.5,92)

(+18,92) h 80

( +21,80 )

(-33,80)

I e

m

- 60

(-35,50) h, PERMISSIBLE

(+25,50)

OPERATING REGION

$-- 40 RESTRICTED j

REGION I

3 I

- 20 l

l l

l l

l l

l 40

-30

-20

-10 0

10 20 30 Asial Poner imoalance (".)

I I

I I

I I

8-51 Babcock & Wilcox

I Figure 8-28.

Axial Power Imbalance Limits, 200 ! 10 to 268 10 EFPD, Four RC Pumps APSRs Withdrawn - Davis Besse 1. Cycle 3 (Tech Spec Figure 3.2-le)

(-28,102)

(15.102) ing

(-30.92)

(18,92)

(-35.80)

- 80 (21,80)

I 60 o

( 50,50)

(25.50)

PERMISSIBLE E

OPERATING REGION W

RESTRICTED REGION l

l l

50

-40

-30 20

- li) 0 10 20 3J Anial Poner imoalance (%)

I I

I I

I I

I 8-52 Babcock & Wilcox I

I I

Figure 8-29.

Axial Power Imbalance Limits, O to 60 EFPD, Three RC Pumps - Davis Besse 1. Cycle 3 (Tech Spec Figure 3.2-2a)

-- 100 I

- - 80

(-12.75.77) C (7.5,77)

(-12.75.69.5)

(9.75,69.5) e (12.75,60.5)

(-18.75.60.5) g- - 60 I

55 S

I

- - 40 E

(-22.5.38.

(18.75.38)

RES7RICTED PERWISSIBLE o

REGION OPERATING D -- 20 REGION t

E a.

I I

I I

I I

l

-30

-20 10 0

+10

+20

+30 Axial Power imoalance (f.)

I I

I I

I I

8-53 Babcock & Wilcox

I Figure 8-30.

Axial Power Imbalance Limits, 50 to 150 10 EFFD.

Three RC Pumps - Davis Besse 1. Cycle 3 (Tech Spec Figure 3.2-2b)

, 100 I

( 15.77

(+9,77) 1

(-17.25,69.5)

(+10.5,69.5)

I W

t a

l

( 24,60.5) 60

(+12.75.60.5) a-

=

=W RESTRICTED

[

REGION b

- 40

( 26 25.38)

(+18.75,38)

=

PERWISSIBLE OPERATING g

REGION E

- 20

=

o l

l l

l l

-30

-20

-10 0

+10

+20

+30 l

Axial Poner imoalance (5)

I I

l I

I I

I I

8-54 Babcock & Wilcox I

I I

Figure 8-31.

Axial Power Imbalance Limits, 150 10 to 200 10 EFPD, Three RC Pumps - Davis Besse 1, Cycle 3 (Tech Spec Figure 3.2-2c) l

- 100

(.18.75.77)

~

~

(+11.25.77)

( -19. 8,69. 5)

(+13.5.69.5)

I 5

(-24.75,60.5)

E-60

(+15.75,60.5)

PERillSSIBLE

[

( 26.25.38) h, OPERATING g

(+18.75,38)

- 40 I

REGION

[

RESTRICIED REGION E

I

{-- 20 t

I E

I I

I I

I I

-30

-20 10 0

+10

+20

+30 Axial Power imoalance ($)

I I

I I

I I

I 8-55 Babcock & Wilcox

I Figure 8-32.

Axial Power Imbalance Limits, 200 10 to 230 ! 10 EFPD, Three RC Pumps - Davis Besse 1, Cycle 3 (Tech Spec Figure 3.2-2d)

. 100 I

(-18.75,77) 80 E

(+11.25,77) 5

(-19.8,69.5)

(+13.5,69.5)

^

W E

(+15.75,60.5)

(-24.75.60.5)

-- 60 d

E E

RESTRICTED

=

REGION O

- - 40

(-26.25.30) );

PERMISS I BL E

(+18.75,38) g

=

OPERATING W

REGION h

-- 20 t

E I

I I

I I

I I

l

-30 20

-10 0

+10

+20

+30 Axial Poser imoalance (%)

I I

I I

I I

l I

6-56 Babcock & Wilcox I

I I

Figure 8-33.

Axial Power Imbalance Limits, 200 ! 10 to 268 1 10 EFPD, Three RC Pumps, APSRs Withdrawn - Davis Besse 1, Cycle 3 (Tech Spec Figure 3.2-2e)

- 100 I

(-21,77)

(11.25,77)

( 22.5,69.5)

(13.5,69.5)

(-26.25.60.5) d' -- 60 (15.75.60.5)

I 5

E

(-37.5.38)

- - 40 (18.75,38)

PERillS$1GLE

=

OPERATING REGION RESTRICTED j - - 20 REGION I

a.

I I

I I

I I

I

-40

-30

-20

-10 0

10 20 30 Axial Poner imoalance (t) i I

I I

I I

I 8-57 Babcock & Wilcox

I Figure 8-34.

Control Rod Core Locations and Group Assignments - Davis Besse 1, Cycle 3 (Tech Spec Figure 3.1-4) 1 l

A B

4 7

4 C

1 6

6 1

0 7

8 5

8 7

E 1

5 5

1 F

4 8

3 7

3 8

4 G

6 2

2 6

-I H -

7 5

7 3

7 5

7 K

6 2

2 6

4 8

3 7

3 8

4 1

5 5

1 N

7 8

5 8

7 0

1 6

6 1

P 4

7 4

I i

i Z

l 1

2 3

4 5

6 7

8 9

10 11 12 13 14 15 GROUP NO. OF RODS FUNCTIONS 1

8 SAFETY X

GROUP NUMBER 2

4 SAFE Y p

4 8

SAFETY 5

8 CONTROL 6

8 CONTROL 7

12 CONTROL 8

8 APSRs TOTAL # E 8-58 Babcock & Wilcox

I I

I 9.

STARTUP PROGRAM - PHYSICS TESTING The planned startup test program associated with core performance is outlined below. 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.

Ptecritical 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.

\\cceptance criteria state that the rod drop time from fully withdrawn to 75. inserted shall be less than 1.66 seconds at the conditions 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 instead of the two-thirds inserted position for data gathering. 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 hot zero power (HZP) steady-state conditions. Acceptance cri-teria require that the measured flow be within allowable limits.

9.1.3.

RC Flow Coastdown The coastdown of RC flow from the tripping of the highest flow RCP from four RCPs running will be measured at HZP conditions. The coastdown of RC flow versus time will then be compared to the required RC flow versus time to de-termine if acceptance criteria are met.

9-1 Babcock & Wilcox

I 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 crit =rion placed on critical boron concentration is that the actual

=

boron concentration must be within 100 ppm boron of the predicted value.

I 9.2.2.

Temperature Reactivity Coefficient The isothermal 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 then decreasing the temperature by 5'F.

During the change in temperature, reactivity feedback is compensated a

by discrete change in rod motion; the change in reactivity is then calculated by the summation of reactivity (obtained from reactivity calculation on a strip chart recorder) associated with the temperature change. Acceptance criteria 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 beem measured, a predicted value of fuel Doppler coefficient of reactivity is added to obtain 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 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-g sists of establishing a deboration rate in the RC system and compensating for m

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 reactimeter data, and differential rod worths are obtained from the measured reactivity worth versus the change in rod group position. The dif ferential rod worths of each of the controlling Babcock & Wilcox 9-2 I

I groups are then summed to obtain integral rod group worths.

The acceptance criteria for the control bank group worths are as follows:

1.

Individual bank 5, 6, 7 worth:

I predicted value - measured value x 100 s 15.

measured value 2.

Sum of groups 5, 6, and 7:

predicted value - measured value x 100 s 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 obtained by adding the incremental changes in reactivity by boration.

Af ter the ejected rod has been borated to 100% withdrawn and equilibrium boron established, the ejected rod is then swapped in versus the controlling rod I

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 x 100 s 20.

measured value 2.

Measured value (error adjusted) s 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 I

at 40, N75, and N100% FP With Nominal Control Rod Position Core power distribution tests are performed at N40, N75, and N100 full power (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 index is established at a nominal FP rod configuration at which the core power distribution was calculated. APSR position is established to provide a I

core power imbalance corresponding to the imbalance at which the core power distribution calculations were performed.

Babcock & Wilcox 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 DNBR must be greater than 1.30.

3.

The value obtained from the extrapolation of the minimum DNBR 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 RPS power / imbalance /

flow trip envelope.

4.

The value obtained from the extrapolation of the worst-case maximum linear heat rate 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 valu x 100 s 8.

measured value W

7.

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

predicted value - measured value S

x 100 12.

measured value 3

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 l

these models for input to safety evaluations.

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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 DNBR and linear heat rate.

1 The power distribution tests performed at 75 and 100% FP are identical to the 40% FP test except that core equilibrium xenon is established prior to the 75 and 100% FP tests.

Accordingly, the 75 and 100% FP measured peak acceptance criteria are as follows:

I Babcock & Wilcox 9-4 I

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1.

The highest measured and predicted radial peaks shall be within the follow-ing limits:

predicted value - measured value x 100 s 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 Vs Excore Detector Imbalance Correlation Verification at %4C% FP Imbalances are set up in the core by control rod positioning.

Imbalances are read simultaneously on the incore detectors and excore power range detectors for various imbalances. The excore versus incore detector offset slopes must be at least 1.15.

If the excore versus incore detector offset slope crite-rion 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 %100% FP The average RC temperature is decreased and then increased by about 5*F at con-stant reactor power. The reactivity associated with each temperature change is obtained from the change in the controlling rod group position. Controlling rod group worth is measured by the fast insert / withdraw method.

The tempera-ture reactivity coefficient is calculated from the measured changes in reac-tivity and temperature.

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

9.3.4.

Power Doppler Reacti.vity Coef ficient at 4100% 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.

Control I

rod group worth is measured using the fast insert / withdraw method. Reactivity 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, adjusted as stated above, and the measured power change.

9-5 Babcock & Wilcox I

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

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

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9.4.

Procedure for Use When Acceptance l

Criteria Are Not Met u

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i If acceptance criteria for any test are not met, an evaluation is performed l

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with participation by Babcock & Wilcox technical personnel as required.

Fur-i ther specific actions depend on evaluation results. These actions can include repeating the tests with more detailed attention to test prerequisites, added tests to search for anomalies, or design personnel performing detailed analy-ses of potential safety problems because of parameter deviation.

Power is not escalated until evaluation shows that plant safety will not be compromised by such escalation, l

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REFERENCES I

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.

I J.11. Taylor (B&W) to S. A. Varga (NRC). Letter, "BPRA Retainer Reinsertion,"

January 14, 1980.

" Program to Determine In-Reactor Performance of B&W Fuels - Cladding Creep Collapse, BAW-10084PA, Rev 2, Babcock & Wilcox, Lynchburg, Virginia, December 1978.

• C. D. Morgan and 11. S. Kao TAFY - Fuel Pin Temperature and Gas Pressure Analysis, BAW-10044 Babcock & Wilcox Lynchburg, Virginia, May 1972.

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

7 B&W Version of PDQ07 Code, BAW-10117A, Babcock & Wilcox Lynchburg, Virginia, January 1977, e Core Calculational Techniques and Procedures, BAW-10ll8A, Babcock & Wilcox.

Lynchburg, Virginia, December 1979.

' Assembly Calculations and Fitted Nuclear Data, BAW-10ll6A Babcock & Wilcox, Lynchburg, Virginia, May 1977.

l' Davis-Besse Nuclear Power Station Unit 1. Cycle 2 Reload Report, BAW-1598, Rev 1. May 1980.

11 Davis-Besse Nuclear Power Station Unit 1. Cycle 2 Reload Report BAW-1598, Rev 2. September 1981.

12 Attachment I to Application to Amend Operating License for Removal of Burn-able Poison Rod and Orifice Kod Assemblies, BAW-1489, Rev 1 Babcock &

Wilcox Lynchburg, Virginia, May 1978.

A-1 Babcock & Wilcox h,

I 18 L. S. Rubenstein (USNRC) to J.11. Taylor (B&W), Letter, " Evaluation of Interim Procedure for Calculating DNBR Reductions Due to Rod Bow," October 18, 1979.

l I" W. L. Bloomfield, et al., ECCS Evaluation of B6W's 177-FA Raised-Loop NSS, l

BAW-10105 Rev 1, Babcock 6 Wilcox, Lynchburg, Virginia July 1975.

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15 J.11. Taylor (B6W) to L. S. Rubenstein (USNRC) Letter, September 5,1980.

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