ML20148J704

Cycle 4 Reload Report

ML20148J704
Person / Time
Site:	Arkansas Nuclear
Issue date:	10/31/1978
From:	ARKANSAS POWER & LIGHT CO.
To:
Shared Package
ML20148J700	List: ML20148J692 ML20148J704
References
BAW-1504, NUDOCS 7811160064
Download: ML20148J704 (54)

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October 1978 ARKANSAS NUCLEAR ONE, UNIT 1

- Cycle 4 Reload Report -

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BABC0CK & WILCOX

} Power Generation Group J Nuclear Power Generation Division l

P, O. Box 1260 Lynchburg, Virginia 24505 L-Babcock & Wilcox r

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CONTENT _S_

Page

1. INTRODUCTION AND

SUMMARY

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

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

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

4. FUEL SYSTEM DESIGN . . . . . . . . . . . . . . . . . . . ..... 4-1 4.1. Fuel Assembly Mechanical Design . . . . . . . . . ..... 4-1 3 4.2. Fuel Rod Design . . . . . . . . . . . . . . . . . ..... 4-1 g 4.2.1. Cladding Collapse . . . . . . . . . . . . ..... 4-1 4.2.2. Cladding Stress . . . . . . . . . . . . . ..... 4-2 4.2.3. Cladding Strain . . . . . . . . . . . . . ..... 4-2 4.3. Therwal Design . . . . . . . . . . . . . . . . . . ..... 4-2 4.4. Material Design . . . . . . . . . . . . . . . . . ..... 4-2 4.5. Operating Experience . . . . . . . . . . . . . . . ..... 4-3

5. NUCLEAR DESIGN . . . . . . . . . . . . . . . . . . . . . ..... 5-1 5.1. Physics Characteristics . . . . . . . . . . . . .. .... 5-1 5.2. Analytical Input . . . . . . . . . . . . . . . . . ..... 5-1 5.3. Changes in Nuclear Design . . . . . . . . . . . . ..... 5-2

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. Raactor 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.

9.2.3.

Temperature Reactivity Coefficient . . . . .....

Control Rod Group Reactivity Worth . . . . .....

9-2 9-2 55 9.3. Power Escalation Tests . . . . . . . . . . . . . . ..... 9-3 Babcock & Wilcox

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l' CONTENTS (Cont 'd) )

l Page 9.3.1. Core Power Distribution Verification at +40, 75, and 100% FP With Nominal Control Rod Group Con-

} figuration . . . . . . . . . . . . . . . . . . . . 9-3 J 9.3.2. Incore Vs Excore Detector Imbalance Corr 21ation Verification at 475% FP . . . . . . . . . . . . . . 9-5

] 9.3.3. Temperature Reactivity Coefficient at 4100% FP . . . 9-5 j 9.3.4. Power Doppler Reactivity Coefficient at 4100% FP . . 9-5 l

9.4. Procedure if Acceptance Criteria Are Not Met . . . . . . . . 9-6

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

[ List of Tables Table 1 4-1. Fuel Design Parameters and Dimensions . . . . . . . . . . . . . . 4-4

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4-2. Fuel Thermal Analysis Parameters . . . . . . . . . . . . . . . . 4-5 5-1. Arkansas Nuclear One, Unit 1 - Cycle 3 and 4 Physics Parameters . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-3

}J 5-2. Shutdown Margin Calculations for ANO-1, Cycle 4 . . . . . . . . . 5-5 6-1. Maximum Design Conditions for. Cycles 3 and 4 . . . . . . . . . . 6-2 7-1. Bounding Values for Allowable LOCA Peak Linear Heat Rates . . . . 7-2 7-2. Comparison of Key Parameters for Accident Analysis . . . . . . . 7-3 List of Figures Figure 3-1. Core Loading Diagram for ANO-1, Cycle 4 . . . . . . . . . . . . 3-3 l 3-2. Enrichment and Burnup Distribution for ANO-1, Cycle 4 . . . . . 3-4

- 3-3. Control Rod Locations and Group Designations for ANO-1 Cycle 4 . 3-5 3-4. BPRA Enrichment and Distribution for ANO-1, Cycle 4 . . . . . . 3-6 5-1. ANO-1 Cycle 4 BOC (4 EFPD) Two-Dimensional Relative Power

, Distribution - Full Power, Equilibrium Xenon, Normal Rod Positions . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-6 8-1. Core Protection Safety Limits . . . . . . . . . . . . . . . . . 8-2 8-2. Protective System Maximum Allowable Setpoints . . . . . . . . . 8-3 8-3. Rod Position Limits for Four Pump Operation From 0 to 100 10 EFPD - ANO-1, Cycle 4 . . . . . . . . . . . . . . . . . . . . . 8-4 8-4. Rod Position. Limits for Four Pump Operation From 100 10 to 250 1 10 EFPD - ANO-1, Cycle 4 . . . . . . . . . . . . . . . . . 8-5

( 8-5. Rod Position Limits for Four Pump Operation After 250 10 EFFD - ANO-1, Cycle 4 . . . . . . . . . . . . . . . . . . . . . 8-6 F

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I, Figures (Cont'd)

Figure Page 8-6. Rod Position Limits for Two and Three Pump Operation From 0 to 100 10 EFPD - ANO-1, Cycle 4 . . . . . . . . . . .... 8.7 8-7. Rod Position Limits for Two and Three Pump Operation From 100 10 to 250 i10 EFPD - AND-1, Cycle 4 . . . . . . . .... 8-8 E

3 8-8. Rod Position Limits for Two and Three Pump Operation After i 250 10 EFPD - ANO-1, Cycle 4 . . . . . . . . . . . . .... 8-9 g 8-9. Operational Power Imbalance Envelope for Operation From 0 to g 100 10 EFPD - ANO-1, Cycle 4 . . . . . . . . . . . . .... 8-10 8-10. Operational Power Imbalance Envelope for Operation From 100 10 to 250 10 EFPD - ANO-1, Cycle 1 . . . . . . . . .... 8-11 8-11. Operational Power Imbalance Envelope for Operation After 250 4

i 10 EFPD - ANO-1, Cycle 4 . . . . . . . . . . . . . . .... 8-12

, 8-12. APSR Position Limits for Operation From 0 to 100 EFPD - ANO-1, g Cycle 4 . . . . . .. . . . . . . . . . . . . NO. . . . .... 8-13 g 8-13. APSR Position Limits for Operation From 100 10 to 250 10 EFPD - ANO-1, Cycle 4 . . . . . . . . . . . . . . . . . .... 8-14 g j

8-14. APSR Position Limits for Operation After 250 10 EFPD - g ANO-1, Cycle 4 . . . . . . . . . . . . . . . . . . . . .... 8-15 8-15. Boric Acid Addition Tank Volume and Concentration Requirements Vs RCS Average Temperature . . . . . . . . . . . . . . .... 8-16 I

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1. INTRODUCTION AND

SUMMARY

This report justifies the operation of the fourth cycle of Arkansas Nuclear One, Unit 1 (ANO-1) at the rated core power of 2568 MWt. Included are the required analyses as outlined in the USNRC document " Guidance for Proposed 3

License Amendments Relating to Refueling," June 1975.

6. To support cycle 4 operation of ANO-1, this report employs analytical tech-niques and design bases established in reports that were previously submitted and accepted by the USNRC and its predececsor (see references).

A brief summary of cycle 3 and 4 reactor parameters related to power capabil-ity is included in section 5 of this report. All of the accidents analyzed in the FSARI have been reviewed for cycle 4 operation.

In those cases where cycle 4 characteristics were conservative compared to those analyzed for pre-vious cycles, no new accident analyses were performed.

The Technical Specifications have been reviewed, and the modifications re-J quired for cycle 4 operation are justified in this report.

Based on the analyses performed, which take into account the postulated effects of fuel densification and the Final Acceptance Criteria for Emergency Core Cool-ing Systems, it has been concluded that ANO-1 can be operated safely for cycle 4 at the rated power level of 2568 MWt.

Because of performance anomalies observed at other B&W plants, orifice rod as-semblies will not be used in ANO-1, cycle 4. The removal of the orifice rod

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-assemblies has been accounted for in the analyses performed for cycle 4. In addition, retainer assemblies will be installed on the fuel assemblies contain-ing BPRAs and on two fuel assemblies containing regenerative neutron sources to provide positive retention during reactor operation.

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2. OPERATING HISTORY 9

The reference cycle for the nuclear and thermal-hydraulic analyses of Arkansas Nuclear One, Unit 1 is the currently operating cycle 3. This cycle 4 design is based on a projected cycle 3 length of 284 EFPD.

No anomalies occurred during cycle 3 that would adversely affect f uel perfor-mance during cycle 4.

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3. GENERAL DESCRIPTION The Arkansas Nuclear One, Unit 1 reactor core is described in detail in sec-tion 3 cf the Arkansas Nuclear Station, Unit 1, Final Safety Analysis Report.I The cycle 4 core, contains 177 fuel assemblies, each of which is a 15 by 15 ar-ray containing 208 fuel rods, 16 control rod guide tubes, and one incore instru-ment guide tube.

The fuel consists of dished-end, cylindrical pellets of urani-um dioxide clad in cold-worked Zircaloy-4. The fuel assemblies in all batches have an average nominal fuel loading of 463.6 kg of uranium. The undensified nominal active fuel lengths, theoretical densities, fuel and fuel rod dimen-sions, and other related fuel parameters may be found in Tables 4-1 and 4-2.

Figure 3-1 is the core loading diagram for ANO-1, cycle 4. The ir.itial enrich-g ments of batches 1C and 4 are 2.06 and 2.64 wt % 235 U, respectively. Batches R S and 6 are enriched to 3,01 and 3.19 wt % 235 U, respectively. All the batch 2 and batch IB assemblies will be discharged at the end of cycle 3. The batch IC assembly will be loaded into the center core location. The batch 4 and 5 assemblics will be shufficd to new locations at the beginning of cycle 4. The fresh batch 6 assemblies will be loaded into the core interior in a symmetric checkerboard pattern. Figure 3-2 is an eighth-core map showing the assembly burnup and enrichment distribution at the beginning of cycle 4.

Reactivity is controlled by 61 full-length Ag-In-Cd control rods, 56 burnable '

poison rod assemblies MP"1.c), c.r.d e>oluble boron shim. In addition to the full-length contcol rods, eight axial pouer shaping rods (APSRs) are provided for additional control of the axial power distribution. The cycle 4 locations of the 69 control rods and the group designations are indicated in Figure 3-3.

The core locations of the total pattern (69 control rods) for cycle 4 are identical to those of the reference cycle indicated in the ANO-1, cycle 3 re-load report.2 The group designations, however, will differ between cycle 4 and the reference cycle in order to minimize power peaking. The cycle 4 loca-l L tions and enrichments of the BPRA clusters are shown in Figure 3-4.

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E The normal system pressure is 2200 psia, and the core average densified nomi-nal heat rate is 5.80 kW/ft at the rated core power of 2568 MWt.

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Figure 3-1. Core Loading Diagram for ANO-1, Cycle 4 Fuel Transf er Canal I

A 5 5 4 5 5 F10 R9 012 R7 F6 7 5 4 5 6 5 6 5 4 5 F12 ull RIO N8 E6 M5 F4 4 6 5 6 4 6 4 6 5 6 4 C

N10 Fil M9 M7 F5 L4 p 5 6 5 6 4 6 4 6 4 6 3 6 5 N14 C3 M10 04 M6 C13 N2 4 5 6 4 6 A 6 4 6 4 6 5 4 E

M12 M14 M11 09 07 MS M2 M4 5 5 6 4 6 5 6 4 6 5 6 4 6 5 5 F t14 gg3 Lig gg N6 K7 L5 L1 L2 5 6 4 6 4 6 4 5 4 6 4 6 4 6 5 K15 Kit K13 N12 R8 N4 K3 K5 K1 y ~H & 5 6 4 6 4 5 E3 5 4 6 4 6 5 4 D13 H12 N13 L12 H15 (Cy 1) Hi F4 D3 H4 N3 K

5 6 4 6 4 6 4 5 4 6 4 6 4 6 5 C15 C11 C13 D12 A8 D4 G3 G5 C1 L 5 5 6 4 6 5 6 4 6 5 6 4 6 5 5 F14 F15 F11 C9 Dio G7 F5 F1 72 M 4 5 6 4 6 4 6 4 6 4 6 5 4 j ti2 E14 Ell C9 C7 E5 E2 E4 i

N 5 6 5 6 4 6 4 6 4 6 5 6 5 D14 03 E10 C12 E6 013 D2 4 6 5 6 4 6 4 6 5 6 4 0

F12 811 E9 E7 85 D6 P 5 4 5 6 5 6 5 4 5 812 D11 A10 P6 A6 D3 84 5 5 4 5 5 I

B10 A9 C4 A7 B6 l l 8

1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 X Batch IIX Previous Core Location Noter Cy It center location contains fuel assembly from cycle 1, i

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I Figure 3-2. Enrichment and Burnup Distribution for ANO-1, Cycle 4 8 9 10 11 12 13 14 15 2,06 3.01 2.64 3.19 2.64 3.19 3.01 2.64 14,854 6,759 18,434 0 15,520 0 12,180 15,524 2.64 3.19 2.64 3.19 2.64 3.19 3.01 17,562 0 19,631 0 16,735 0 7,340 3.01 3.19 2.64 3.19 3.01 3.01 <

12,264 0 15,928 0 6,782 10,957 ,

g 2.64 3.19 3.01 2.64 17,899 0 9,446 19,237 3.01 3.19 3.01 7,495 0 6,764 0 '

18,457 p

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XXX Initial Enrichment Il XXX BOC Burnup, mwd /mtU I

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Figure 3-3. Control Rod Locations and Group Designation for ANO-1 Cycle 4 X

A B

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

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1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 Group No. of rods Function j fX Group No. 1 8 Safety 2 12 Safety 3 8 Safety 4 9 Safety 5 8 Control l 6 8 Control

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5 Figure 3-4. BPRA Enrichment and Distribution for ANO-1, Cycle 4 8 9 10 11 12 13 14 15 l 11 1.15 0.80 K 0.50 1.15 l

L 0.50 0.80 0.80 I

i M 1.15 0.80 1.40 N 1.15 1.40 0.80 l

0 0.80 0.80 0.80 l

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X.XX BPRA Concentration, wt % BqC in A1 203 I!

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4. FUEL SYSTEM DESIGN 4.1. Fuel Assembly Mechanical Design The type of fuel assemblies and pertinent fuel design parameters for ANO-1, cycle 4 are listed in Table 4-1. All fuel assemblies are identical in con-cept and are mechanically interchangeable. All results, references, and identified conservatisms presented in section 4.1 of the ANO-1, cycle 3 re-load report 2 are applicable to the cycle 4 reload core.

Retainer assemblies will be used on fuel assemblies containing BPRAs to pro-vide positive retention during reactor operation. The justification for the design and use of the retainers is described in reference 3, which is appli-cable to the extended cycle 4 of ANO-1. Similar retainer assemblics will be used on the two fuel assemblies containing the regenerative neutron sources.

_4 . 2 . Fuel Rod Design The mechanical evaluation of the fuel rod is discussed below.

4.2.1. Cladding Collapse Creep collapse analyses were performed for three-cycle assembly power histories for ANO-1. The batch 4 fuel is more limiting than the other batches because of its previous incore exposure time. The batch 4 assembly power histories were analyzed and the most limitirg assembly determined.

The power history for the most 11uiting assembly was used to calculate the fast neutron flux level for the anergy range above 1 MeV. The collapse time for the most limiting assembly was conservatively determined to be more than 30,000 EFPH (effective full-power hours), which is longer than the maximum projected three-cycle exposure life of 23,112 hours0.0013 days <br />0.0311 hours <br />1.851852e-4 weeks <br />4.2616e-5 months <br /> (see Table 4-1). The creep collapse analysis was performed based on the conditions set forth in references 2 and 4.

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_4 . 2 . 2 . Cladding Stress The ANO-1 stress parameters for batch 4 and subsequent fuel are enveloped by g

a conservative fuel rod stress analysis. For design evaluation, the primary 5 membrane stress must be less than two-thirds of the minimum specified unir-radiated yield strength, and all stresses must be less than the minimum speci-fled unirradiated yield stiength. In all cases, the margin is in excess of 30%. The following conservatisms with respect to ANO-1 fuel were used in the analysis:

1. Low post-densification internal pressure.

2. Low initial pellet density.

3. High system pressure.

4. Iligh thermal gradient across the cladding.

For the batch 1 reinserted fuel, the results presented in the ANO-1 fuel den-sification reports are applicable.

4.2.3. Cladding Strain The fuel design criteria specify a limit cf 1.0% on cladding plastic tensile circumferential strain. The pellet is designed to assure that cladding plas-tic strain is less than 1% at design local pellet burnup and heat generation rate. The design burnup and heat generation rate are higher than the worst-case values ANO-1 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 value for the cladding ID.

4.3. Thermal Design All fuel assemblies in this core are thermally similar. The fresh batch 6 fuel inserted for cycle 4 operation introduces no significant differences in fuel thetmal performance relative to the other fuel remaining in the core.

The design minimum linear heat rate (LHR) capability and the average fuel temperature for each batch in cycle 4 are shown in Table 4-2. LHR capabili-ties are based on centerline fuel melt and were established using the TAFY-3 code 6 with fuel densification to 96.5% of theoretical density.

4.4. Material Design The chemical compatibility of all possible fuel-cladding-coolant-assembly I interactions for the batch 6 fuel assemblies is identical to that of the present fuel.

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4.5. Operating Experience Babcock & Wilcox operating experience with the Mark B, 15 by la fuel assembly has verified the adecctacy of its design. As of May 31, 1978, the following experience has been accummulated for the eight operating B&W 177-fuel assem-bly plants using the Mark B fuel assembly:

Maximum assembly burnup, mwd /mtU Cumulative net Current electrical Reactor cycle Incore Discharged output, MWh Oconee 1 4 28,850 25,300 21,556,695 Oconee 2 3 29,200 26,800 16,694,459 Oconee 3 3 28,600 27,200 18,016,725 TMI-1 4 25,310 32,200 19,185,215 ANO-1 3 25,230 28,300 15,208,313 Rancho Seco 2 24,770 17,170 11,465,355 Crystal River 3 1 10,700 --

4,978,690 Davis-Besse 1 1 4,000 --

1,734,732

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Table 4-1. Fuel Design Parameters and Dimensions Twice-burned Once-burned assemblies F h hel assemblies, assemblies, batch 4 Batch 1 Batch 5 batch 6 Fuel assembly type Mk-B4 Mk-B3 Mk-B4 Mk-B4 Number of assablies 56 1 56 64 Fuel rod OD (nom.), in. 0.430 0.430 0.430 0.430

, Fuel rod ID (nom.), in. 0.377 0.377 0.377 0.377 l .

Flexible spacers, type Spring Corrugated / Spring Spring j

spring

Rigid spacers, type Zr-4 Zr0 Zr-4 Zr-4 l Uadensified active fuel 142.6 144.0 142.25 142.25

& length (nom.), in.

Fuel pellet OD (mean 0.3700 0.3700 0.3695 0.3695

! specified), in.

f Fuel pellet init:.C density 93.5 92.5 94.0 94.0

(nom.), % TD 4

Initial fuel enri ctment, 2.64 2.06 3.01 3.19 i wt % 235U i

Burnup, BOC, mwd /mtU 17,604 14,854 8662 0 j Cladding collapse time, EFPH >30,000 >30,000 >30,000 >30,000 l Estimated residence time 23,112 21,528 26,112 28,584 c2 (max.), EFPH c-8 o

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Tab 1r +-2. Fuel Thermal Analysis Parameters Batch 1 Batch 4 Batch 5 Batch 6 No. of assemblies 1 56 56 64 Initial density, % TD 92.5 93.5 94.0 94.0 Pellet diameter, in. 0.370 0.370 0.3695 0.3695 Stack height, in. 144 142.6 142.25 142.25 Densified Fuel Parameters Pellet diameter, in. 0.3632 0.3645 0.3646 0.3646 Fuel stack height, in. 141.1 140.49 140.5 140.5 Nominal linear heat rate 5.77 5.80 5.80 5.80

/ 0 2568 MWt, kW/ft Average fuel temperature 1330 1315 1320 1320 at nominal LHR, F LHR capability (center- 20.1 20.15 20.15 20.15 line fuel nelt), kW/ft based on fuel spec values

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5. NUCLEAR DESIGN 5.1. Physics Characteristics Table 5-1 lists the core physics parameters of design cycles 3 and 4. The values for both cycles were generated using PDQ07. The average cycle burnup will be higher in cycle 4 than in the design cycle 3 because of the longer cycle 4 length. Figure 5-1 illustrates a representative relative power dis-tribution for the beginning of cycle 4 at full power with equilibrium xenon and normal rod positions.

f Although both cycles 3 and 4 are rodded cycles with a rod pull at 250 EFPD, the initial BPRA loading, longer design life, and different shuffle pattern for cycle 4 make it difficult to compare the physics parameters of the two cycles. Calculated ejected rod worths and their adherence to criteria are considered at all times in life and at all power levels in the development of the rod position limits presented in section 8. The meximum stuck rod worth for cycle 4 is less than that for the design cycle 3 at both BOC and EOC. All safety criteria associated with these worths are met. The adequacy of the shutdown margin with cycle 4 stuck rod worths is demonstrated in Table 5-2.

The fo.'. lowing conservatisms were applied for the shutdown calculations.

1. Poison material depletion allowance.

2. 10% uncertainty on net rod worth.

3. Flux redistribution penalty.

Flux redistribution was accounted for since the shutdown analysis was calcu-lated using a two-dimensional model. The reference fuel cycle shutdown margin is presented in ANO-1 cycle 3 reload report.2 5.2. Analytical Input The cycle 4 incore measurement calculation constants to be used for computing core power distributions were prepared in the same manner as those for the reference cycle, l

5-1 Babcock & Wilcox

B 5.3. Changes in Nuclear Design There is only one significant core design change between the reference and re-load cycles. This change is the increase in cycle lifetime to 387 EFPD and the subsequent incorporation of BPRAs to aid in reactivity control. The cal-culational methods and design infortnation used to obtain the important nuclear design parameters for this cycle were the same as those used for the reference l cycle.

3; I

! I I

I I

I I

I I

I~

t l

l I 5-2 Babcock & Wilcox

i Table 5-1.

ArkansasNuc1carOne)

Physics Parameters (a Unit 1-Cycle 3and4 Cycle 3 Cycle 4( )

Cycle length, EFPD 294 387 Cycle burnup, mwd /mtU 9200 12,111 Average core burnup - E0C, FMd/mtU 19,405 20,505 Initial core loading, mtU 82.1 82.1 Critical boron - BOC, ppm (no Xe)

HZP(d), group 8 (37.5% wd) 1318 1562 HZP, groups 7 and 8 inserted 1248 1453 HFP, groups 7 and 8 inserted 1065 1246 Critical boron - EOC, ppm (eq Xe) group 8 (37.5% wd) 3 8 Control rod worths - HFP, BOC, % Ak/k Group 6 1.07 1.18 Group 7 0.77 1.02 Group 8 (37.5% wd) 0.42 0.37 Control rod worths - HFP, EOC, % Ak/k Group 7 1.05 1.00 Group 8 (37.5% wd) 0.45 0.47 Max ejected rod worth - HZP, % Ak/k(*

BOC (P8) 0.64 0.76 EOC (N-12) 0.57 0.82 Max stuck rod worth - HZP, % Ak/k BOC (P8) 2.50 1.92 EOC (N-12) 2.22 1.86 Power deficit, HZP to HFP, % Ak/k B0C 1.27 1.38 EOC 2.03 2.28 Doppler coeff - BOC, 10-5 (Ak/k/"F) 100% power (0 Xe) -1.47 -1.57 Doppler coeff - EOC, 10-5 (Ak/k/*F) 100% power (eg Xe) -1.61 -1.71 Moderator coeff - HFP, 10-4 (Ak/k/*F)

BOC (0 Xe, crit ppm, group 8 ins) -0.66 -0.48 EOC (eq Xe, 17 ppm, group 8 ins) -2.62 -2.78 Boron worth - HFP, ppm /% Ak/k BOC 107 118 EOC 97 105 5-3 Babcock & Wilcox

8 Table 5-1. (Cont'd)

Cycle 3 Cycle 4 Xenon worth - HFP, % Ak/k BOC (4 EFPD) 2.66 2.59 3 E0C (equilibrium) 2.75 2.75 g Effective delayed neutron fraction - HFP g BOC 0.00591 0.00617 g EOC 0.00524 0.00517 Cycle 4 data are for the conditions stated in this report.

The cycle 3 core conditions are identified in reference 2.

(

Based on 272 EFPD at 2568 MWt, cycle 2.

Based on a projected cycle 3 lifetime of 284 EFPD.

(

HZP:

hot zero power (532F T"V8); HFP: hot full power (579 T##8).

(e) Ejected rod worth for groups 5 through 8 inserted. '

I I'

I I

I I

B B

I 5-4 Babcock & Wilcox

Table 5-2. Shutdown Margin Calculations for ANO-1, Cycle 4 BOC, % Ak/k 250 EFPD, % Ak/k EOC, % Ak/k Available rod worth Total rod worth, IIZP 9.35 9.64 9.65 Worth reduction due to burnup of poison material -0.35 -0.45 -0.51 Maximum stuck rod, llZP -1.92 ~2.26 -1.86 Net worth 7.08 6.93 7.28 Less 10% uncertainty -0.71 -0.69 -0.73 Total available worth 6.37 6.24 6.55 Required rod worth Power deficit, HFP to HZP 1.38 2.19 2.28

Max allowable inserted rod worth 1.20 1.35 1.35 Flux redistribution 0.40 0.74 1.15 Total required worth 2.98 4.28 4.78 Shutdown margin (total avail.

worth minus total required worth) 3.39 1.96 1.77 Note: Required shutdown margin is 1.00% Ak/k I

I f

5-5 Babcock & Wilcox

8 Figure 5-1. ANO-1 Cycle 4 BOC (4 EFPD) Two-Dimensional Relative Power Distribution - Full Power, Equilibrium Xenon, Normal Rod Positions 8 9 10 11 12 13 14 15 H 0.688 0.941 0.981 1.277 1.116 1.218 0.64 0.436 7

K 0.587 1.272 1.079 1.277 1.073 1.248 0.606 L

1.348 1.384 X 1.012 1.290 1.037 0.462 l

M 1.145 1.328 1.174 0.577 N 1.331 1.003 0.482 l'

I, 0

0.460 li g

, E Il 4 I

XXX 3 1 _ meo Ree 0 _ , No.

Relative Power Density l

E l'

5-6 Babcock & Wilcox

l l

6. THERMAL-HYDRAULIC DESIGN The incoming batch 6 fuel is hydraulically and t ametrically identical to that of batch 5. The cycle 3 and 4 maximum design conditions and significant param-eters are shown in Table 6-1. The methods and models used in the thermal-hydraulic evaluation for cycle 4 are the same as for cycle 3 with the following two exceptions:

(1) orifice rod assemblies (ORAs) will be removed from the core, and (2) new burnable poison rod assembly (BPRA) retainers will be includ-ed for cycle 4 operation.

The cycle 4 core configuration is as follows:

I 56 fuel assemblies with BPRAs and new retainers 69 fuel assemblies with control rods 2 fuel assemblies with neutron sources and retainers 50 fuel assemblies without ORAs Evaluations for cycle 3 operation assumed 44 fuel assemblies with ORAs removed.

g Since there are only six more assemblies with ORAs removed for cycle 4, the B core bypass flow was virtually unchanged. The new BPRA retainers are described in B&W report BAW-1496.3 The thermal-hydraulic design and calculational as-sumptions for the retainers are discussed in section 5 of that report. B&W has completed flow and pressure drop tests that confirm the conservatisms of the retainer analysis.

For cycle 4 evaluations, the hot-maximum-design assembly was assumed to contain a BPRA, and as a result the minimum calculated steady-state DNBR decreased by two points (0.02) relative to cycle 3, from 1.90 to 1.88 (Table 6-1) at 112% overpower.

The potential effect of fuel rod bow on DNBR was considered by incorporating suitable margins into DNB-limited core safety limits and RPS setpoints. The I maximum rod bow magnitude was calculated using the NRC interim model, AC/C 0.065 + 0.001449 /BU, where AC is the rod bow magnitude (mtis), C is the

=

initial gap, and BU is the assembly burnup (mwd /mtU). The resultant DNBR pen-alty, based on the maximum predicted assembly burnup at EOC 3 regardless of F

l 6-1 Babcock & Wilcox h

L l 8 l batch, is 11.2%. The variable low pressure trip functien provides the ther-f mal margin required for L:ie 11.2% DNBR rod bow penalty.

j For cycle 4, the flux /

g flow trip setpoint will be reduced slightly (see section 8) as a result of add-

• f

ing the BPRA retainers and the additional six assemblies with ORAs removed.

The setpoint is based on an assumed two-pump coastdown from 102% indicated power (108% core power), which provides sufficient thermal margin to offset the 11.2% rod bow penalty.

Table 6-1. Maximum Design Conditions for Cycles 3 and 4 I

l Cy 3 Cycle 4 I

besign power level, MWt 2368 2568 System pressure, psia 2200 2200 Reactor coolant flow, % design 106.5 106.5 Vessel inlet / outlet coolant temp at 555.6/602.4 555.6/602.4 100% p'ower, F Ref desig radial-local power 1.78 1.78 peaking fat Ref design axial flux shape .5 cosine 1.5 cosine Hot channel factors: Enthalpy rise 1.011 1.011 Heat flux 1.014 1.014 Flow area 0.98 0.98 g

Active fuel length, i r. . 140.2 140.2 EP Avg heat flux at 100% power, Btu /h-ft 2(a) 175 x 103 175 x 10 3 Max heat flux at 100% power, "eu/h-ft 20) 468 x 103 468 x 103 Ch correlation BAW-2 BAW-2

.h, DNBR (% power) 1.90 (112) 1.88 (112)

.2.03 (108) 2.01 (108) 2.31 (100) 2.30 (100)

(" Heat flux was based on densified length (in hottest core location).

I

) Based on average heat flux with reference peaking.

I:

I 6-2 Babcock & Wilcox g

7. ACCIDFNT AND TRANSIENT ANALYSIS 7.1. Gener_a_1 Saf ety Analysis Each FSARI accident analysis has been examined with respect to changes in cycle 4 parameters to determine the effect of the cycle 4 reload and to en-sure that thermal performance during hypothetical transients is not degraded.

The effects of fuci decsification on the FSAR accident results have been eval-uated and are reported in reference 5. Since batch 6 reload fuel assemblies contein fuel rods whose theoretical density is higher than those considered in the reference 5 report, the conclusions in that reference are still valid.

7.2. Accident Evaluation The key parameters t*mt have the greatest effect on determining the outcome of a transient can typically be classified in three major areas: core thermal parameters, thermal-hydraulic parameters, and kinetica pr.rameters, including the reactivity feed'ack coefficients and control rod worths.

Core thermal properties used in the FSAR ac. __at analysis were design operat-values based on calculational values plus uncertainties. First-core values

'tues) of core thermal parameters and subsequent fuel batches are com-arameters used in cycle 4 analyses in Tabic 4-2. The cycle 4 thermal-t design' conditions are compared to the previous cycle 3 values Ehese parameters are common to 4t1 the accidents considered in A comparison of the key kinetics parameters from the FSAR and grovided in Table 7-2.

a analysis for a B&W 177-FA, lowered-loop NSS has been performed

.m

..o the 1inal Acceptance Criteria ECCS Evaluation Model (reported in BAW-10103)7 The analysis in BAW-10103 is generic since the limiting values of 1

parameters for all plants in this category were used. Furthermore, the combination of average fuea temperature as a function of LHR and lifetime pin pressure data used in the BAW-10103 LOCA lir a analysis is conservative compared to those calculated for this reload. Thus, the analysis and the LOCA 7-1 Babcock & \Vilcox

l' limits reported in BAW-10103 and substantiated by reference 8 provide conserva-tive results for the operation of the reload cycle. Table 7-1 shows the bound- l ing values for allowable LOCA peak LHRs for ANO-1 cycle 4 fuel.

l It is concluded from the examination of cycle 4 core thermal and kinetics prop- l erties, with respect to acceptablc previous cycle values, that this core reload will not adversely affect the ANO-1 plant's ability to operate safely during I

cycle 4. Considering the previously accepted design basis used in the FSAR and subsequent cycles, the transient evaluation of cycle 4 is considered to be bounded by previously accepted analyses. The initial conditions for the tran- I sients in cycle 4 (except the moderator dilution event) are bounded by the l

FSAR , the fuel densification reports, and/or subsequent cycle analyses. The combination of a higher initial boron concentration with a higher boron worth '

will result in insignificantly more severe consequences for the same dilution g

rate than presented in the FSAR for the mederator dilution eveut. An adequate 5 safety margin exists so that the safet; evaluation criteria are not violated even for the maximum hypothetical dilution rate of 500 gpm considered in the FSAR.I I!

Table 7-1. Bounding Values for Allowable LOCA Peak Linear Heat Rates Core elevation, ft Allowable pe,ak linear heat rate, kW/ft I

l l

2 15.5 l 4 16.6 l

t, 18.0 1

8 17.0  !

10 16.0 l'

I!

I 7-2 Babcock & Wilcox

Table 7-2. Comparison of Key Parameters for Accident Analysis FSAR and densification Predicted Parameter report value value Doppler coeff (B0C), Ak/k/'F -1.17 x 10-5 -1.57 x 10-5 Doppler coeff (EOC), Ak/k/*F -1.30 x 10-5 -1.71 x 10-5 Moderator coeff (BOC), Ak/k/*F 0. 0 (") -0.48 x 10-4 Moderator coeff (EOC), Ak/k/*F -4.0 x 10 t,(b) -2.78 x 10-4 All-rod group worth (llZP), % Ak/k 12.9 9.35 Initial boron conc, ppm 1150 1366 Boren reactivity worth (HFP), 100 118 ppm /1% Ak/k Max ejected rod worth (HFP), % Ak/k 0.65 0.55 Dropped rod worth (HFP), % Ak/k 0.65 0.20

("}+0.5 x 10-4 Ak/k/ F was used for the moderator dilution analysis.

( } 3.0 x 10-4 Ak/k/"F was used for the steam line failure analysis, l

7-3 Babcock & Wilcox

~

1

)

l

8. PROPOSED MODIFICATION' TO TECHNICAL SPECIFICATIONS The Technical Specifications have been revised for cycle 4 operstion to account for changes in power peaking and control rod worths inherent with the conver-sion to an 18-month, lumped burnable poison cycle.

The cycle 4 flux / flow setpoint is reduced from 1.06 to 1.057 primarily as a result of the addition of the BPRA retainers as described in B&W design report BAW-1496.3 i

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. Figures 8-1 through 8-15 are revisions to previous Technical Specification limits.

6 8 Babcock & Wilcox I

k . .. . - . .

~

Figure 8-1. Core Protection Satety Limits

- - 120

-18,112 25,112 UNACCEPTABLE OPERATION 1

100 ACCEPTABLE l

4 PUMP 45,95

-45,92 OPERATION g

-18,85.4 25,25.4 3

-. 80 ACCEPTABLE 2 3 & 4 PUMP 45,68.4

-45,65.4 # s OPERATION

-18,58.5 25,58.5 60 I

ACCEPTABLE 3 2,3 & 4 PUMP '

-41.3.41.3

~~ 40 l OPERATION l l

-- 20 I, '

I

. . . _. . . , g

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

I CURVE E' AC' icd COOL ANT FLOW (GPM) 1 374,800 2

3 280,035 184,441 l

I 8-2 Babcock & Wilcox

t -.

Figure.8-2. Protectivu System Maximum Allowable Setpoints THERMAL POWER LEVEL, %

. 120 UNACCEPTABLE OPERATION 10,105.7 105.7 15,?05.7 Mj = 0.747 " 100-y2 = -0. 913 ACCEPTABLE

_31,90

. 4 PUMP 30,92 OPERATION

-10,78.95 80

, , 15,78.95 ACCEPTABLE

-31,63.25 / 3 & 4 PUMP OPERATION

- . 60 30,65.25

-10,52.0 52 15.52.0

. . 40 31,36.3 ACCEPTABLE 30,38.3 2,3 & 4 PUMP OPERATION

[

- - 20

~

S' S M

+ +

ll ll ll 11 m

~

E E E i a 1 i I 8

-60 40- 20 0 20 40 60 Power imoalance, %

l 8-3 Babcock & Wilcox  !

7 t

Figure 8-3. Rod Position Limits for Four Pump Operation From 0 to 100 10 i EFFD - ANO-1, Cycle 4 l 110 1

(180,102) (206,102) gl 100 -

/ RESTRICTED REGION g

(175,92)[ (206.92) 90 -

80 -

I i . (215,80) 70 - 5 a

E 60 -

l 50 '

• ) 40 PERMISSIBLE OPERATING REGION l'

s E 30 g'

20 -

10 (0,0) 0 1 ' ' ' ' ' ' ' ' ' ' ' ' Il W

0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 l Rod index, 5 WD l

l l l 1 l l W 0 20 40 60 80 100 Group 7 1  ! I I I l 0 20 40 60 80 100 1

Group 6 '

l I I I i 1 0 20 40 60 80 100 ,

Group 5 '

l I I l

8-4 Sabcock & Wilcox ,

i Figure 8-4. Rod Position Limits for Four Pump  ;

Operation From 100 10 to '

[ 250 10 EFPD - ANO-1, Cycle 4 110 ,

(132,102) (175,102) (206,102) 100 -

I RESTRICTED OPERATION IN THIS REGION (175,92) o < 2%,92)  !'

90 - REGION NOT ALLOWED 3a _ (174,80) , (230,80) 70 -

SHUTOOWN MARGIN @

LIMIT r g6 E

E 50 -

(60,50)

(300,50) >

o

, 40 -

PERMISSIBLE OPERATING REGION t

5 -

A 30 '

(0,27p 20 -

10 -

l (0,0) 0 ' 8 ' i t i i e i 1 1 I i 1 0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300

- J Index. f WD t i l i l 1 0 20 40 60 80 100 Group 7 I i l I f f 0 20 40 60 80 100 Group 6 1 1 i t i f 0 20 40 60 80 100 Group 5

(

8-5 Babcock & Wilcox  ;

i

.J

1 I

Figure 8-5. Rod Position Limits for Four Pump El Operation After 250 10 EFPD - I ANO-1, Cycle 4 l

110 (154,102) (280,102) 100 ' '

g g

OPERATION IN THIS .

90 - REGION IS NOT ALLOWED (280,92) l go _ RESTRICTED (270,80)

REGION 70 -

W

, E 60 -

h N

(68,50) 50 -

?

40 -

I; l

$ sh E 30 0,30) U p PERHISSIBLE OPERATING REGION l 10

~$

ll (0,0) 0 1  ! ' ' ' I I ' I I i l l I '

0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 Rod index. WD I I i 1 I i 0 20 40 60 80 100 Group 7 t t i I t t 0 20 40 60 80 100 Group 6 I i i l i 1 0 20 40 60 60 Group 5 100 g

m I.

I, l

I 8-6 Babcock & Wilcox

Figure 8-6. Rod Position Limits for Two and Three Pump Operation From 0 to 100 10 EFPD - ANO-1, Cycle 4 RESTRICTED FOR 2 & 3 PUMP l10 (24,102) (175,102) (215,102) ,

i 100 -

RESTRICTED RESTRICTED FOR 3 PUMP FOR 3 PUMP l 90 -

80 -

70 -

E" 60 -

(24,64)

W 50 -

o

', 40 - PERMISSIBLE OPERATING REGION D

R E 30

~

20 -

10 g t i i I i t i l I i i i 1 0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 Rod Index.

• WD I I l I i 1 0 20 40 60 80 100 Group 7 I I I I I J 0 20 40 60 80 100 Group 6 I I I i 1 1 0 70 40 60 80 100 Group 5 8-7 Babcock & Wilcox 3

I Figure 8-7. Rod Position Limits for Two and Three Pump Operation From 100 10 to 250 10 EFPD - ANO-1, Cycle 4 10 -

W (132,102) (174,102) (230,102) 100 -

RESTRICTED OPERATIOR IN THIS [g4 FOR 3 PUNP "' l 90 - RE'll0H IS NOT 8 ALLOWED 80 -

I (79,64)

(300,64) '

ly 60 -

$ RESTRICTED FOR su 50 -2 & 3 PUNP (60,50) r RESTRICTED FOR 3 PUMP g _

f (22,36) 3 PERMISSIBLE OPERATlHG REGION 20 -

10 I (0,0)

O i f i i I l i L i l i l  ! l __

0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 Rod index.

• WD I l l l l 1 0 20 40 60 80 100 t i t i i 1 0 20 40 60 80 100 .

Group 6 i i t i i 1 0 20 40 60 80 Group 5 100 g

3 I

I 8-8 Babcock & Wilcox I

m... - _ _ . . _ _ _ _ ._ _ _ _ _ _ . _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ . _ _ - -

T

-Figure 8-8. Rod' Position Limits for Two and Three Pump Operation After 250. 10 EFPD - l ANO-1, Cycle 4 l 110 S

(154,102) (270,102) l 100 -

l RESTRICTED 90 - OPERATION IN THIS REGION FOR 3 PUNP l lS NOT ALLOWED -l 80 -

70 - (100,70)

E" 60 -

l I

• PERNISSIBLE OPERATING REGION N 50 - (68,50)

T> -(50,45) a '

40

-(26 i + ,

E 30 0,301  !

d RESTRICTED FOR 3 PUNP 20 -

10 -

RESTRICTED FOR 2 & 3 PUNP 0 I t i t ' I i t i I i i 1 1 0 20 40 60 80 100 (20 140 160 180 200 220 240 260 280 300 l Rod index. 7. WD 1 l l 1 i i

{ 0 20 40 Group 7 60 80 100 I i l I i I O 20 40 60 80 100 ,

Group 6

.l l

l 1 1 1 i 1 0 20 40 60 80 100 Group 5 i

l l

l l

1 u.

8-9 Babcock 8. Wilcox

+ -~

) il Ililj)l )11' il' ') ,

MncCMC o

o o8c* O7an%eOQC* TOcmn BC0W%DoW s T2 t t3< OHO" O mon @oMWe*OD mMOa O trO HOO +i HO t

n tmmO i O8H

• oMOg b

[te,. a Ow Mn(mo> .

a N.o= y +o>.N* oM a

I oo g

.m D( - +@.m*eN I o to 0

D g- o.N t o>o - i 0o _ + NCo l No G

N I

O - T I mo n xmCHm oImQ A mmGHx o4mO R

Mmo o2 E mmo o2 P e r

O 4 to N O

E I L G B E I

S 1

&o R S

I M

R o I

Co E

P I

No I

o

- ~ _ - - - ~ ~ - -

i Jo o o bo L. o o g Mo Mo eo 1

1 L

>X e r

oOiD

3CN" N8G

• M o8gO mmToOOX C

0 9 k _OOX tilIl l' >i

Figure 8-10. Operational Power Imbalance Envelope for Operation From 100 10 to ,

250 10 EFPD - ANO-1, Cycle 1 l Power, % of 2568 MWt

-15.3,102 *

- 100 i

-13.4,92 ) . ,

+9.6,92 90

-18.0,80 +11.2,80 80 ,,

i

- 70 '

e 5 -

60 M x E' - 50 RESTRICTED = RESTRICTED REGION -

d 40 w

REGION w = '

) i -

30 w" ,

Q.

20 f ~

- 10 i

i I I f f i l i

) l I

) -50 -40 20 -iO O i0 20 30 40 50 Axial Power imbalance, %

(

r 8-11 Babcock & Wilcox ;i

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m . .

. .k

. . l l R

e E R S E T G R I I O C F 4 i N T i E g 0 D -

2 u 1 r

. c

- ,5 3 i 8

0 8 - -

0 - 1 1 1

0 1 0

8 A 2 i ,

x ,4 AfO i 0 9 1

0 Nop a 2 2 Ore l

P

- r P 1

'$ $j u

%$5 g ,'

o ,1 pOt a o 0 i

w ei w e Cr o e ,

r y an 8 r ct a

- li l 1 i 0 - - - - - - - - - - % eo 2 m 1 2 o nP b

0 0 3 4 5 6 7 8 1

1 4 o a 0 0 0 0 0 9 0 1 f

Aw l 0 0 0 0 fe a 1 2 t r 5 e n 0 c

i "h % 6 8 rI

,e i 2b m

M 5a

% 2 W 0l 0 i

+ + t a

+ 1 1 n

l 5 4 c 6 . . 1e R .

E 4 3 0 3 R S ,4 , ,

E 0 i E T 9 1 En 8 2 0 Fv G R 0 I I 2 Pe O C Dl N T o 4 p B 0 i E e

a D b

c o

c 5 J

k 0 W

i l

c o

x _

g I ' I I l I I 55 I l) , 3 '{i

!'igure 8-12. APSR Position Limits for Operation From 0 to 100 EFPD - ANO-1, Cycle 4 t/o 6,102 24,102

. o o 100 90 -

' '9 ' 4'8 RESTRICTED REGION 30,80

( 0,80 80 ,

RESTRICTED REGION

[ 70 -

h 60 -

8 m

50 -

, 100,50 o

DEt,

. 40 -

l 30 -

PERMISSIBLE OPERATING REGION 20 -

10 -

0 ' ' ' ' ' ' ' ' I O 10 20 30 40 50 60 70 80 90 100 APSR, % WITHORAWN

[- ,

l il 8-13 Babcock & Wilcox  !

I I:

1 Figure 8-13. APSR Position Limits for Operation From 100 10 to 250 10 EFPD -

ANO-1, Cycle 4 17,102

' ' g loo _

5l RESTRICTED REGION I i 17,92 i 90 -

l l

80 -

~

[ 60 -

=

m o 50 -

100,50 f 40 -

I 30 -

PERMISSIBLE OPERATING REGION 20 -

10 -

0 i , , ,  ! , i i i 0 10 20 30 40 3

50 60 70 80 90 100 APSR, % Withdrawn I

I I

I 8-14 Babcock & Wilcox

Figure 8-14. APSR Position Limits for Operation After 250 10 EFPD - ANO-1, Cycle 4 20,102 100 -

< 20,92 RESTRICTED REGION 90 -

80 - 30,80 70 -

E e 60 -

N 50 a 100,50 40 PERMISSIBLE OPERATING REGION 30 20

( 10 -

(~ o l I I I I I I I l 0 l'0 20 30 40 50 60 70 80 90 IOO APSR, % Withdrawn

[

)

i 8-15 Babcock & Wilcox g

\

Figure 8-15. Boric Acid Addition Tank Volume and Concentration Requirements Vs RCS Average Temperature 7000 I.

(579 F,6138 gal)

Il n 1 E 5000 -

l 2

l w" OPERATION AB0VE AND T0 (532 F,4427 gal) (579 F,4392 gal) 3 g THE LEFT OF THE CURVES 3 IS ACCEPTABLE 5

4000 ' -

(500 F,3826 gal) m

[ (532 F,3166 gal) g a 3

{ 3000- -

4

!j 8700 PPM BORIC I

3 a

~

ACIO (400 F,2103 gal) l 12000 PPM BORIC ACIO 1000 - (300 F,877 gal) l (300'F,628 gal) t i i  :

0 200 300 400 500 600 RCS Average Temperature (F)

]

8-16 Babcock & Wilcox

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 with removal of the orifice rods and addition of the BPRAs.

9.1. Precritical Tests 9.1.1. Control Rod Trip Test Precritical control rod drop times are recorded for all control rods at hot full flow conditions before zero power physics testing begins. Acceptance criteria state that the rod drop time from fully withdrawn to 3/4-inserted shall be less than 1.46 seconds at the conditions above.

It should be noted that safety analysis calculations are based on a rod drop from fully witi. drawn to 2/3-inserted. Since the most accurate position in-dication is obtained from the zone reference switch at the 3/4-inserted posi-( tion, this position is used instead of the 2/3-inserted position for data gathering.

9.1.2. Reactor Coolant Flow i

Reactor coolant flow with four RC pumps running will be measured at hot zero power, steady-state conditions. Acceptance criteria require that the measured flow be within allowable limits.

9.1.3. RC Flow Coastdown The constdown of reactor coolant flow from the tripping of the most limiting reactor coolant pump combination from four pumps running will te measured at hot zero power conditions. The coastdown of reactor coolant flow versus time will then be compared to the required coolant flow versus time to determine if acceptance is me;..

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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 concentration determined. The critical boron concentration is calculated by 1

correcting for any rod withdrawal required in achieving equilibrium boron, g

l The acceptance criteria placed on critical boron concentration is that the Wl 1

actual baron concentration must be within 100 ppm B of the predicted value. '

9.2.2. Temperature Reactivity Coefficient The isothermal temperature coefficient is measured at approximately the all- 1 rods-out configuration and at the hot zero power rod insertion limit. The '

average coolant temperature is varied by first decreasing then increasing tem- l perature by 5*F. During the change in temperature, reactivity feedback is com- .

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pensated by discrete change in rod motion if necessary; change in reactivity is then calculated by the summation of reactivity (obtained from reactivity l calculation on strip chart recorder) associated with the temperature change.

Acceptance criteria state that the measured value shall not dif"er from the predicted value by more than 0.4 x 10~4 Ak/k/*F.

The moderattsr coefficient of reactivity is calculated in conjunction with the tamperature coefficient measurement. After the temperature coefficient has been measure:1, a predicted value of f uel Doppler coefficient of reactivity is added to obtain the moderator coefficient. This value must not be in excess of the Technical Specification limit of +0.5 x 10" Ak/k/*F.

9.2.3. Control Rod Group Reactivity Worth Control bank group reactivity worths (groups 5, 6, and 7) are measured at hot zero power conditions using the boron / rod swap method. The boron / rod swap method consists of establishing a deboration rate in the RCS and compensating for the reactivity changes of this deboration by inserting control rod groups 7, 6, and 5 in incremental steps. The reactivity changes that occur during these treasutements are calculated based on reactivity calculator data, and differential reactivity worths.are obtained from the measurcd reactivity worth versus the change in rod group position. The differential reactivit y values =

are then summed to obtain integral rod group worths for each of the controlling groups. g W

9-2 Dabcock & Wilcox

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

1. Individual bank 5, 6, 7 worth:

predicted value - measured value measured value x 100 -< 15

2. Sum of groups 5, 6, ar.d 7:

predicted value - measured value measured value x 100 < 10 9.2.4. Ejected Control Rod Reactivity Worth After the CRA groups have been positioned approximately at the rod insertion limit, the ejected rod is borated to 100% wd and the worth obtained by adding the incremental changes in reactivity.

Af ter the ejected rod has been borated to 100% wd and equilibrium boron estab-f lished, the ejected rod is then swapped in versus the controlling rod group and the worth determined by the change in controlling rod group position.

Acceptance criteria for the ejected rod worth test are as follows:

predicted value - measured valu measured value x 100 < 20

2. Measured value (error adjusted) 5 1.0% ok/k 9.3. Power Escalation Tests 9.3.1. Core Power Distribution Verification at 40, 75, and 100% FP With Nominal Control Rod Group Configuration Core power distribution tests are performed at 40, 75, and 100% FP. The test at 40% FP is essentfally a check on power distribution in the core to bring attention to any abnormalities before escalating to the 752 FP plateau. Rod index is establishe.d at a nominal full power configuration which it where the core power distribution calculations are performed. APSR position is estab-11shed to provide a core power imbalance corresponding to the imbalance where the core power distribution calculaticna are performed

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

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

2. The minimum DNBR must be greater then 1.30.

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I l 3. The value obtained from the extrapolation of the minimum DNBR to the next M power plateau overpower trip setpoint must be greater than 1.30, 4

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4. The value obtained from the extrapolation of tha worst-case maximum linear I j heat rate to tPc next power plateau overpower trip setpoint must be less than the fuel melt limit. W

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

6. The highest measured radial peak and the highest predicted radial peak shall be within the following limits:

i f predicted value - measured value measured value x 100 58 l g

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7. The highest measured total peak and the highest predicted total peak shall r

be within the following limits:

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predicted value - measured value.x 100 < 12 l

measured value l Itents 1, 2, 5, 6, and 7 above are established for the purpose of verifying i-l core nuclear and thermal calculational methods, thereby verifying the accept-l abili.ty of data fron theso models for input to safety evaluations. 3 l Items 3 and 4 establish the criteria whereby escalation to the next power j

plateau may be accomplished without exceeding any safety limits specified b y g

W the safety analysis with regard to DNBR and linear heat rate.

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

75 and 100% FP tests. Aacordingly, the 75 and 100% FP measured radial and total peak acceptance criteria are as follows:

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1. The highest n:easurud radial peak and the highest predicted radial peak i

shall be within the following limits:

predicted value - measured vah e measured value x 100 < 5 I

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2. The highest measured total peak and the highest predicted total peak shall be within the following limits:

( predicted value - measured value measured value x 100 s 7.5

( 9.3.2. Incore Vs Excore Detector Imbalance Correlation Verification at %75% FP Imbalances are set up in the core by control rod positioning while operating

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at a steady-state power icvel. Imbalances are read simultaneously on the in-core detectors and excore power range detectors for various imbalances. The various incore and excore measured imbalance values are least-squares fit to determine the ratio of excore to incore imbalances. If the ratio is less than 1.0 (compensated for measurement and analytical uncertainties), gain amplifiers in the excore detector signal processing equipment are adjusted to obtain the required ratio.

9.3.3. Temperature Reactivity Coefficient at 4100% FP The average reactor coolant temperature is decreased and then increased by about 5*F at constant reactor power. The reactivity associated with each tem-perature change is obtained from the change in the controlling rod group posi-tion. The temperature reactivity coefficient is calculsted from the measured

( reactivity change and the measured temperature change.

Acceptance criteria are that the moderator temperature coefficient shall be negative when above 95% full power.

9.3.4. Power Doppler Reactivity Coefficient at N100% 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. Reactiv-

[' ity corrections are made for changes in x on and reactor coolant temperature which occur during the measurement. The power Doppler reactivity coefficient 1 is calculated from the measured reactivity change, adjusted as stated above, and the measured power change.

Acceptance criteria state that the measured value shall be more negative than

-0.55 x 10-4 Ak/k-% FP.

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9. 4 .' Procedure if Acceptance Criteria Are Not Met

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4 An evaluation is performe(1 before the test program is continued if acceptance criteria for any test are not met. This evaluation is performed by site test personnel, with participation by Babcock & Wilcox technical personnel as re-quired. Further specific actions depend on evaluation results. Theskagions j .

can include test reperformance with more detailed attention to test prerequi-sites, added tests to search for anomalies, or detailed analysis of potential i

safety problems because of parameter deviation by design personnel. The plant is not escalated in power until eval.uation shows that plant safety will not be 1 j compromised by escalation.

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10. REFERENCES 1

Arkansas Nuclear One, Unit 1 Final Safety Analysis Report, Docket No. 50-313,

( Arkansas Power & Light Co.

2 Arkansas Nuclear One, Unit 1 - Cycle 3 Reload Report, BAW-1471, Babcock &

Wilcox, Lynchburg, Virginia, Nove- c 1977.

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

4 Program to Determine In-Reactor Performance of B&W Fuels - Cladding Creep Collapse, BAW-10084, Rev. 1, Babcock & Wilcox, Lynchburg, Virginia, Novem-ber 1976.

5 Arkansas Nuclear One, Unit 1 - Fuel Densification Report, BAW-1391. Babcock

& Wilcox, Lynchburg, Virginia, June 1973.

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

7 ECCS Analysis of B&W's 177-FA Lowered Loop NSS, BAW-10103, Rev. 1, Babcock

& Wilcox, Lynchburg, Virginia, September 1975.

8 J. H. Taylor (B&W Licensing) to R. L. Baer (Reactor Safety Branch, USNRC),

Letter, July 8, 1977.

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