ML19339A550

Cycle 5 Reload Rept.

ML19339A550
Person / Time
Site:	Arkansas Nuclear
Issue date:	07/31/1980
From:	BABCOCK & WILCOX CO.
To:
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ML19339A548	List: ML19339A547 ML19339A550
References
BAW-1606, NUDOCS 8011040249
Download: ML19339A550 (80)

Text

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Io BAW-1606 July 1980 I

I ARKANSAS NUCLEAR ONE, UNIT 1

- Cycle 5 Reload Report -

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I I BABCOCK & WILCOX

,l Power Generation Group B Nuclear Power Generation Division ,

P. O. Box 1260 Lynchburg, Virginia 24505 Babcock & Wilcox I .. _____ T o /t o Vo X V ? . -- - .- -

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CONTENTS Fage

1. INTRODUCTION AND

SUMMARY

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

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

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

4. FUEL SYSTEM DESIGN . . . . . . . . . . . . . . . . . . . . .... 4-1

E 4.1. Fuel Assembly Mechanical Design . . . . . . . . . . .... 4-1 3 4.2. Fuel Rod Design . . . . . . . . . . . . . . . . . . .... 4-1 4.2.1. Cladding Collapse . . . . . . . . . . . . . .... 4-1 4.2.2. Cladding Stress . . . . . . . . . . . . . . .... 4-2

4.2.3. Cladding Strain . . . . . . . . . . . . . . .... 4-2 4.3. Thermal Design . . . . . . . . . . . . . . . . . . . .... 4-2 4.4. Material Design . . . . . . . . . . . . . . . . . . .... 4-3 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 l

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.

.g STARTUP PROGRAM - PHYSICS TESTING . . . . . . . . . . . . .... 9-1 5 9.1. Precritical Tests . . . . . . . . . . . . . . . . . .... 9-1 9.1.1. Control Rod Trip Test . . . . . . . . . . . .... 9-1 l 9.1.2. RC Flow . . . . . . . . . . . . . . . . . . .... 9-1 l

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 9.2.3. Control Rod Group Reactivity Worth . . . . . . . . . 9-2 9.2.4. Ejected Control Rod Reactivity Worth . . . . .... 9-3 I l

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I, CONTENTS (Cont'd)

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9.3. Power Escalation Tests . . . . . . . . . . . . . . . .... 9-3 1 9.3.1. Core Power Distribution Verification at 440, 75, and 100% FP With Nominal Control Rod Position . . . . . . . . . . . . .... 9-3 9.3.2. Incore Vs Excore Detector Imbalance Correlation Verification at 440% FP . . . . .... 9-5 g 9.3.3. Temperature Reactivity Coefficient at 4100% FP . .. 9-5 3 9.3.4. Power Doppler Reactivity Coefficient at %100% FP . . 9-5 9.4. Procedure for Use if Acceptance Criteria Not Met . . .... 9-6 REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . .... A-1 I

List of Tables Table 4-1. Fuel Desigr. Parameters and Dimensions . . . . . . . . . .... 4-4 4-2. Fuel Thermal Analysis Parameters . . . . . . . . . . . . .... 4-5 5-1. Physics Parameters for ANO-1, Cycles 4 and 5 . . . . . . .... 5-2 5-2. Shutdown Margin Calculations for ANO-1, Cycle 5 . . . . .... 5-4 6-1. Maximum Design Conditions, Cycles 4 and 5 . . . . . . . .... 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 g 8-1. Reactor Protection System Trip Setting Limits . . . . . .... 8-16 g List of Figures I

3-1. Core Loading Diagram for ANO-1, Cycle 5 . . . . . . . . .... 3-3 3-2. Enrichment and Burnup Distribution.for ANO-1, Cycle 5 . .... 3-4 3-3. Control Rod Locations and Group Designations for ANO-1, Cycle 5 . . . . . . . . . . . . . . . . . . . . . . . 3-5 3-4. BPRA Enrichment and Distribution for ANO-1, Cycle 5 . . .... 3-6 5-1. ANO-1 Cycle 5 BOC Two-Dimensional Relative Power Distribution - Full Power, Equilibrium Xenon, Normal Rod Positions . . . . . . . . . . . . . . . . . . .... 5-5 8-1. Core Protection Safety Limits . . . . . . . . . . . . . .... 8-17 8-2. Core Proteciton Safety Limits . . . . . . . . . . . . . .... 8-18 8-3. Core Protection Safety Limits . . . . . . . . . . . . . .... 8-19 3 8-4. Protective System Maximum Allowable Setpoints . . . . . .... 8-20 g 8-5. Protective System Maximum Allowable Setpoints . . . . . .... 8-21 8-6. Boric Acid Addition Tank Volume and Concentration Requirements Vs RCS Average Temperature . . . . . . . . .... 8-22

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l Figures (Cont'd)

Figure Page 8-7. Rod Position Limits for Four-Pump Operation From 0 to 300 ! 10 EFPD - ANO-1, Cycle 5 . . . . . . . . . . . . . . 8-23 I 8-8.

8-9.

Rod Position Limits for Four-Pump Operation From 300 10 to S60 10 EFPD - ANO-1, Cycle 5 . . . . . . ....

Rod Position Limits for Four-Pump Operation From 8-24 360 1 10 to 387 1 10 EFPD - ANO-1, Cycle 5 . . . . . . .... 8-25 8-10. Rod Position Limits for Three-Pump Operation From 0 to 300 10 EFPD - ANO-1, Cycle 5 . . . . . . . . . . . . . . 8-26 8-11. Rod Position Limits for Three-Pump Operation From I 8-12.

300 10 to 360 2 10 EFPD - ANO-1, Cycle 5 . . . . . . ....

Rod Position Limits for Three-Pump Operation From 360 1 10 to 387 ! 10 EFPD - ANO-1, cycle 5 . . . . . . ....

8-27 8-28 8-13. Rod Position Limits for Two-Pump Operation From I 8-14.

0 to 300 ! 10 EFF3 - ANO-1, Cycle 5 . . . . . . . . . . . . . . 8-29 Rod Position Limits for Two-Pump Operation From 300 10 to 360 2 10 EFPD - ANO-1, Cycle 5 . . . . . . .... 8-30 I 8-15.

8-16.

Rod Position Limits for Two-Pump Operation From 360 ! 10 to 387 10 EFPD - ANO-1, Cycle 5 . . . . . . ....

Operational Power Imbalance Envelope for Operation 8-31 I 8-17.

From 0 to 300 1 10 EFPD - ANO-1, Cycle 5 . . . . . . . ....

Operational Power Imbalarce Envelope for Operation From 300 2 IL to 360 10 EFPD - ANO-1, Cycle 5 . . . . . . . .

8-32 8-33 8-18. Operational Pows- Imbalance Envelope for Operation I 8-19.

From 360 1 10 to 307 ! 10 EFPD - ANO-1, Cycle 5 . . . . . . . .

APSR Position Limits for Operation From 8-34 0 to 300 10 EFPD - ANO-1, Cycle 5 . . . . . . . . . . .... 8-35 8-20. APSR Position Limits for Operation From 300 ! 10 I to 360 2 10 EFPD - ANO-1, Cycle 5 . . . . . . . . . . . .... 8-36 8-21. APSR Position Limits for Operation From 360 ! 10 to 387 10 EFFD - ANO-1, Cycle 5 . . . . . . . . . . . . . . . 8-37 I

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1. INTRODUCTION AND

SUMMARY

This report justifies the operation of the fifth 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 License Amendments Relating to Refueling," June 1975.

To support cycle 5 operation of ANO-1, this report employs analytical tech-niques and design bases established in reports that have been submitted to and accepted by the USNRC and its predecessor, the USAEC (see references). l The cycle 4 and 5 reactor parameters related to power capability are summarized briefly in section 5 of this report. All of the accidents analyzed in the FSAR I have been reviewed for cycle 5 operation. In those cases where cycle 5 characteristics were conservative compared to those analyzed for previous cy-eles, no new accident analyses were performed.

The Technical Specifications have been reviewed, and the modifications required fiir cycle 5 operation are justified in this report.

Bosed on the analyses performed, which take into account the postulated effects of fuel densification and the Final Acceptance Criteria for Emergency Core Cooling Systems, it has been concluded that ANO-1 can be operated safely for cycle 5 at a rated power level of 2568 MWt.

The cycle 5 core for ANO-1 will contain four lead test assemblies (LTAs).

These assemblies are part of a Department of Energy Extended Burnup Test pro-gram. The LTA design is dnscribed in reference 2.

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2. OPERATING HISTORY j

k The reference cycle for the nuclear and thermal-hydraulic analyses of ANO-1 is l the currently operating cycle 4. This cycle 5 design is based on a design cy-cle 4 length of 387 ! 10 EFPD (effective full-power days).

1 j No anomalies occurred during cycle 4 that would adversely affect fuel per-formance during cycle 5.

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3. GENERAL DESCRIPTION The ANO-1 reactor core is described in detail in section 3 of the Arkansas Nuclear Station, Unit 1, Final Safety Analysis Report (FSAR).1 The cycle 5 core contains 177 fuel assemblies, each of which is a 15 by 15 array containing 208 fuel rods, 16 control rod gu_Je tubes, and one incore instrument guide tube. The fuel comprises dished-end, cylindrical pellets of uranium dioxide clad in cold-worked Zircaloy-4. The fuel assemlbies in all batches have an average nominal fuel loading of 463.6 kg of uranium, with the exception of four batch 7 lead test assemblies (LTAs), which have a nominal loading of 440.0 kg uranium. The undensified nominal active fuel lengths, theoretical densities, fuel and fuel rod dimensions, and other related fuel parameters are given in Tables 4-1 and 4-2 for all fuel assemblies except the LTAs; the corresponding parameters for the LTAs are included in reference 2.

Figure 3-1 is the core loading diagram for ANO-1, cycle 5. The initial en-richments of batches 5B, 6, and 7 are 3.01, 3.19, and 2.95 wt % 2ssU, respec-tively. All the batch 1c and batch 4 assemblies and 11 of the twice-burned l batch 5 assemblies will be discharged at the end of cycle 4. The remaining 45 twice-burned batch 5 assemblies (designated batch 5B) will be shuffled to the core interior. The 64 once-burned batch 6 assemblies will be shuffled to locations on or near the core periphery. The 68 fresh batch 7 assemblies will be loaded in a symmetric checkerboard pattern throughout the core. Figure 3-2 is an eighth-core map showing the assembly burnup and enrichment distribution at the beginning of cycle 5.

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Reactivity is controlled by 61 full-length Ag-In-Cd control rods, 64 BPRAs, and soluble boron shim. In addition to the full-length control rods, eight l l

axial power shaping rods (APSRs) are provided for additional control of the axial power distribution. The cycle 5 locations of the 69 control rods and the grc op designations are indicated in Figure 3-3. The core locations of the total pattern (69 control rods) for cycle 5 are identical to those of the reference cycle indicated in the reload report for ANO-1, cycle 4.3 The Babcock & Wilcox 3-1

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group designations, however, differ between cycle 5 and the reference cycle in

.I in order to minimize power peaking. The cycle 5 locations and enrichments of j the BPRAs are shown in Figure 3-4.

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Figure 3-1. Core Loading Diagram for ANO-1, Cycle 5 FULL TRANSTER CANAL Y 6 6 7 6 6 A K4 K2 K14 K12 6 6 6 58 7 6 6 6 8 I M4 L3 N3 P6 N13 L13 M12 6 6 6 6 6 6 I I I # I C H13 K6 L5 L11 K10 08 6 6 7 58 58 58 7 6 6 D I I I I 011 MS P A7 P8 A9 P H5 05 g 6 6 7 58 58 58 58 7 6 6 7 7 7 C10 F9 LTA 04 A6 A10 012 LTA F7 C6 I F 6

09 6

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58 G1 I

58 P10 I

58 K8 I

58 L2 I

58 GIS I

6 C4 6

07 6 6 58 58 58 58 58 6 6 G I I I I I I 39 E10 F1 P12 P4 N2 FIS E6 B7 58 58 58 58 58 SB $8 58 58 -- Y W-H I I I I L14 H14 H9 N14 011 02 H7 H2 F2 6 6 58 58 58 58 58 6 6 K I I I I 7 P9 M10 L1 D14 812 84 L15 M6 P7 6 6 58 58 58 58 58 6 6 I I I I I L M9 012 K1 F14 G8 B6 KIS 04 N7 6 6 7 58 58 58 58 7 6 6 M I I I 010 L9 LTA N4 R6 RIO N12 LTA L7 06 y 6 6 7 Sis 58 $8 7 6 6 7 7 7 7 N11 H11 P R7 88 R9 P E8 NS 6 6 6 6 6 6 I I I I I 0 C8 C6 F5 Fil GIO H3 p 6 6 6 58 6 6 6 7 7 E4 F3 03 810 013 F13 E12 6 6 6 6 I

R G4 G2 G14 G12 I

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

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Batch Previous core location (LTA = Lead Test Assembly P = Precharacterized Standard Mark B Assembly) 3-3 Babcock & Wilcox

Figure 3-2. Enrichment and Burnup Distribution f or ANO-1, Cycle 5 8 9 10 11 12 13 14 15 3.01 3.01 l3.01 2.95 3.01 2.95 3.01 2.95 H

22777 12750 20232 0 22251 0 18566 0 3.01 2.95 3.01 2.95 3.19 2.95 3.19 K

12750 0 16714 0 16635 0 14433 3.01 2.95 3.01 2.95 3.19 3.19 L

18559 0 14941 0 11872 15770 3.01 2.95 3.19 3.19 22391 0 16322 23336 3.19 2.95 3.19 16162 0 21295 3.19 0

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Initial Enrichment, X.XX wt x 235u XXXXX BOC Burnup, mwd /mtU 3-4 Babcock & Wilcox

I I Figure 3-3. Control Rod Locations and Group Designations for ANO-1, Cycle 5 5 x 1

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3 4 7 4 C l 5 5 1 D 6 8 3 8 6 E l 3 2 2 3 1 F 4 8 7 6 7 8 4 o 5 2 4 4 2 5

'I w- 7 3 6 2 6 3 7 -Y H

K 5 2 4 4 2 5 L 4 8 7 6 8 I

7 4 g 1 3 2 2 3 1 N 6 8 3 8 6 0 1 5 5 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 No. of

, Group Rods Function

' 1 8 Safety X Group Number 2 9 Safety 3 8 Safety 4 12 Safety 5 8 Control 6 8 Control I 7 8

8 Control APSRs

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I Figure 3-4. BPRA Enrichment and Distribution for ANO-1, Cycle 5 8 9 10 11 12 13 14 15 H 1.1 0.8 I

K 0.8 1.1 0.2 l

L 0.8 0.8 0.5 I

M 1.1 0.8 0.8

,., 0.8 0.2 I 0.8 I

O 0.5 0.2 I

p 0.2 R

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X.X LBP Concentration, wt % Br,C in Al203 I

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4. FUEL SYSTEM DESIGN I 4.1. Fuel Assembly Mechanical Design The type of fuel assemblies and pertinent fuel design parameters for ANO-1 cy-I cle 5 are listed in Table 5-1. All fuel assemblies listed are identical in concept and are mechanically interchangeable. All results, references, and 3

identified conservatisms presented in section 4.1 of the cycle 4 reload report are applicable to these assemblies. In addition to the assemblies listed, four lead test assemblies (LTAs) are being inserted with batch 7. One stan-dard Mark B fuel assembly will contain annealed guide tubes to compare with the LTAs. The analysis and justification for the LTAs and annealed guide tubes are reported in reference 2.

Retainer assemblies will be used on the fuel assemblies that contain BPRAs to provide positive retention during reactor operation. This will be the second cycle of operation for the retainer assemblies. The justification for the de-sign and use of the retainers for two cycles is described in reference 4, and g is applicable to ANO-1, cycle 5.

4 Similar retainer assemblies will be used on 5 the two fuel assemblies containing the regenerative neutron sources.

4.2. Fuel Rod Design 4.2.1. Cladding Collapse Creep collapse analyses were performed for three-cycle assembly power histories for ANO-1. The batch 5 fuel is more limiting than the other batches because of its previous incore exposure time. The batch 5 assembly power histories were l analyzed and the most limiting assembly determined.

The power history for the most limiting assembly was used to calculate the fast neutron flux level for the energy range above 1 MeV. The collapse time for the most limiting assembly was conservatively determined to be more than 30,000 I EFPH, which is longer than the maximum projected three-cycle exposure time of 25,356 EFPH (Table 4-1). The creep collapse analysis was performed based on the conditions set forth in references 3 and 5.

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I 4.2.2. Cladding Stress The ANO-1 stress parameters for batch 4 and subsequent fuel are enveloped by a conservative fuel rod stress analysis. For design evaluation, the primary 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-fied unirradiated yield strength. In all cases, the margin is greater than 30%. The following conservatisms with respect to the ANO-1 fuel were used in the analysis:

1. Low post-densification internal pressure.

2. Low initial pellet density.

3. High system pressure.

4. liigh thermal gradient across the cladding.

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 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 that 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

.t1 fuel in the cycle 5 core is thermally similar except the four LTAs. Ther-i.1 analysis of the LTA fuel was performed with the TACO-2 code6 , while all remaining fuel was analyzed with the TAFY-3 code.7 Comparison of these analy-nes shows that the LTAs are less limiting than the standard fuel at all tires during cycle 5. The results of the thermal design evaluation of the LTAs are given in reference 2. The minimum linear heat rate (LilR) capability and the average fuel temperature for each batch are given in Table 4-2. LilR capabil- IW ities are based on centerline fuel melt limits. All cycle 5 core protection limits are based on an LliR of 20.15 kW/ft as determined by TAFY-3. No credic was taken for the increased MIR capability of the LTAs.

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l 4.4. Material Design The chemical compatibility of all possible fuel-cladding-coolant-assembly in-I teractions for the batch 7 fuel assemblies is identical to that of the present fuel.

4.5. Operating Experience Babcock & Wilcox operating experience with the Mark B 15 by 15 fuel assembly has verified the adequacy of its design. As of March 31, 1980, the following experience has been accumulated for the eight operating B6W 177-fuel assembly I plants using the Mark B fuel assembly:

"* W Cumulative ne h g y Current electr. output, Reactor cycle Incore Discharged MWh Oconee 1 6 19,600 40,000 29,238,460 Oconee 2 5 23,400 33,700 26,189,723 Oconee 3 5 26,300 29,400 25,705,247 TMI-l 4 32,400 32,200 28,840,053 ANO-1 4 25,100 33,222 22,996,378 Rancho Seco 3 37,729 29,378 20,317,332 Crystal River 3 2 23,194 23,194 11,400,975 Davis-Besse 1 1 14,600 --

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(" As of March 31, 1980.

As of February 29, 1980, the latest data available.

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Table 4-1. Fuel Design Parameters and Dimensions Batch 5 Batch 6 Batch 7 Fuel assembly type Mark B4 Mark B4 Mark B4 No. of assemblies 45 64 64(*)

Fuel rod OD (nom), in. 0.430 0.430 0.430 Fuel rod ID (nom), in. 0.377 0.377 0.377 Flexible spacers, type Spring Spring Spring Rigid spacers, type Zr-4 Zr-4 Zr-4 Undensitied active fuel 142.25 142.25 141.80 length (nom), in.

Fuel pellet OD (mean speci- 0.3695 0.3695 0.3686 fied), in.

Fuel pellet initial density 94.0 94.0 95.0 (nom), % TD Initial fuel enrichment, 3.01 3.19 2.95 23s wt % U Burnup, BOC, mwd /mtU 17,467 15,208 0 l Cladding collapse time, EFPH >30,000 >30,000 >30,000 Estimated residence time, 25,536 27,864 27,864 (max), EFPH (a)Fout LTAs were also analyzed; the results are reported in reference 2.

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Table 4-2. Fuel Thermal Analysis Parameters I= Batch SB Batch 6 Batch 7 No. of assemblies 45 64 64(a)

Initial density, % TD 94.0 94.0 95.0

"" "' d'"*"'*" '"- '3 '"

.W E , Stack height, in. 142.25 142.25 141.80 i

i3 Densified Fuel Parameters ig Pellet diameter, in. 0.3646 0.3646 0.3649 i Fuel stack height, in. 140.5 140.5 140.74

= Nominal linear heat rate 5.80 5.80 5.79

, at 2568 MWt, kW/ft Avg fuel temperature at 1320 1320 1310 nominal LHR, F LHR capability, kW/ft( ) 20.15 20.15 20.15

I Nominal core avg LHR = 5.80 kW/ft at 2568 MWt.

(# Four LTAs were also analyzed; the results are reported in reference 2.

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Centerline fuel melt based on fuel specification values.

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5. NUCLEAR DESIGN 5.1. Physics Characteristics Table 5-1 lists the core physics parameters of design cycles 4 and 5. The values for both cycles sere generated using PDQ07. Since the core has not yet reached an equilibrium cycle, differences in core physics parameters are to be expected between cycles. Figure 5-1 illustrates a representative rela-tive power distribution for the beginning of cycle 5 at full power with equi-librium xenon and nominal rod positions.

Although cycles 4 and 5 are both rodded, 387-EFPD cycles, differences in feed enrichment, BPRA loading, shuffle pattern, and rod group designations 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 maximum stuck rod worth for cycle 5 is less than that for the design cycle 4 at BOC and EOC. All safety criteria associated with these worths are met. The adequacy of the shutdown margin with cycle 5 stuck rod worths is demonstrated in Table 5-2. The following conservatisms were applied for the shutdown calculations:

1. Poison material depletion allowance.

2. 107. 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 mar-gin is presented in the ANO-1 cycle 4 reload report.3 5.2. Analytical Input The cycle 5 incore measurement calculation constants to be used for computing core power distributions were prepared in the same manner as those for the reference cycle.

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I 5.3. Changes in Nuclear Design Ill There are no significant core design changes between the reference and reload cycles. The calculational methods and design information used to obtain the important nuclear design parameters for this cycle were the same as those used for the reference cycle. The only significant operational change from the reference cycle is the withdrawal of the APSRs during the last 27 EFPD of cycle 5. The stability and control of the core in this mode have been ana-lyzed; the calculated stability index without APSRs in --0.0407 h-1, which demonstrates the axial stability of the core. The operating limits (Techni-cal Specification changes) for the reload cycle are shown in section 8 I

Table 5-1. Physics Parameters for ANO-1, Cycles 4 and 5(#)

Cycle 4 Cycle 5(C)

Cycle length, EFPD 387 387 Cycle burnup, mwd /mtU 12,111 12,129 Avg core burnup, EOC, mwd /mtU 20,505 22,069 Initial core loading, mtU 82.1 82.0 Critical boron - BOC, ppm (no Xe)

HZP(d), gp 8 ins. 1562 1508 l HZP, gps 7 and 8 ins. 1458 1404 HFP, gps 7 and 8 ins. 1246 1179 Critical boron - EOC, ppm HZP, gp 8 100% wd, eq Xe 418 408 HFP, gp 8 100% wd, eq Xe 86 87 Control rod worths - HFP, BOC, % Ak/k Group 6 1.18 1.068 Group 7 1.02 0.98 Group 8 0.37 0.45 l Control rod worths - HFP, 300 EFPD, % Ak/k Group 7 1.00 1.227 I Group 8 0.47 0.48 l Max ejected rod worth - HZP, % Ak/k(*} )

BOC (N-12), group 8 ins. 0.76 0.49 '

360 EFPD (N-12), group 8 out 0.82 0.50 Max stuck rod worth - HZP, % Ak/k BOC (N-12) 1.92 1.33 360 EFPD (H-14) 1.86 1.39 Babcock & Wilcox

Table 5-1. (Cont'd)

Power deficit, HZP to HFP, % Ak/k BOC 1.38 1.36 EOC 2.28 2.29 Doppler coeff - 100% pwr, 10-5(Ak/k/*F)

BOC, no xenon -1.57 -1.52 EOC, equil xenon -1.71 -1.73 Moderator coeff - HFP, 10 "(Ak/k/*F)

BOC (0 Xe, crit ppm, gp 8 ins.) -0.48 -0.62 EOC (eq Xe, 0 ppm, gp 8 out) -2.78 -2.83 Boron worth - HFP, ppm /% Ak/k BOC 118 122 EOC 105 106 Xenon worth - HFP, % Ak/k 2.59 2.58 I BOC (4 EFPD)

EOC (equilibrium)

Effective delayed neutron fraction (HFP) 2.75 2.67 0.00622 I BOC EOC 0.00617 0.00517 0.00520 I (* Cycle 5 data are for the conditions stated in this report.

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The cycle 4 core conditions are identified in reference 2.

Based on 294 EFPD at 2568 MWt, cycle 3.

( Based on a projected cycle 4 lifetime of 387 EFPD.

( )HZP: hot zero power (532F T 8), HFP: hot full power (579F T g).

(*) Ejected rod worth for groups 5 through 7 inserted, group 8 as stated.

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I' Table 5-2. Shutdown Margin Calculations for ANO-1, Cycle 5 BOC, 300 EFPD, 360 Fr/D,

% Ak/k  % Ak/k  % Ak/k Available Rod Wcrth Total rod worth, HZP 9.16 9.44 9.10 I' 1

Worth reduction due to -0.42 -0.42 -0.42 poison material bunup Maximum stuck rod, HZP -1.33 -2.17 -1.39 Net worth 7.41 6.85 7.29 Less 10% uncertainty -0.74 -0.69 -0.73 Total available worth 6.67 6.16 6.56 Required Rod Worth I

Power deficit, HFP to HZP 1.36 2.32 2.29 Allowable inserted rod 1.19 1. 4 ~/ 0.35 worth Flux redistribution 0.66 1.16 1.19 Total required worth 3.21 4.95 3.83 Shutdown margin (total 3.46 1.21 2.73 available worth minus total required worth)

Note: The required shutdown margin is 1.00% Ak/k.

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I Figure 5-1. ANO-1 Cycle 5 BOC (4 EFPD) Two-Dimensional Relative Power Distribution - Full Power, Equ11brium Xenon, Normal Rod Positions 8 9 10 11 12 13 14 15 1.12 1.27 1,17 1.30 1.16 1.17 0.5 0.58 I K 1.27 1.24 1.18 1.27 1.22 1.01 0.52 I

t a N 0.70 1 .16 1.10N 1 .27 0.97 0.41 I M 1.07 1.26 1.13 0.71 I

1.23 1.02 0.47 I N 0.59 0

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I Inserted Rod Group No. l I XXX Relative Power Density

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-N I 6. THERMAL-HYDRAULIC DESIGN h fresh batch 7 fuel ir hydraulically and geometrically similar to the previo: sly irradiated batch 53 and 6 fuel. The four batch 7 LTAs have been analyzed to ensure that they are never the limiting assembl'es during cycle 5 gperation.2 The thermal-hydraulic evaluation of cycle 5 incorporated the methods and models described in references 1, 3, and 8. The cycle 5 nuclear design al-lowed a reduction of the design radial-local peak from 1.78 to 1.71. As a re-sult of this peakir.g reduction, the steady-state design overpower minimum DNBR increased from 1.85 te 2.05. Table 6-1 summarizes the cycle 4 and 5 maximum design conditions.

A rod bow DNBR penalty has been calculated for cycle 5 according to procedures approved by reference 9. The burnup used to calculate the penalty is conserva-tively chosen as the highest assembly burnup in the cycle 5 core of 36,213 mwd /mtU. The resultant rod bow penalty is 4.0%, which includes 1% credit for use of a flow area reduction factor in the DNBR analyses.

A flux / flow setpoint of 1.07 has been established for cycle 5 operation. This setpoint and other plant operating limits based on DNBR criteria include a min-imum of 10% DNBR margin to offset the impact of any rod bow penalty.

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I Table 6-1. Maximum Design Conditions, Cycles 4 and 5 Cycle 4 Cycle 5 Design power level, MWt 2568 2568 System pressure, psia 2200 2200 Reactor coolant flow, % design 106.5 106.5 Vessel inlet / outlet coolant temp 555.6/602.4 555.6/602.4 at 100% power, F Reference design radial-local 1.78 1.71 power peaking factor Refe nee design axial flux 1.5 cosine 1.5 cosine shape Hot channel factors Enthalpy rise 1.011 1.011 Heat flux 1.014 1.014 Flow area 0.98 0.98

/etive fuel length, in. 140.2 140.2 Asg heat flux a 100% power, 175 175 3

1C Btu /h-ft2(a Max heat flux at 100% power, 468 449 l 3

10 Btu /h-ft2(b)

CHF correlation BAW-2 BAW-2 Minimum DNBR -

At 112% power 1.88 2.05 At 108% power 2.01 2.18 At 100% power 2.30 2.39

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

(

Based on average heat flux with reference. peaking.

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7. ACCIDENT AMD TRANSIENT ANALYSIS 7.1. General Safety Analysis Each FSAR1 accident analysis has been examined with respect to changes in cycle 5 parameters to determine the effect of the cycle 5 reload and to en-sure that thermal performance during hypothetical transients is not degraded.

The effects of fuel densification on the FSAR accident results have been eval-uated and are reported in reference 8.

I Since batch 7 reload fuel assemblies contain fuel tods whoac theoretical density is higher than those considered in the reference 3 report, the conclusions in that reference are still valid.

A study of the major FSAR Chapter 14 accidents using the cycle 5 iodine and noble gas inventories concluded that the thyroid and whole body doses are less than 4.1% of the 10 CFR 100 limits for all accidents except the MHA. For the MHA, the 2-hour dose to the thyroid at the exclusion area boundary increased by 2.7% to 157 Rem, which represents 52% of the 10 CFR 100 limits. The cor-responding 2-hour whole body dose for the MHA decreased by 29.1% to 7.09 Rem, which represents 28% of the 10 CFR 100 limits.

7.2. Accident Evaluation The key parameters that have the greatest effect on determining the outcome of a traneient can typically be classified in three major areas: core thermal parameters, thermal-hydraulic parameters, and kinetics parameters, including the reactivity feedback coefficients and control rod worths.

Core thermal properties used in the FSAR accident analysis were design oper-ating values based on calculational values plus uncertainties. First-core values (FSAR values) of core thermal parameters and subsequent fuel batches are compared to parameters used in cycle 5 analyses in Table 4-2. The cycle 5 thermal-hydraulic maximum design conditions are compared to the previous cycle 4 values in Table 6-1. These parameters are common to all the accidents l considered in this report. The key kinetics parameters from the FSAR and cycle 5 are compared in Table 7-2.

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l A generic LOCA analysis for a ESW 177-FA, lowered-loop NSS has been performed using the Final Acceptance Criteria ECCS Evaluation Model (reported in BAW-10103)." This analysis is generic since the limiting values of key param- l eters for all plants in this category were used. Furthermore, the combination of average fuel temperatures as a function of LHR and lifetime pin pressure data used in the BAW-10103 LOCA limits analysis is conservative compared to those calculated for this reload. Thus, the analysis and the LOCA limits reported in BAW-10103 and substantiated by reference 11 provide conservative results for the operation of the reload cycle. Table 7-1 shows the bounding values for allowable LOCA peak LHRs for ANO-1 cycle 5 fuel.

It is concluded from the examination of cycle 5 core thermal and kinetics properties, with respect to acceptable previous cycle values, that this core reload will not adversely affect the ANO-1 plant's ability to operate safely during cycle 5. Considering the previously accepted design basis used in the su FSAR and subsequent cycles, the transient evaluation of cycle 5 is considered to be bounded by previously accepted analyses. The initial conditions for the transients in cycle 5 are bounded by the FSAR, the fuel densification re-port, and/or subsequent cycle analyses.

I Table 7-1. Bounding Values for Allowable LOCA Peak Linear Heat Rates Core Allowable elevation, peak LHR, ft kW/ft 2 15.5 4 16.6 6 18.0 8 17.0 10 16.0 I: l I

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I Table 7-2. Comparison of Key Parameters for Accidene Analysis FSAR and densification ANO-1 Parameter report value cycle 5 g

Doppler coeff (BOC), 10-5 Ak/k/*F -1.17 -1.52 Doppler coeff (EOC), 10-5 Ak/k/*F -1.30 -1.73 Moderator coeff (BOC), 10-" Ak/k/*F 0.0(*) -0.62 Moderator coeff (E0C), 10-" Ak/k/*F -4.0 -2.83 All-rod group worth (HZP), % Ak/k 12.9 9.16 I Initial boron concentration, ppm 1150 1179 I Boron reactivity worth (HFP),

ppm /% Ak/k 100 122 Max ejected rod worth (HFP), % Ak/k 0.65 0.54 Dropped rod worth (HFP), % Ak/k 0.65 0.20

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

(b) 3.0 x 10-" Ak/k/*F was used for the steam line failure analysis.

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I 8. PROPOSED MODIFICATIONS TO TECHNICAL SPECIFICATIONS 1

The Technical Specifications have been revised for cycle 5 operation to ac-count for changes in power peaking and control rod worths inherent with the transition to 18-month LBP fuel cycles. The cycle 5 flux / flow setpoint is increased from 1.057 to 1.07.

I This is the result of a flatter radial power distribution.

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. The following pages con-tain the revisions to previous Technical Specifications.

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I:1 2 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS, REACTOR CORE Applicability Applies to reactor thermal power, reactor power imbalance, reactor coolant system pressure, coolant temperature, and coolant flow during power operation of the plant.

Objective To maintain the integrity of the fuel cladding.

Specification 2.1.1 The combination of the reactor system pressure and coolant temperature shall not exceed the safety limit as defined by the locus of points es-tablished in Figure 2.1-1. If the actual pressure / temperature point is below and to the right of the pressure / temperature line, the safety limit is exceeded.

2.1.2 The combination of reactor thermal power and reactor power imbalance (power in the top half of the core minus power in the bottom half of g the core expressed as a percentage of the rated power) shall not exceed 3 the safety limit as defined by the locus of points for the specified flow set forth in Figure 2.1-2. If the actual-reactor-thermal-power /

reactor-power-imbalance point is above the line for the specified flow, the safety limit is exceeded.

Bases To maintain the integrity of the fuel cladding and to prevent fission product release, it is necessary to prevent overheating of the cladding under normal g operating conditions. This is accomplished by operating within the nucleate 5 boiling regime of heat transfer, wherein the heat transfer coefficient is large enough so that the cladding surface temperature is only slightly greater than the coolant temperature. The upper boundary of the nucleate boiling re-gime is termed departure from nucleate boiling (DNB). At this point there is a sharp reduction of the heat transfer coefficient, which would result in high cladding temperatures and the possibility of cladding failure. Although DNB is not an observable parameter during reactor operation, the observable param-eters of neutron power, reactor coolant flow, temperature, and pressure can be related to DNB through the use of the BAW-2 correlation. (1) The BAW-2 corre-lation has been developed to predict DNB and the location of DNB for axially uniform and nonuniform heat flux distributions. The local DNB ratio (DNBR),

defined as the ratio of the heat flux that would cause DNB at a particular core location to the actual heat flux, is indicative of the margin to PNB.

The minimum value of th: DNBR during steady-state operation, normal opera-tional transients, and anticipated transients is limited to 1.3. A DNBR of 1.3 corresponds to a 95 percent probabiJity at a 95 percent confidence g level that DNB will not occur; this is considered a conservative margin to W DNB for all operating conditions. The difference between the actual core outlet pressure and the indicated reactor coolant system pressure has been Babcock & Wilcox

I I considered in determining the core protection safety limits. The difference in these two pressures is nominally 45 psi; however, only a 30 psi drop was assumed in reducing the pressure trip setpoints to correspond to the ele-vated location where the pressure is actually measured.

I The curve presented in Figure 2.1-1 presents the conditions at which a mini-mum DNBR of 1.3 is predicted for the maximum possible thermal power (112 percent) when the reactor coolant flow is 131.3 x 106 lb/h, which is the design flow rate for four operating reactor coolant pumps. This curve is based on the following nuclear power peaking factors (2) with potential fuel densification effects:

F = 2.56; F = 1.71; F = 1.50.

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These design limit power 1aaking factors are the most restrictive calculated at full power for the rang from all control rods fully withdrawn to maximum allowable control rod insertion, and for the core DNBR design basis.

I The curves of Figure 2.1-2 are based on the more restrictive of two thermal limits and include the effects of potential fuel densification:

l. The 1.3 DNBR limit produced by a nuclear power peaking factor I' of FN = 2.56 or the combination of the radial peak, axial peak, and Ehe position of the axial peak that yic1ds no less than I

1.3 DNBR.

2. The combination of radial and axial peaks that prevents central fuel melting at the hot spot. The limit is 20.1 kW/ft.

Power peaking is not a directly observable quantity, and therefore, limits have been established on the basis of the reactor power imbalance produced by the power peaking.

The flow rates for curves 1, 2, and 3 of Figure 2.1-3 correspond to the ex- l pected minimum flow rates with four pumps, three pumps, and one pump in each loop, 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 Figure 2.1-3. The curves of Figure 2.1-3 represent the conditions at which a minimum DNBR of 1.3 is predicted at the mav.iaum possible thermal power for the number of I

reactor coolant pumps in operation or the local quality at the point of min-l imum DNBR is equal to 22 percent (3), whichever condition is more restrictive. l Using a local quality limit of 22 percent at the point of minimum DNBR as a basis for curve 3 of Figure 2.1-3 is a conservative criterion even though the quality at the exit is higher than the quality at the point of minimum DNBR.

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The DNBR as calculated by the BAW-2 correlation continually increases from the point of minimum DNBR, so that the exit DNBR is always higher and is a fuention of the pressure. g gi The maximum thermal power for three-pump operation is 86.42 percent due to a  ;

power level trip produced by the flux-flow ratio (74.7 percent flow x 1.07 =

79.92 percent power; plus the maximum calibration and instrumentation error.

The maximum thermal power for other reactor coolant pump conditions is pro-duced in a similar manner.

For each curve of Figure 2.1-3, a pressure-temperature point above and to the left of the curve would result in a DNBR greater than 1.3 or a local quality at the point of minimum DNBR less than 22 percent for that particular reactor g coolant pump situation. Curves 1 and 2 of Figure 2.1-3 are the most re- g strictive because any pressure / temperature point above and to the left of this curve will be above and to the left of the other curve.

REFERENCES (1) Correlation of Critical Heat Flux in a Bundle Cooled by Pressurized Water, BAW-10000A, May 1976.

(2) FSAR, Section 3.2.3.1.1.c.

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I 2.2 SAFETY LIMITS - REACTOR SYSTEM PRESSURE Applicability Applies to the limit on reactor coolant system pressure.

Objective I

To maintain the integrity of the reactor coolant system and to prevent the release of significant amounts of fission product activity.

Specification 2.2.1 The reactor coolant system pressure shall not exceed 2750 psig when there are fuel assemblies in the reactor vessel.

2.2.2 The setpoint of the pressurizer code safety valves shall be in ac-cordance with the ASME Boiler and Pressure Vessel Code,Section III, Article 9, Summer 1968.

I The reactor coolant system (1) serves as a barrier to prevent radionuclides in the reactor coolant from reaching the atmosphere. In the event of a fuel cladding failure, the reactor coolant system is a barrier against the release I of fission products. Eastablishing a system pressure limit helps to assure the integrity of the reactor coolant system. The maximum transient pressure allowable in the reactor coolant system pressure vessel under the ASME Code,Section III, is 110 percent of design pressure.(2) The maximum transient pressure allowable in the reactor coolant system piping, valves, and fittings under ANSI Section B31.7 is 110 percent of design pressure. Thus, the safety limit of 2750 psig (110 percent of the 2500 psig design pressure) has been I es tablished. (2) The settings for the reactor high-pressure trip (2355 psig) and the pressurizer code safety valves (2500 tsig !1%)(3) have been estab-lished to ensure that the reactor coolant system pressure safety limit is not exceeded. The initial hydrostatic test is conducted at 3125 psig (125 per-I cent of design pressure) to verify the integrity of the reactor colant system.

Additional assurance that the reactor coolant system pressure does not exceed the safety limit is provided by setting the pressurizer electromatic relief valve at 2255 psig.(4)

REFERENCES (1) FSAR, Section 4.

(2) FSAR, Section 4.3.10.1.

(3) FSAR, Section 4.2.4.

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(4) FSAR, Table 4-1.

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I 2.3 LIMITING SAFETY SYSTEM SETTINGS, PROTECTIVE INSTRUMENTATION Applicability Applies to instruments monitoring reactor power, reactor power imbalance, re-actor coolant system pressure, reactor coolant outlet temperature, flow, num-ber of pumps in operation, and high reactor building pressure.

Objective To provide automatic protection action to prevent any combination of process variables from exceeding a safety limit.

I Specification 2.3.1 The reactor protection system trip setting limits and the permissible g bypasses for the instrument channels shall be as stated in Table 2.3-1 and Figure 2.3-2.

g Bases The reactor protection system consists of four instrument channels to monitor each of reveral selected plant conditions which will cause a reactor trip if 3 any one of these conditions deviates from a preselected operating range to the 3 degree that a safety limit may be reached.

The trip setting limits for protection system instrumentation are listed in Table 2.3-1. The safety analysis has been based on these protection system instrumentation trip setpoints plus calit stion and instrumentation errors.

Nuclear Overpower A reactor trip at high power level (neutron flux) is provided to prevent dam- g age to the fuci cladding from reactivity excursions too rapid to be detected 3 by pressure and temperature measurements.

During normal plant operation with all reactor coolant pumps operating, reac-tor trip is initiated when the reactor power level reaches 105.5 percent of rated power. Adding to this the possible variation in trip setpoints due to calibration and instrument errors, the maximum actual power at which a trip l would be actuated could be 112%, which is the value used in the safety W j analysis.

A. 07erpower Trip Based on Flow and Imbalance

'The power level trip setpoint produced by the reactor coolant system flow is based on a power-to-flow ratio which has been established to acco mnodate the most severe thermal transient considered in the de-sign, the loss-of-coolant-flow accident from high power. Analysis l

has demonstrated that the specified power-to-flow ratio is adequate l to prevent a DNBR of less than 1.3 should a low flow condition exist W due to any electrical malfunction.

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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 trip setpoint produced by the power-to-flow ratio provides overpower DNB protection for all modes of pump operation. Fo every flow rate there is a maximum permissible power level, and for every power level there is a minimum permissible icw flow rate. Typical power level and low flow rate combinations for the pump situations of Table 2.3-1 are as follows:

1. Trip would occur when four reactor coolant pumps are operating if power is 107 percent and reactor flow rate is 100 percent or flow rate is 93.45 percent and power level is 100 percent.

2. Trip would occur when three reactor coolant punps are operating if power is 79.92 percent and reactor flow rate is 74.7 percent or flow rate is 70.09 percent and power level is 75 percent.

3. Trip would occur when one reactor coolant pump is operating in I each loop (total of two pumps operating) if the power is 52.64 percent and reactor flow rate is 49.2 percent or flow rate is 45.79 percent and the power level is 49.0 percent.

The flux / flow ratios account for the maximum calibration and instrumentation errors and the maximum variation from the average value of the RC flow signal in such a manner that the reactor protective system receives a conservative indication of the RC flow.

No penalty in reactor coolant flow through the core was taken for an open core vent valve because of the core vent valve surveillance program during each refueling outage. For safety analysis calculations the maximum calibra-tion and instrumentation errors for the power level were used.

The power-imbalance boundaries are established in order to prevent reactor thermal limits from being exceeded. These thermal limits are either power peaking kW/ft limits or DNBR limits. The reactor power imbalance (power in top half of core minus power in bottom half of core) reduces the power level trip produced by the power-to-flow ratio so that the boundaries of Figure 2.3-2 are produced. The power-to-flow ratio reduces the power level trip I associated with reactor power-to-reactor power imbalance boundaries by 1.07 percent for a 1 percent flow reduction.

l B. Pump Monitors In conjunction with the power imbalance / flow trip, the pump monitors prevent the minimum core DNBR from decreasing below 1.3 by tripping the reactor due to the loss of reactor coolant pump (s). The pump monitors also restrict the power level for the number of pumps in operation.

I C. RCS Pressure I

During a startup accident from low power or a slow rod withdrawal from high power, the system high-pressure trip setpoint is reached before the nuclear overpower trip setpoint. The trip setting limit Babcock & \Milcox 8-7

I shown in Figure 2.3-1 for high RCS pressure (2300 psig) has been es-tablished to maintain the system pressure below the safety limit (2750 psig) for any design transient .(2)

The low-pressure (1800 psig) and variable low-preasure (11.75 Tout

- 5103) trip setpoints shown in Figure 2.3-1 have been established to maintain the DNB ratio greater than or equal to 1.3 for those design accidents that result in a pressure reduction.(2,3)

Due to the calibration and instrumentation errors, the safety analy-sis used a variable low reactor coolant system pressure trip value of (11.75 Tout - 5143).

D. Coolant Outlet Temperature The high reactor coolant outlet temperature trip setting limit (619F) E shown in Figure 2.3-1 has been established to prevent excessive core g coolant temperatures in the operating range. Due to calibration and instrumentation errors, the safety analysis used a trip setpoint of l 620F.

E. Reactor Building Pressure The high reactor building pressure trip setting limit (4 psig) pro-vides positive assurance that a reactor trip will occur in the un-likely event of a steam line failure in the reactor building or a loss-of-colant accident, even in the absence of a low reactor coolant system pressure trip.

F. Shutdown Bypass in order to provide for control rod drive tests, zero power physics testing, and startup procedures, there is provision for bypassiag cer- g tain segments of the reactor protection system. The reactor protec- g tion system segments that can be bypassed are shown in Table 2.3-1.

Two conditions are imposed when the bypass is used:

1. A nuclear overpower trip setpoint of $5.0 percent of rated power is automatically imposed during reactor shutdown.

2. A high reactor coolant system pressure trip setpoint of 1720 psig is automatically imposed.

The purpose of the 1720 psig high-pressure trip setpoint is to prevent normal operation with part of the reactor protection system bypassed.

This high-pressure trip setpoint is lower than the normal low-pressure trip setpoint so that the reactor must be tripped before the bypass is initiated. The overpower trip setpoint of $5.0 prevents any signifi-cant reactor power from being produced when performing the physics tests. Sufficient natural circulation (5) would be available to remove 5.0 percent of rated power if none of the reactor coolant pumps were operating.

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I REFERENCES (1) FSAR, Section 14.1.2.3.

(2) FSAR, Section 14.1.2.2.

(3) FSAR, Section 14.1.2.7.

(4) FSAR, Section 14.1.2.8.

(5) FSAR, Section 14.1.2.6.

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I' 3.2 MAKEUP AND CHEMICAL ADDITION SYSTEMS Applicability Applies to the operational status of the makeup and the chemical addition systems.

Objective To provide for adequate boration under all operating conditions to assure ability to bring the reactor to a cold shutdown condition.

Specification 3.2.1 The reactor shall not be heated or maintained above ROF unless the following conditions are met:

3.2.1.1 Two makeup pumps are operable except as specified in Specification 3.3.

3.2.1.2 A source of concentrated boric acid solution in addition to that in the borated water storage tank is available and operable. This re-quirement is fulfilled by the boric acid addition tank. This tank shall contain at least the equivalent of the boric acid volume and concentration requirements of Figure 3.2-1 as boric acid solution with a temperature of at least 10F above the crystallization temper-ature. System piping and valves necessary to establish a flow path from the tank to the makeup system shall also be operable and shall have at least the same temperature as the boric acid addition tank.

One associated boric acid pump is operable.

3.2.1.3 The boric acid addition tank and associated piping, valves, and both pumps may be out of service for a maximum of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. After the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, if the system is not returned to service and operable, the reactor shall be brought to the hot shutdown condition within an additional 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Bases The makeup and chemical addition systems provide control of the reactor cool- g ant system boron concentration.(1) This is normally accomplished by using any g of the three makeup pumps in series with a boric acid pump associated with the boric acid addition tank. The alternate method of boration will be the use of the makeup pumps taking suction directly from the borated water storage tank.(2)

The quantity of boric acid in storage from either of the two above-mentioned sources is sufficient to borate the reactor coolant system to a 1% suberitical margin in the cold condition (200F) at the worst time in core life with a stuck control rod assembly and after xenon decay.

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Minimum volumes (including a 10% safety factor) as specified by Figure 3.2-1 for the boric acid addition tank or 44,549 gallons of 2270 ppm boron as borie l

acid solution in the borated water storage tank (3) will each satisfy this re-quirement. The specification ensures that adequate supplies are available 8-10 Babcock & Wilcox

I whenever the reactor is heated above 200F so that a single failure will not prevent boration to a cold condition. The minimum volumes of boric acid solu-tion given include the boron necessary to account for xenon decay.

The pricipal method of adding boron to the primary system is to pump the concentrated boric acid solution (9500 ppm boron, minimum) into the makeup tank using the 25 gpm boric acid pumps.

The alternate method of addition is to inject boric acid from the borated water storage tank using the makeup pumps.

Concentration of boron in the boric acid addition tank may be higher than the concentration that would crystallize at ambient conditions. For this reason and to ensure that a flow of boric acid is available when needed, this tank and its associated piping will be kept 10'F above the crystallization temper-ature for the concentration present. Once in the makeup system, the concen-trate is sufficiently well mixed and diluted so that normal system tempera-tures ensure boric acid solubility.

REFERENCES I (1) FSAR, Sections 9.1, 9.2.

(2) FSAR, Figure 6-2.

(3) FSAR, Section 3.3.

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I 3.5.2 Control Rod Group and Power Distribution Limits Applicability ,

This specification applies to power distribution and operation of control rodo,during power operation.

Objective To ensure an acceptable core power distribution during power operation, to set a limit on potential reactivity insertion from a hypothetical control rod ejection, and to ensure core suberiticclity after a reactor trip.

Specification 3.5.2.1 The available shutdown margin shall be not less than 1% Ak/k with the highest worth control rod fully withdrawn.

3.5.2.2 Operation With Inoperable Rods:

1. Operation with more than one inoperable rod, as defined in Speci- l fications 4.7.1 and 4.7.2.3, in the safety or regulating rod W groups shall not be permitted.

2. If a control rod in the regulating or safety rod groups is de-clared inoperable in the withdrawn position as defined in Speci-fications 4.7.1.1 and 4.7.1.3, an evaluation shall be initiated immediately to verify the existence of 1% Ak/k available shutdown margin. Boration may be initiated either to the worth of the inoperable rod or until the regulating and transient rod groups are withdrawn to the limits of Specification 3.5.2.5.3, whichever occurs first. Simultaneously, a program of exercising the re-maining regulating and safety rods shall be intiated to verify operability.

3. If within one hour of determination of an inoperable rod as de-fined in Specifiation 4.7.1, 't is not determined that a 1% Ak/k available shutdown margin exists combining the worth of the in-operable rod with each of the other rods, the reactor shall be brought to the hot standby condition until this margin is estab-lished.

4. Following the determination of an inoperable rod as defined in Specification 4.7.1, all remaining rods shall Le exercised within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and exercised weekly until the rod problem is solved.

5. If a control rod in the regulating or safety rod groups is de-clared inoperable per 4.7.1.2, power shall be reduced to 60% of the thermal power allowable for the reactor coolant pump combi-nation.

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6. If a control rod in the regulating or axial power shaping groups is declared inoperable per Specification 4.7.1.2, operation above 60% of the thermal power allowable for the reactor coolant pump combination may continue provided tha rods in the group are positioned such that the rod that was declared inoperable is con-tained within allowable group average position limits of Speci-fication 4.7.1.2 and the withdrawal limits of Specification 3.5.2.5.3.

3.5.2.3 The worth of single inserted control rods during criticality is lim-ited by the restrictions of Specification 3.1.3.5 and the Control Rod Position Limits defined in specification 3.5.2.5.

3.5.2.4 Quadrant Tilt:

Except for physics tests, if quadrant tilt exceeds 4.92%, power I 1.

shall be reduced immediately to below the power level cutof f (see Figures 3.5.2-1A and 3.5.2-1B). Moreover, the power level cutoff value shall be reduced 2% for each 1% of tilt in excess of 4.92%

tilt. For less than four-pump operation, thermal power shall bc reduced 2% of the thermal power allowable for the reactor coolant pump combination for each 1% tilt in excess of 4.92%.

2. Within a period of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, the quadrant power tilt shall be re-duced to less than 4.92% except for physics tests, or the follow-ing adjustments in setpoints and limits shall be made:

a. The protection system maximum allowable setpoints (Figure 2.3-2) shall be reduced 2% in power for each 1% tilt.

b. The control rod group and APSR withdrawal limits shall be reduced 2% in power for each 1% tilt in excess of 4.92%.

c. The operational imbalance limits shall be reduced 2% in power for each 1% tilt in excess of 4.92%.

3. If quadrant tilt is in exess of 25%, except for physics tests or diagnostic testing, the reactor will be placed in the hot shutdown condition. Diagnostic testing during power operation with a quad-I rant power tilt is permitted provided the thermal power allowable for the reactor coolant pump combination is restricted as stated in 3.5.2.4.I above.

4. Quadrant tilt shall be monitored on a minimum frequency of once every 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> during power operation above 15% of rated power. i 3.5.2.5 Control Rod Positions:

1. Technical Specification 3.1.3.5 (Safety Rod Withdrawal) does not I prohibit the exercising of individual safety rods as required by Tabic 4.1-2 or apply to inoperable safety rod limits in Technical Specification 3.5.2.2.

2. Operating rod group overlap shall be 20% 5 between two sequen-tial groups, except for physics tests.

Babcock & Wilcox I 8-13

I

3. Except for physics tests or exercising control rods, (a) the control rod withdrawal limits are specified on Figures 3.5.2-1A, 3.5.2-1B, and 3.5.2-IC for four-pump operation, on Figures 3.5.2-2A, 3.5.2-2B, and 3.5.2-2C for three-pump operation, and on Figures 3.5.2-2D, 3.5. 2-2E, and 3.5.2-2F for two-pump opera-tion; and (b) the axial power shaping control rod withdrawal limits are specified on Figures 3.5.2-4A, 3.5.2-4B, and 3.5.2-4C.

If any of these control rod position limits are exceeded, correc-tive measures shall be taken immediately to achieve an acceptable control rod position. Acceptable control rod positions shall be l attained within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

4. Except for physics tests, powe. shall not be increased above the power level cutoff (92% full ,,ower) unless the xenon reac-tivity is within 10% of the equilibrium value for operation at rated power and asymptotically approaching stability.

3.5.2.6 Reactor Power Imbalance shall be monitored on a frequency not to ex-ceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> during power operation above 40% rated power. Except for physics tests, imbalance shall be maintained within the envelopes defined by Figures 3.5.2-3A 3.5.2-3B, and 3.5.2-3C. If the imbal-ance is not within the envelopes defined by Figures 3.5.2-3A, 3.5.2-3B, and 3.5.2-3C, corrective measures shall be taken to achieve an acceptable imbalance. If an acceptable imbalance is not achieved within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, reactor power shall be reduced until imbalance limits are met.

3.5.2.7 The control rod drive patch panels shall be locked at all times with limited access to be authorized by the superintendent.

Bases The poucr-imbalance envelopes defined in Figures 3.5.2-3A, 3.5.2-3B, and 3.5.2-3C are based on (1) LOCA analyses which have defined the maximum linear heat rate (see Figure 3.5.2-4) such that the maximum cladding temperature will not exceed the Final Acceptance Criteria and (2) the Protective System Maximum Allowable Setpoints (Figure 2.3-2). Corrective measures will be taken immediately should the indicated quadrant tilt, rod position, or imbalance be outside their specified boundaries. Operation in a sisoation that would cause the Final Acceptance Criteria to be approached should a LOCA occur is highly improbable because all of the power distribution parameters (quadrant tilt, rod position, and imbalance) must be at their limits while simultaneously all other engineering and uncertainty factors are also at their limits.* Con-servatism is introduced by application of:

a. Nuclear ut. certainty factors.

b. Thermal calibration.

c. Fuel densification effects.

• Actual operating limits depend on whether or not incore or excore detectors are used and their respective instrument and calibracion errors. The method used to define the operating limits is defined in plant operating procedures.

Babcock & Wilcox

i 1

_ '...+~-..f ; .?. 1. _l.. .- -.

i

d. Hot rod manufacturing tolerance factors.

e. Fuel rod bowing.

The 20 25% overlap between successive control rod groups is allowed since the worth of a rod is lower at the upper and lower parts of the stroke. Control rods are arranged in groups or banks defined as follows:

Group Function 1 Safety 2 Safety 3 Safety 4 Safety 5 Regulating 6 Regulating 7 Xenon transient override 8 APSR (axial power shaping bank)

The rod position limits are based on the most limiting of the following F ree criteria: ECCS power peaking, shutdown margin, and potential ejected rod worth. As discussed above, compliance with the ECCS power peaking criterion is ensured by the rod position limits. The minimum available rod worth, con-sistent with the rod position limits, provides for achieving hot shutdown by reactor trip at any time, assuming the highest worth control rod that is wit-drawn remains in the full-out position (1). The rod position limits also ensure that inserted rod groups will not contain single rod worths greater than 0.65% Ak/k at rated power. These values have been shown to be safe by the safety analysis (2) of the hypothetical rod ejection accident. A maximum single inserted control rod worth of 1.0% Ak/k is allowed by the rod position limits at hot zero power. A single inserted control rod worth of 1.0% Ak/k at beginning of life, hot zero power, would result in a lower transient peak thermal power and therefore less severe environmental consequences than a 0.65% Ak/k ejected rod worth at rated power.

Control rod groups are withdrawn in sequence beginning with group 1. Groups I 5, 6, and 7 are overlapped 20%. The normal position at power is for groups 6 and 7 to be partially inserted.

The quadrant power tilt limits set forth in Specification 3.5.2.4 have been I established within the thermal analysis design base using the definition of quadrant power tilt given in Technical Specifications, Section 1.6. These limits, in conjunction with the control rod position limits in Specification I 3.5.2.5.3, ensure that design peak heat rate criteria are not exceeded during normal operation when including the effects of potential fuel densification.

The quadrant tilt and axial imbalance monitoring in Specifications 3.5.2.4.4 and 3.5.2.6, respectively, will normally be perforind in the plant computer.

The 2-hour frequency for monitoring these quantities will provide adequate surveillance when the computer is out of service.

i During the physics testing progrin, the high flux trip setpoints are administra-tively set as follows to ensure that an additional safety margin is provided:

8-15 Babcock 8.Wilcox

Table 8-1. Reactor Protection System Trip Setting Limits (Specifictions Table 2.3-1)

Four RC pumps operating Three RC pumps operat'ing One RC pump operating in (nominal operating (nominal operating each loop (nominal Shutdown power, 100%) power, 75%) operating power, 49%) bypion Nuclear power, % of 105.5 105.5 105.5 5.0*

rated, max Nuclear power based on 1.07 times flow minus 1.07 times flow minus 1.07 times flow minus Bypassed flowb and imbalaace, % reduction due to im- reduction due to im- reduction due to im-of rated, max balance (s) balance (s) balance (s)

Nuclear power based on NA NA 55 Bypassed pump monitors, I of rated, max" High RC system pressure, 2300 2300 2300 1720*

psig, max co

,L Low RC system pressure, 1800 1800 1800 Bypassed

• psig, min d d Variable low RC system 11.75 T "

- 5103 11.75 T - 5103 11.75 T - 5103 Bypassed pressure, psig, min RC temp, F, max 619 619 619 619 High reactor building 4(18.7 psia) 4(18.7 psia) 4(18.7 psia) 4(18.7 psia) pressure, psig, max

• Automatically set when other segments of the RPS (as specified) are bypassed.

[ Reactor coolant system flow.

The pump monitors also produce a trip on (a) loss of two RC pumps in one RC loop, and (b) loss of one or two RC pumps during

$ two-pump operation.

Y d p T is given in degrees Fahrenheit (F).

I

=

O

1. 4 M

E _

E E E E E E E

I Figure 8-1. Core Protection Safety (Tech Spec Figure 2.1-1)

Limits I 2600 2400 lI E

a i .- 2200 r

'I t a

E ACCEPTABLE OPERATION l .5 5 2000

, o

!h "

S O UNACCEPTABLE OPERATION 1800 e r

, 1600 ll 560 580 600 620 Reactor Outlet Temperature, *F 640 660

I i

r l Babcock & Wilcox 8-17 j

Figure 8-2. Core Protection Safety Limits (Tech Spec Figure 2.1-2)

POWER $R 5

- 120

(-35.84,112) (21.28,112) _ ,g l M

i - 1.31 ACCEPTABLE

- 110 4 PUMP I l

(-45,100) l OPERATION -

- 100 l (35,100) i l

l 86.42

- 90

'. i l

I i I

' l ACCEPTABLE 3 & 4 PUMP

- 80 l l

l l

(-45,74.42) I OPERATION l (35,74.42) g

- 70 l l l  !

! 59.14 _ - 60 l ACCEPTABLE l 8

I 2,3 & 4 PUMP I

- 50 E

(-43,47.14) I OPER&fl0N I (35,47.14) g i l l l 1

- 40 l l UNACCEPTABLE I g

I

- 30 I l UNACCEPTABLE OPERATION l I OPERATION l - -

0

%l l 91I 41 u u l

- 10 gll i

u al u I I

E El E EI l i i a i i e i i __

i a i l i i I 3

-60 -50 -40 .10 -20 -10 0 10 20 30 40 50 60 i Reartar Power t;analance, 5 I

I I

8-18 Babcock & Wilcox

I Figure 8-3. Core Protection Safety Limits (Tech Spec Figure 2.1-3)

I 2600 I

2400

'I i 5 -2200 I =

! O'2

=

lI { 2000

/

3 ll 1800 7

I 560 580 600 620 640 660 Reactor Outlet Temperature, *F CURVE GPM POWER PUMPS OPERATING (TYPE OF LIMIT) 1 374,880 (100",)* 112% FOUR PUMPS (ON8R LIMIT) ,

l 2 3

280,035 (74.7%)

184,441 (49.2%)

86.7%

59.0%

THREE PUMPS (DN8R LIMIT)

ONE PUMP IN EACH LOOP (QUALITY LIMIT)

• 106.5% OF DESIGN FLOW I

I I .

I a cock & Elcox 8-19

I Figure 8-4. Protective System Maximum A11ovable Setpoints (Tech Spec Figure 2.3-1) 2500 P e 2300 PSIG T = 619*F 2300

.{" ACCEPTABLE OPERATION i

2100

/ l

[ P = 11.75 TOUT 3O 5103 PSIG

- 1900

/

UhACCEPTABLE E CPERATION l

P = 1800 PSIG 1700 I

580 600 620 640 660 560 Reactor Outlet Temperature, F l

I I

I I

e.ececx . w,icex l

I Figure B-5. Protective System Maximum Allowable Setpoints (Tech Spec Figure 2.3-2) l POWER $EA5 l

(-18,107) 107 110 (9,107)

.0 #CCEPI'"C k l

• Puwe l OPERATION g

-100l M2 = -1.0

(-34,91) l (26,90)

- 90 l

' 79.92

! 20 f ACCEPTA KE l l l I I

la s s I l Pipe -- 70 l  :

l

(-34,63.92) l0PERATION 1

- 60 l l l l g l 52.64 ACCEPTABLE - - 50 I

i 2,3 a e I (-34,36.64) g i iPuwe l OPERATION -- 40 l I

I i l (26,35.64) l l

- 30 I l 1

I 1

I

[

I I -

- 20 i i i j uaccEemu DuccEPmu . _

OPERATION u I I I

- 10 l u

'* El u

OPERATION u i l

i i a:

e ii il i l i n

li i i i

-60 -50 -40 -30 -20 -10 0 10 20 30 40 50 60 Power imoalance, %

I I

lI l# 8-21 Babcock 8. Wilcox l

I:l Figure 8-6. Boric Acid Addition Tank Volume and Concentration Requirements Vs RCS Average Temperature (Tech Spec Figure 3.2-1)

OPERATION AB0VE AND TO 3 (579F,6375 GAL) o E THE LEFT OF THE CURVES 6000 I

IS ACCEPTABLE

=

l i

(579F, 5007 GAL) g 5000 _

5 I i I l 5 4000 -

(532F,4065 GAL) /

- / a g (500F,3582 GAL) / 5 (532F,3190 GAL)

[

= -

3000

[ /(500F,2811 GAL)

E /

12,000 PPM 5 2000 -

a BORIC ACIO 9500 PPM 400F,1657 GAL) 1000 -

BORIC ACID

/

[ g (300F,920 GA[L (300F,722 GAL) 5 0

( 00f,0 GAL) (200F; O W , ,

0 100 200 300 400 500 600 RCS Average Temperature, F l, I

I II I

8-22 Babcock & Wilcox I

I Figure 8-7. Rod Position Limits for Four-Pump Operation J From 0 to 300 1 10 EFPD - ANO-1, Cycle 5

3 (Tech Spec Figure 3.5.2-1A)

g 110 - , -

100

, o . 102)

OPERATION IN THIS REGION IS NOT ALLOWED q (206.7,92)

RESTRICTiD

}

SHUT 00WN REGION 80 -

(240.0,80)

MARGIN l _

E 70 -

LIMIT

=

60 i 50 -

(100,50) (300,50) >

i

% PERMISSIBLE i o 40 -

' OPERATING REGION 30 _

l 20 -

(43,15) 10 d (0,11)

I m t , , , , , , , i i , ,

j O + .

4 0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 Rod index, % WO O 20 40 60 80 100 e i i t i  !

Group 7 j 0 20 40 60 80 100 i f I f f i j Group 6 0 20 40 60 80 100

!' f I t  ! f I Group 5 l

lI I

ll I

8-23 Babcock & LVilcox

I Figure 8-8. Rod Position Limits for Four-Pump Operation From 300 i 10 to 360 ! 10 EFPD ANO-1, Cycle 5 (Tech Spec Figure 3.5.2-1B)

=

110 (270.102) ,

100 _

JPERATION iN 90 - THIS REGION ;S (261.92)

NOT ALLOWED 80~

(251,80)

S!!UT00WN RESTRICTED 70 g MARGIN REGION

- LIMIT PERMISSIBLE 60

- OPERATING (100,50) (200,50)

I

,- 50 REGION

=

E 40 30 -

20

- (43,15) 10 t (0,11)

' ' ' ' ' ' ' ' ' ' ' ' i i 0'

(0.0) 0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 Roa Index, t WO O 20 40 60 80 100 t t ' t 1 I Group 7 0 20 40 60 80 100 f I I t  !  !

Group 6 0 20 40 C. 80 100 t f  !  ! I f Group 5 I

I I

I 8-24 Babcock & Wilcox

I Il Rod Posit' ion Limit,i for Four-Pump Operation From

s Figure 8-9.

360 ! 10 to 387 10 EFPD - ANO-1, Cycle 5

}g (Tech Spec Figure 3.5.2-lC)

! g 110

(*bl.102) (274,102) l 100 - OPERATION IN THIS REGION IS NOT ALLOWE0 (267,92) I 90 -

! SHUTOOWN (240,80)

!g 80 MARGIN LlMIT RESTRICTED REGION ll g

= 70 -

n E

" 60 -

PERMLSSIBLE c OPERATING

1. (66,50) (175.50) REGION g 50 -

ig J l E l 2 40 il ,

20 (0,14) ( 5.15) l In 10

g ' ' ' ' ' ' ' ' ' ' ' ' '

0 0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 i (0,0)

Rod Inaex, % WO O 20 40 60 80 100 l t 1 i f f I

!'Em 0 20 40 60 80 t

100 I

' t I  ! t

! Group 6 0 20 40 60 -80 100 t i e i j Group 5 lI

\I

!I lI 8-25 Babcock & Wilcox

I Figure 8-10. Rod Position Limits for Three-Pump Operation From 0 to 300 2 10 EFPD - ANO-1, Cycle 5 (Tech Spec Figure 3.5.2-2A) 110 -

100 -

l 90 -

l GPERATION IN THIS REGION l (247.0,76.5) 80 ._ lS NOT ALLOWED (180.76.5)  !

RESTRICTED g

70 -

REGION g

=a h

60 -

SHUT 00NN PERMISSIBLE E

MARGIN LIMIT OPERATING g

'o REGION 50 -

(300.0.50) i g 40 -

(100.38.2) n.

30 20 -

10

_ (43.11.21 Da . > < > e < r i i , , t (0,0) 0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 Rod index, % WD 0 20 40 60 80 100 i I t  ! I I Group 7 0 2,0 40 60 80 100 E

t. i t i i g Group 6 y 20 40 6,0 80 1p0 l Group 5 l

I-)

I I

Il 8-26 Babcock & Wilcox

I Figure 8-11. Rod Position Limits for Three-Pump Operation Frem 300 t 10 to 360 2 10 EFPD - ANO-1. Cycle 5 (Tech Spec Figure 3.5.2-2B) )

110 100 90 80 - OPERATION IN THIS (160,76.5) (24S.1,76.5)

REGION IS NOT ALLOWED 3 70 -

g = PERMISSIBLE j N 60 -

SHUTOOWN MARGIN LIMIT OPERATING REGION g RESTRICTED e 50 - REGION

Ii t

(200.0,50) o 40 -

(100,38.2) 30 -

1 I

20 -

(43.11.2) 10

I:

J 0* ' ' ' ' ' ' ' ' ' ' ' ' ' '

) (0,0) 0 3 40 60 80 100 120 140 160 180 200 220 240 260 280 300 0 20 40- 60 80 100 Rod Index, 5 #D t t t f f a Group 7 0 20 40 60 80 100 t t I e f I

! Group 6 0 20 40 60 80 100 t I f I t l Group 5

I 4

.I I

8-27 Babcock & Wilcox l

I Figure 8-12. Rod Position Limits for Three-Pump Operation From 360 t 10 to 387 2 10 EFPD - ANO-1, Cycle 5 (Tech Spec Figure 3.5.2-2C) 100 -

90 -

OPERATION IN THIS REGION IS NOT ALLOWED 80 -

(151.76.5) (232.5.76.5) j 70 -

g SHUTDOWN MARGIN LIMli PERU SM 8d

" OPERATING

RESTRICTED REGION REGION a 50

. (175.50) 40 -

[ (66.38.2) 30 I

(5.11.2)l0 0- i ' ' ' ' ' t i i i i i i i (0,0) 0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 Rod Index, fi WO O 20 40 60 80 100 I I f I f l Group ?

O 20 40 60 80 100 i l I I t t Group 6 0 20 40 60 80 100 t t i  ! t j Group 5 I

I I

I 8-28 Babcock & Wilcox

g i

B Figure 8-13. Rod Position Limits for Two-Pump Operation From 0 to 300 10 EFPD - ANO-1, Cycle 5 (Tech Spec Figure 3.5.2-2D) 110 100 -

90 -

80 E

= 70 -

~

i .

N 60 -

l OPERATION IN THIS REGION (180,51) g 50 jg

iS NOT ALLOWE0 5 40 PERMISSIBLE li OPERATING REGION

! 30 ~

SHUTDOWN MARGIN LIMIT ' (100,25) lm

]g 20 -

l 10 (43,7.5) l ' ' , , , , , , , , , , ,

i 0 >

.0) 0 20 >> 60 80 100 120 140 160 180 200 220 240 260 280 300 Roa 1.dex, % WO O 20 40 60 80 100 I I i i f l

. Group 7 0 20 40 60 80 100 j

i i  ! i i i G 20 40 60 80 ica f , I f , ,

Group 5 l

{

8-29 Babcock & Wilcox

I Figure 8-14. Rod Position Limits for Two-Pump Operation From 300 3 10 to 360 10 EFPD - ANO-1, Cycle 5 (Tech Spec Figure 3.5.2-2E) 100 -

90 I

80 -

g 70 -

=

=

60 -

OPERATION IN THIS REGION IS N LLOWE0 (180,51) (201.7,51) 50 -

i g 40 ~

SHUT 00WN MARGIN LIMIT 96

' PERMISSIBLE OPERATING 30 REGION (100,25) 20 -

0

- (43,7.5) 0; i i i i e i , i i i i i , i

• ) 0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 40 8,0 y 100 0

Rod Index, 's WD ,

2,0 , 6,0 Group 7 0 20 40 60 80 100 t i i i l' i Group 6 0 20 40 60 80 100 m Li i i t i g Group 5 I

I I

I Babcock & Wilcox 8-30

!I

I Figure 8-15. Rod Position Limits for Two-Pump Operation From 360 i 10 to 387 2 10 EFPD - ANO-1, Cycic 5 1

(Tech Spec Figure 3.5.2-2F) '

' 110 100 -

90 -

- 80 -

I E 70 -

ll 8 N

l3  ;

60 -

OPERATION IN THIS REGION IS NOT ALLOWED j

(151.51) -- (177.2,51)

50 i

g6

~

SHUTDOWN MARGIN D

j LIMIT

' 30 -

PERMISSIBLE OPERATING REGION 1

(66.25) 20 -

l

!g 10 -

lg (5.7.5)

I i i , , , , , , , , , ,

Os e (0,0) 0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 Rao index, % WD 0 20 40 60 80 fr0 8 f f I t 3 Group 7 0 20 40 60 80 100 1

i e i i i i Group 6

.'E 0 20 40 60 80

!3 i i i i i 1,00 4

Group 5

!I lI 8-31 Babcock & Wilcox

I Figure 8-16. Operational Power Imbalance Envelope for Operation From 0 to 300 2 10 EFPD - ANO-1, Cycle 5 (Tech Spec Figure 3.5.2-3A)

--110

(-14.28,102) b -

-100

(-13.79,92)i > -

- 90 (1 .6,92) I

(.18.4,80) 4 -

- 80 > (24.0,80)

I PERMISSIBLE OPERATING REGION

- 60 RESTRICTED RESTRICTED g

- 50 5 REGION h- REGION

'o

- 40 l

o E

- 30 l

J'

- 20

- 10 i 1 1 1 t t t 1 t I I

-50 -40 -30 -20 -10 0 10 20 30 40 50 Axial Power imDalance, %

I I

I I

8-32 Babcock & Wilcox

Figure 8-17. Operational Power Imbalance Envelope for Operation From 300 2 10 to 360 ! 10 EFPD - ANO-1 Cycle 5 I (Tech Spec Figure 3.5.2-3B)

-- 110

(-17.0,102)  ; _

_ ggg

- (15.3,102)

(-17.48,92) '

90 (16.6,92)

I

(-22.4,80)4 -

- 80 > (24.0,80)

I PERillSSIBLE OPERATING i

REGION l

- 60 RESTRICTED REGION RESTRICTED -

- 50 REGION g -- 40 m

I N

'o - 30 e

- 20 l -

- 10 I - 50 -40 -30 -20 -10 0 10 20 30 40 50 Axial Power Imbalance, 5 I

I 8-33 Babcock & Wilcox (

l

I Figure 8-18. Operational Power Imbalance Em clope for Operation From 360 2 10 to 387 1 10 EFPD - ANO-1, Cycle 5 (Tech Spec Figure 3.5.2-3C)

-- 110

(-20.0,102) ' '

- 100

(-20.0,92) 0 '

- 90

(-24.0,80) 4 -

- 80 > ( 24. 0,80)

I PERMISSIBLE OPERATING REGION

- 60 RESTRICTE0 E REGION -

- 50 RESTRICTE0 5 REGION G

" -- 40 m

W m

-- 30 a

l f- - 20 a.

- 10 t i i t i l i I I f

-50 30 -20 -10 0 10 20 30 40 50 Axial Power Imbalance, 5 I

I 8-34 Babcock & Wilcox l _- _ _ _ . . -

l I

I Figure 8-19. APSR Position Limits for Operation From 0 to 300 2 10 EFPD - ANO-1, Cycle 5 (Tech Spec Figure 3.5.2-4A) 110 I 100 -

(8.0,102)

' ? " ' '

RESTRICTED

.9 a ( 6.0,92) REGION 90 -

80 < (0.0,80) (61.2,80)

=

I =

=

8 70 -

PERMISSIBLE OPERATING m

l _

O es**

60 -

REGION l g (100.0,50) '

50 40 -

30 -

20 -

10 I O 0

i i i , , , , , ,

10 20 30 40 50 60 70 80 90 100 APSR, % Witndrawn I

lI I

g e.35 eeececx , wiice,

I l Figure 8-20. APSR Position Limits for Operation From 300

! 10 to 360 i 10 EFPD - ANO-1, Cycle 5 (Tech Spec Figure 3.5.2-4B) 110 I

(6.0,102) (41.7,102) 100 -

i(6.0,92) I (41.7.92) RESTRICTE0 0

REGION 80 4 (0.0,80) (61.2,80) l

70 -

PERMISSIBLE

", OPERATING

$ 60 REGION (100.0,60) h O

a 50 -

N 5 40

~

30 -

I 20 -

10 -

0 0 10 20 30 40 50 60 70 80 90 100 APSR, % WItndrawn I

I I'

I, 8-36 Babcock & Wilcox

I

. Figure 8-21. APSR Position Limits for Operation From 360 l  ! 10 to 387 ! 10 EFPD - ANO-1, Cycle 5 13 (Tech Spec Figure 3.5.2-4C)

'E:

110

! 100 i

l 90 -

l i

80 -

i 1 -

ll E 70 -

15 m

=

APSR INSERTION NOT ALLOWE0 i

60 -

• IN THIS TIME INTERVAL l

< a jl 50 -

l i t '

l E o

1 40 ll -

1 30

!I j 20 i

l 10 Q f f f r t t t t t l 0 10 20 30 40 50 60 10 80 90 100 APSR, '4 Witndrawn I

!I

!I ll

8-37 Babcock & Wilcox

l 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. 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 oegins. Acceptable

[ criterin state that the rod drop time from fully withdrawn to 75% 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 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.

9.1.2. RC Flow

-I RC 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 two-pump coastdown of RC flow from the tripping of the highest-flow RC pump in each loop from four RC pumps running will be measured at hot zero lI power conditions. The coastdown of RC flo'- oraus time will then be compared to the required RC flow versus time to @ca int whe' 2r acceptance is met.

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I 9.2. Zero Power Physics Tests 9.2.1. Critical Boron Concentration Criticality 19 obtained by deboration at a constant dilution rate. Once criticality is achieved, equilibrium boron is obtained and the critical boron concentration determined. The critical boron concentration is calculated by 3 correcting for any rod withdrawal required to achieve equilibrium boron. The acceptance critarion placed on critical boron concentration is that the ac-tual boron concentraiton must be within !100 ppm boron of the predicted value.

9.2.2. Temperature Reactivity Coefficient The isothermal temperature coefficient is measured at approximately the all-rods-out configuration and at the hot zero power rod insertion limit. The 3 average coolant temperature is varied by first decreasing and then increasing temperature by 5*F. During the change in temperature, reactivity feedback is compensated by discrete changes in rod motion; the change in reactivity is then calculated by the summation of reactivity (obtained from reactivity cal-culation 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 10.4 x 10-" (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 tempetature coefficient has been 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~" (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 condit.ons using the boron / rod swap method. This technique con-sists of establishing a deboration rate in the reactor coolant system and compensating for the reactivity changes of this deboration by inserting con-trol 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 differential rod worths of each of the 9-2 Babcock & Wilcox

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1 controlling 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 5 15 measured value

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

I predicted value - measured value measured value x 100 $ 10 9.2.4. Ejected Control Rod Reactivity Worth After CRA groups 7, 6, and 5 have been positioned near the minimum rod inser-tion limit, the ejected rod is borated to 100% withdrawn and the worth ob-tained by adding the incremental changes in reactivity by boration.

After the ejected rod has been borated to 100% withdrawn and equilibrium boron established, the ejected rod is swapped in versus the controlling rod group, and the worth is determined by the change in the previously calibrated controlling rod group position. The boron swap and rod swap values are aver-aged and error-adjusted to determine ejected rod worth. Acceptance criteria I for the ejected rod worth test are as follows:

1. lpredictedvalue-measuredvalue measured value x 100l520

2. Meaoured value (error-adjusted) 5 1.0% Ak/k The predicted ejected rod worth is given in the Physics Test Manual.

9.3. Power Escalation Tests 9.3.1. Core Power Distribution Verification at 440, 75, and 100% FP With Nominal Control Rod Position Cure power distribution tests are performed at 40, 75, and 100% full power I (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.

Rad index is established at a nominal full-power rod configuration at which the core power distribution was calculated. APSR position is established to pro-vide a core power imbalszce corresponding to the imbalance at which the core powe- distribution calculations were performed.

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

1. The worst-case maximum LHR must be less than the LOCA limit.

2. The minimum DNBR must be greater than 1.30.

3. The value obtained from extrapolation of the minimum DNBR to the next power plateau overpower trip setpoint must be greater than 1.30, or the extrapolated value of imbalance must fall outside the RPS power / imbalance / flow trip envelope.

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

5. The quadrant power tilt shall not exceed the limits specified in the Technical Specifications.

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

predicted value - measured value 100 5 8.

measured value

7. The highest measured and predicted total peaks shall N within the following limits:

predicted value - measured value measured value x 100 5 12.

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

Items 3 and 4 establish the criteria whereby escalatl'n to the next power plateau may be accomplished without exceec~ing the safety limits specified by 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 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:

1. The highest measured and predicted radial peaks shall be within the following limits: l l

l predicted value - measured value x 100 55  !

measured value Babcock a.Wilcox

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

predicted value - measured value measured value x 100 5 7..

I 9.3.2. Incore Vs Excore Detector Imbalance Correlation Verificatioe at %40% FP I Imbalances, set up in the core by control rod positioning, are read simultan-eously on the incore detectors and excore power range detectors. The excore detector offset versus incore detector offset slope must be at least 1.15.

I If this criterion is not met, gain amplifiers on the excore detector signal processing equipment are adjusted to provide the required gain.

9.3.3. Temperature Reactivity Coefficient at %100% FP The average reactor coolant temperature is decreased and then increased by about 5*F at constant 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 temperature reactivity coefficient is calculated from the measured changes in reactivity and temperature. Acceptance criteria state that the moderator temperature coefficient shall be negative.

9.3.4. Power Doppler Reactivity Coefficient at %100% FP Reactor power is decreased and then increased oy about 5% FP. The reactivity change is obtained from the change in controlling rod group position. Con-I trol rod group worth is measured using the fast insert / withdraw method.

activity corrections are made for changes in xenon and reactor coolant tem-Re-perature 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. The predicted value of the power Doppler reactivity coefficient is given in the Physics Test Manual. I I Acceptance criteria state that the measured value shall be more negative than

-0.55 x IO " (Ak/k)/% FP.

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I 9.4. Procedure for Use if Acceptance Criteria Not Met If the acceptance criteria for any test are not met, an evaluation is per-formed before the test program is continued. This evaluation is performed by site test personnel with participation by Babcock & Wilcox technical per-sonnel as required. Further specific actions depend on the evaluation

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results. These actions can include repeating the tests with more detailed l attention to test prerequisites, added tests to search for anomalies, or design personnel perfor. ting detailed analyses of potential safety problems '

caused by parameter deviation. Power is not escalated until an evaluation ,

shows that plant safety will not be compromised by such escalation.

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I REFERENCES 1

I 1 Arkansas Nuclear One, Unit 1 - Final Safety Analysis Report, Docket 50-313, i

Arkansas Power & Light.

I 2 T. A. Coler.an and J. T. Willse, Ext. ended Burnup Lead Test Assembly irradi-

)

ation Program, BAW-1626, Babcock & Wilcox (to be published).

I 3 Arkansas Nuclear One, Unit 1 - Cycle 4 Reload Report, BAW-1504, Babcock &

Wilcox, October 1978.

4 J. H. Taylor to S. A. Varga, Letter, "BPRA Retainer Reinsertion," January 14, 1980.

5 Program to Determine In-Reactor Performance of B&W Fuels - Cladding Creep Collapse, BAW-10084, Rev. 1, Babcock & Wilcox, November 1976.

8 Y. H. Hsii, et al., TACO Fuel Pin Performance Analysis, BAW-10141P, Babcock & Wilcox, January 1979.

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

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

& Wilcox, June 1973.

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L. S. Rubenstein (NRC) to J. H. Taylor (B&W), Letter, " Evaluation of In-terim Procedure for Calculating DNBR Reduction Due to Rod Bow," Octouer 18, 1979.

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

& Wilcox, September 1975.

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

Letter, July 8, 1977.

Note: All Babcock & Wilcox reports mentioned above are from NPGD, Lynchburg, Virginia.

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