Text

AR Nuclear One,Unit 1,Cycle 5 Reload Rept.
	ML19341B537
Person / Time
Site:	Arkansas Nuclear
Issue date:	01/31/1981
From:	; BABCOCK & WILCOX CO.
To:	;
Shared Package
ML19341B532	List: ML19341B531; ML19341B537;
References
	BAW-1658, NUDOCS 8102020481
	Download: ML19341B537 (78)
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ARKANSAS NUCLEAR ONE, UNIT 1 ,

, :; - Cycle 5 Reload Report -

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1. INTRODUCTION AND SUKMARY . .. . . . . . . . . . .. . . . . . . . 1-1 3 lm

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

3. GENERAL DESCRIPTION ... .. .. . . .. . . . ... . . . . . . 3-1 2 4 FUEL SYSTE!! DESIGN . .. . .. .. . . . . . . . .. . . . . . . . 4-1 4.1. Fuel Assembly Mechanical Design . . . . . ... . . . . . . 4-1 4.2. Fue 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

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I 6. THERMAL-HYDRAULIC DESIGN . . . .. . . . . . . . .. . . . . . . . 6-1

7. ACCIDENT AND TRANSIENT ANALYSIS . . . . . . .. . . . . . . . . . 7-1 7.1. General Saf ety Analysis . . . . . . . . . .. . . . . . . . 7-1 7.2. Accident Evaluation .. . . . . . . . . . . . . . . . . . . 7-1 o n.

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

9. STARTUP PROGRAM - PHYSICS' TESTING . . . . . .. .. . . . . . . . 9-1 i

,o 9.1. Precritical Tests . ..... .. . . . . . . . . . . . .. 9-1 9.1.1. Control Rod Trip Test . . . . . . . - , . . . . . . . 9-1 p-7._

9.2. Zero Power Physics Tests . . . .. . . . . .. . . . . . .- . 9-2 9.2.1. Critical Boron Concentration . . . . . . . . . . . . 2 G Temperature Reactivity Coefficient ... .. . . . . 9-2 I" 9.2.2.

9.2.3. Control Rod Group Reactivity Worth . . . . . . . . . 9-2 9.2.4. Ejected Control Rod Reactivity Porth . . . . . . . .

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CONTENTS (Cont'd)

Page 9.3. Power Escalation Tests . . . . . . . . . . . . . . . .... 9-3 9.3.1. Core Power Distribution Verific : tion at s40, p 75, and 100% FP With Nominal Control Rod Position . . . . . . . . . . . . . . . . . . .... 9-3 9.3.2. Incore Vs Excore Detector Imbalance Correlation Verification at s40% FP . . . . .... 9-5 79

{l 9.3.3. Temperature Reactivity Coefficient at s100% FP . . . . . . . . . . . . . . . . . . .... 9-5 g 9.3.4. Power Doppler Reactivity Coefficient at

2 s100% FP . . . . . . . . . . . . . . . . . . .... 9-5 D 9.4. Procedure for Use if Acceptance Criteria Not Met .. . .... 9-6 El REFERENCES . .. . . . . . . . . . . . . . . . . . . . . . . . . -. A-1 E

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List of Tables

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Table r 4-1. Fuel Design 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 g 5-2. Shutdown Margin Calculations for ANO-1, Cycle 5 . . . . .... 5-4

[; 6-1. Maximum Design Conditions, Cycles 4 and 5 .. . . . . . .... 6-2 en 7-1. Bounding Values for Allowable LOCA Peak Linear Heat Rates ... 7-2 7-2. Comparison of Key Parameters for Accident Analysis . . . .... 7-3 8-1. Reactor Protection System Trip Setting Limits . . . . ... ... 8-17 w.

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List of Figures Figure 3-1. Fuel Shuffle for ANO-1 Cycle 5 . . . . . . . . . . .. . .... 3-3 g.

m; 3-2. Enrichment and Burnup Distribution, ANO-1 Cycle 5 Off b 320-EFPD Cycle 4 . . . . . . . . . . . . . . . ... . . .... 3-4 3-3. Control Locations and Group Designations for ANO-1 Cycle 5 . .- 3-5 3-4. LEP Enrichment and Distribution, ANO-1 Cycle 5 . . . . .... .3-6

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L 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 . . . . . . . ... . . .. .:. . . .

8-18 8-2. Core Protection Safety Limits . . . . . . . . . . .. . . .... 8-19 8-3. Core Protection Saf ety Limits . . . . . . . . . . . . . . . . : . 8-20 8-4 Protective System Maximum Allowable Setpoints . . . . . . . . . . . 8-21

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8-5. Protective System Maximum Allowable Setpoints . . . . . .... 8-22 8-6. Boric Acid Addition Tank Volume and Requirements Vs RCS Average Temperature . . . . . . . . . . . . . . . . .... 8-23 :

8-7. Rod Position Limits for Four-Pump Operation From 0 to 60 EFFD -- ANO- 1, Cycle 5 . . . . . . . . . . . . . . .... 8-24 8-8. Rod Position Limits for Four-Pump Operation From 50 to 200 1 10 EFPD - ANO- 1, Cy cle 5 . . . . . . . . . . . . . . . 8-25 ;.

8-9. Rod Position Limits for Four-Pump Operation From 200 1 10 to 400 1 10 EFPD -- ANO-1, Cycle 5 . . . . . . .... 8-26 8-10. Rod Position Limits for Four-Pump Operation From 400 10 to 435 2 10 EFPD - ANO-1, Cycle 5 . . . . . . .... 8-27 8-11. Rod Position Limits for Three-Pump Operation From 0 to 60 EFPD -- ANO-1, Cycic 5 . . . . . . . . . . . . . . . . . 8-28 8-12. Rod Position Limits for Three-Pump Operation From 50 to 200 1 10 EFPD -- ANO-1, Cycic 5 . . . . . . . . . .... 8-29 8-13. Rod Position Limits for Three-Pump Operation From 200 2 10 to 400 2 10 EFPD - ANO-1, Cycle 5 . . . . . . .... 8-30 8-14. Rod Position Limits for Three-Pump Operation From ..

400 1 10 to 435 1 10 EFPD - ANO-1, Cycle 5 . . . . . . .... 8-31 8-15. Rod Position Limits for Two-Pump Operation From 0 t o 6 0 E FPD - ANO- 1, Cyc le 5 . . . . . . _ . . . . . . . . . . . 8-32 8-16. Rod Position- Limits for TVo-Pump Operation From 50 to 200 10 EFPD -- ANO-1, Cycle 5 . . . . . . . . . .... 8-33 (2 8-17. Rod Position Limits for Two-Pump Operation From 200 1 10 to 400 1 10 EFPD - ANO-1, Cycle 5' . . . . . . . . . . 8-34 8-18. Rod Position Limits for Two-Pump Operation From 400 2 10 to 435 1 10 EFPD -- ANO-1, Cycle 5 . . . . . . .... 8-35 8-19. Operational Power Imbalance Envelope for Operation From 0 to 60 EFPD -- ANO-1, Cycle 5 . . . . . . .. . . . .... .

8-36 8-20. Operational Power Imbalance Envelope for Operation From 50 to 200 ! 10 EFPD - ANO-1, Cycle 5 . . . . . . . . . . . 8-37 8-21. Operational Power Imbalance Envelope for Operation From 200 1 10 to 400 1 10 EFPD - ANO-1, Cycle 5 . . . . . . . . 8-38 8-22. Operational Power Imbalance Envelope for Operation From 400 1 10 to 435 ! 10 EFPD -- ANO-1, Cycle 5 . . . . . . . . 8-39 8-23. APSR Position Limits for Operation From 0 to 60 EFPD - ANO-1, Cycle 5 . . . . . . . . . . . . . . . . . . . . . 8-40

24. APSR Position Limits for Operation From 50 to 200 1 10 EFPD - ANO-1, Cycle 5 . . . . . . . . . . . . .... 8-41 .

APSR Position Limits for Operation From 200 1 101 i 8-25. 8 L to 400 2 10 E FPD -- ANO-1, Cy c l e 5 . . . . . . . . . . . . . . .

8-26. APSR Position Limits for Operation From 400 1 10 8-43 to 435 10 EFPD -- ANO-1, Cycle 5 . . . . . . . . . . . . . . .

8-44 8-27. LOCA Limited Maximum Allowable Linear Heat Rate . . . . .... [g 3.

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1. INTRODUCTION AND SUTiARY b

j This report justifies the operation of the fifth cycle of Arkansas Nuclear One, y,

Unit 1 (ANO-1) at the rated core power of 2568 Wt. Included are the required k analyses as outlined in the USNRC document, " Guidance for Proposed License Amendments Relating to Refueling," June 1975.

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5i 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 g cecepted by the USNRC and its predecessor, the USAEC (see references).

The cycle 4 and 5 reactor parameters related to power capability are summarized E: 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 r characteristics were conservative compared to those analyzed for previous cy-j cles, no new accident analyses were performed.

t E The Technical Specifications have been reviewed,'and the modifications required E

for cycle 5 operation are justified in this report.

Based on the analyses performed, wh.tch 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 Wt.

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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-li gram. The LTA design is described in reference 2.

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b5 2. OPERATING HISTORY 3 The reference cycle for- the nuclear and thermal-hy'draulic analyses of Arkansas Nuclear One, Unit .1 is the currently operating cycle 4. This cycle 5 design q,

h,i- is based on a design cycle 4 length of 320 ef fective full power days (EFPD).

No anomalies occurred during cycle 4 that would adversely aff ect fuel per-formance during cycle 5.

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b j? 3. GENERAL DESCRIPTION 9

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{ The ANO-1 reactor core is described in detail in section 3 of the Arkansas Nuclear Station, Unit 1 Final Safety Analysis Report (FSAR). ,

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{J The cycle 5 core contains 177 fuel assemblies, each of which is a 15 by 15 array containing 208 fuel rods, 16 control rod guide tubes, and one incore

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[ instrument guide tube. The fuel comprises dished-end, cylindrical pellets of

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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 hava 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.

gg Figure 3-1 is the fuel shuffle diagram for ANO-1, cycle 5. The initial en-ice richments of batches 5B, 6, and 7 are 3.01, 3.19, and 2.95 wt % assU, respec-tively. All the batch IC and batch 4 assemblies and 11 of the twice-burned batch 5 assemblies will be discharged at the end of cycle 4. The remaining 45 twice-burned batch 5 assemblies (designated batch SB) 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

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j at the beginning of cycle 5.

3" Reactivity is controlled by 61 full-length Ag-In-Cd control rods, 64EBPRAs,'

gg and soluble boron shim. In addition to the full-length control rods, eight-b5 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 M 'the group 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 h5 3-1 e- - n , - - .

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group designations, however, differ between cycle 5 and the reference cycle in - t in order to minimize power peaking. Thc cycle 5 locations and enrichments of It the BPRAs are shown in Figure 3-4. {g

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Figure 3-1. Fuel Shuf fle for ANO-1 Cycle 5 Futt TRAM 5FER CAMAL I I

6 6 7 6 6 A K4 K2 Kl4 K12 6 6 6 58 7 6 6 6 I

B M4 L3 N3 P6 N13 L13 M12 6

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Figure 3-2. Enrichment and Burnup Distribution, ANO-1 Cycle 5 of f 320-EFPD Cycle 4 .

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3.01 3.01 3.01 2.95 3.01 2.95 3.01 2.95 H 11750 17438 0 19860 0 16636 0 -

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3.01 2.95 3.01 2.95 3.19 2.95 3.19 X

11750 0 15711 0 13950 0 11861 3.01 2.95 3.01 2.95 3.19 3.19 L

16625 0 13524 0 9937 13118 3.01 2.95 3.19 3.19 M

20117 0 13381 12643 ,

3.19 2.95 3.19 13377 0 13216 3.19

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Figure 3-3. Control Locations and Group Designations for ANO-1 Cycle 5 i

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1 6 6 1 fff;) C 0 7 8 5 8 7 1 5 2 2 5 1 E

F 4 8 3 7 3 8 4

{!! g 6 2 4 4 2 6 r .: }] W 7 5 7 2 7 5 7 y h 6 2 4 '4 2 6 4 8 3 7 3 8 4 L L p ..

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1 8 Safety 2 9 Safety 3 4 Safety 4 12- Safety 5 8 Control -

6 8 Control 7 12- Control 8 8 APSRs Total 69 <

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Figu re 3-4. LBP Enrichment and Distribution, ANO-1 Cycle 5 ,

8 9 10 11 12 13 14 15 H 1.1 0.8 K 1.1 0.8 0.2 ,..-

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X,X LBP Concentration, wt % BSC in Al203 3-6 Babcock s Wilcox 4

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sf e 4. FUEL SYSTEM DESIGN

!f 4.1. Fuel Assembiv Mechanical Design The type of fuel assemblies and pertinent fuel design parameters for ANO-1 cycle 5 are listed in Table 4-1. All fuel assemblies listed are identical in

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concept and are mechanically interchangeable. All results,. references, and s;, identified conservatisms presented in section 4.1 of the cycle 4 reload reort I are applicabic to Fark B4 assemblies. In addition to the assemblies listed,

~e-four lead test assemblies (LTAs) are being inserted with batch 7. One staa-bc 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 q.

y tubes are reported in reference 2.

Retainer assemblies will be used on the fuel assemblies that contain BPRAs to

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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 is applicable to ANO-1, cycle 5. Similar retainer assemblies will be used on

the two fuci assemblics containing the regenerative neutron sources.

4.2. Fuel Rod Design i;.

The batch 7 internal fuel rod design differs from batches 5 and 6.in several respects. As outlined in Table 4-1, these include an increase in initial y

pellet density from 94 to 95% TD, ~a decrease in the nominal fuel pellet diame-ter from 0.3695 to 0.3686 inch, and a reduction in stack length from 142.25 to 141.8 inches. These co=bined changes' vere implemented to improve fuel perfor-

[i mance as well as maintain a constant assembly uranium loading. The mechanical g evaluation of the fuel rod is discussed below.

b 4.2.1. Cladding Collapse p The batch 5 fuel is more limiting than batches 6 and 7 because of its previous incore exposure time. The batch 5 assembly power histories were analyzed to i

determine the most limiting three-cycle power history for creep collapse.

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This worst-case power history was then compared against a generic analysis to ensure that creep-ovalization will not aff ect fuel performance during ANO-1 cycle 5. The generic analysis was performed based on reference 5 and is ap-plicable for the batch S fuel design. p The creep collapse analyses predicts a collapse time greater than 35,000 ef-fective full-power hours (EFPH), which is longer than the maximum expected residence time of 30,960 EFPH (Table 4-1).

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- is radiated yield strength, and all stresses must be less than the minimum speci-fled unirradiated yield strength. In all cases, the margin is greater than h:

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. High thermal gradient across the cladding.

4.2.3. Cladding Strain

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' 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 i

All fuel in the cycle 5 core is thermally similar. The design of 'the four gi, E

batch 7 lead test assembifes is such that the thermal performance of this fuel is equivalent to or slightly better than the standard Mark B design used ,..

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in the remainder of the fuel. The thermal design analysis'~of the LTAs using 8

the TA00-2 code is described in reference 2.

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t The results of the thermal design evaluation of the cycle 5 core are summarized in Table 4-2. Cycle 5 core protection limits were based on a linear heat rate 7

f$ (LHR) to centerline fuel melt of 20.15 kW/f t as determined by the TAFY-3 code ,

with no credit taken for the increased LHR capability of the LTA fuel. The maximum fuel rod burnup at EOC 5 is predicted to be less than 42,000 mwd /mtU.

Fuel rod internal pressure has been evaluated with TAFY-3 for the highest burn-h up fuel rod and is predicted to be less than the nominal reactor coolant system pressure of 2200 psia.

4.4. Material Design

,y The chemical compatibility of all possible fuel-cladding-coolant-assembly in-3 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 x 15 fuel assembly has verified the adequacy of its design. As of July 31, 1980, the following

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experience has been accumulated for the eight operating B&W 177-fuel assembly plants using the Mark B fuel assembly:

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Cumulative net

Current electrical output ,

Reactor cycle Incore Discharged MWh Oconee 1 6 23,300 40,00C 32,457,943 Oconee 2 5 26,100 33,100 27,786.436

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Oconee 3 5 30,200 29,400 28,483,452 T'iI-1 4 32,400 32,200 23,840,053 b ANO-1 4 28,100 33,222 25,006,003 Rancho Seco 4 27,900 37,730 22,625,102 g Crystal River 3 3 20,53,0 23,194 12,113,632 k, Davis-Besse 1 1 14,884 -- 7,654,365 fi p -

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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, Mark BEB l No. of assemblies (*} 45 64 64 Mark B4, 4 :~ ..

Mark BEB Fuel rod OD (nom), in. 0.430 0.430 0.430 ;

v Fuel rod ID (nom), in. 0.377 0.377 0.377 1 Flexible spacers Spring Spring Spring ..

Rigid spacers, type Zr-4 Zr-4 Zr-4 i{f Undensified active fuel 142.25 142.25 141.80 length (nom), in.

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

Fuel pellet initial 94.0 94.0 95.0 density (nom), % TD .

Initial fuel enrichment, 3.01 3.19 2.95 wt % 23sU Average burnup, BOC, mwd /mtU 15,806 12,639 0 Cladding collapse time, >30,000 >30,000 35,000 ;.

EFPM' Estimated residence time, 25,440 28,320 30,960 EFPH (max) ,

(" Four lead test assemblies (Mark BEB) make up a tota'l batch 7 reload of

-h 68 fuel assemblies.- These LTAs were analyzed and reported in reference 2.

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Table 4-2. Fuel Thermal Analysis Parameters

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9., Batch SB Batch 6 Batch 7 m No. of assemblies 45 64 64(a) l- Initial density, % TD 94.0 94.0 95.0 Pellet diameter, in. 0.3695 0.3695 0.36861

[ Stack height, in. 142.25 142.25 141.80 Densified Fuel Parameters Pellet diameter, in. 0.3646 0.3646 0.3649 Fuel stack height, in. 140.5 140.5 140.74 h.:.:.

i- 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 Nominal core avg LHR = 5.80 kW/f t at 2568 MWt.

(*)Four LTAs were also analyzed; the results are reported

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(b) Centerline fuel melt based on fuel specification values.

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'S ff 5. NUCLEAR DESIGN h

ff 5.1. Physics Characteristics c: Table 5-1 lists the core physics parameters of design cycles 4 and 5. The t::.

O values for both cycles were generated using PDQ07. Since the core has not yet reached an equilibrium cycle, differences in core physics parameters are h.! 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.

c Operational changes as well as differences in cycle length, feed enrichment, E' BPRA loading, shuffle pattern, and rod group designations make it difficult

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to compare the physics parameters of cycles 4 and 5. Calculated ejected I rod worths and their adherence to criteria are considered at all times in L "

life and at all power levels in the development of the rod p5sition limits presented in section 8. The maximum stuck rod. worth for cycle 5 is less than n

Ei 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. 10% uncertainty a net rod worth.

6 B 3. Flux redistribution penalty.

l' Flux redistributien was accounted for since the shutdown analysis was calcu-l..@ lated using a two-dimensional model. The reference fuel cycle shutdown mar-gin is presented in the ANO-1 cycle 4 reload report.

Il Analytical Input -

./2. 5.2.

The cycle 5 incore measurement calculation constants to be used for computing r'. core power distributions were prepared in the same manner as these for the reference cycle.

G 5-1 Babcock s.Wilcox ik *

~_ _ .. _. . . _ _

E 3 ,

Changes in Nuclear Design f 5.3.

1 There are no significant core design changes betseen the reference and reload cycles. The calculational methods and design information used to obtain the important nuclear design parameters for this cycle ware the same as those used for the reference cycle. There are two significant operational changes f rom .hh the reference cycle: the withdrawal of the APSRs during the last 35 EFPD of cycle 5 and a change f rom a rodded to a feed-and-bleed mode of operation. The stability and control of the core in the feed-and-bleed mode with APSRs removed E

have been analyzed. The calculated stability index without APSRs is -0.0334 h~1, which demonstrates the axial stability of the core. The operating limits (Tech-nical Specification changes) for the reload cycle are given in section 8. ,,

=

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

Cycle 4( } Cycle 5(c)

Cycle length, EFPD 387 435

~

Cycle burnup, mwd /mtU 12.111 13,633 Avg core burnup, EOC, mwd /mtU 20,505 22,221 Initial core loading, mtU 82.1 82.0 Critical boron - BOC, ppm (no Xe)

HZP(d), group 8 ins 1562 1576 HFP, group 8 ins 1246 1412 Critical boron - EOC, ppm HZP, group 8100% wd, no Xe 418 631

[ 48 HFP, group 8100% wd, eq Xe 86 i

Control rod worths - HFP, BOC, % Ak/k Group 6 1.18 1.21 Group 7 1.02 1.46 Group 8 0.37 0.45 Control rod worths -- HFP, 435 EFPD, % Ak/k i

1.00 1.48 iE Group 7 Max ejected rod worth -- HZP, % Ak/k(*}

BOC (N-12), group S ins 0.76 0.44 5 t

400 EFPD (N-12), group 8 ins 0.82 0.51 ff i

Max st,ck rod worth -- H2P, % Ak/k j.

1.92 1.56 f.E BOC (N-12) 400 EFPD (N-12) 1.86 1.68-p 5-2 Babcock &Wilcox .7

f Table 5-1. (Cont'd) h:

}; Cycle 4 Cycle 5 m Power deficit, HZP to HFP, % ak/k BOC 1.38 1.33 EOC 2.28 2.40 Doppler coef t - BOC,10-s (ak/k/*F) fe 100% power (no Xe) -1.57 -1.52 Doppler coef f - EOC,10~ 5 (ak/k/*F)

g 100% power (eq Xe) -1.71 -1.78 Moderator coef f - HFP ,10-" ak/k/*F)

BOC, (no Xc crit ppm, group 8 ins) -0.48 -0.47 EOC, (eq Xe, O ppm, group 8 out) -2.78 -2.87 Boron worth - HFP, ppm /% ak/k BOC 118 123 EOC 105 106 Xenon worth - HFP, % ak/k BOC (4 EFPD) 2.59 2.58 EOC (equilibrium) 2.75 2.66

~E ffective delayed neutron fraction - HFP BOC

0.00617 0.00628 I.. EOC 0.00517 0.00517 t ...

(* Cycle 5 data are for the conditions stated in this report. The cycle 4 core conditions are identified in reference 2.

( Based on 294 EFPD at 2568 We, cycle 3. *

~

(*) Based on 320 EFPD at 2568 Wt, cycle 4.

(

HZP denotes hot zero power'(532F Tavg), HFP denotes hot full power (579F T,y ).

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

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1 Table 5-2. Shutdown Margin calculations for ANO-1, Cycle 5 .(

BOC, 400 EFPD, 435 EFPD,

% Ak/k % Ak/k % Ak/k Available Rod Worth 'f:

Total rod worth, HZP 9.09 9.40 9.13 Worth reduction due to -0.42 -0.42 -0.42 poison material burnup Maximum stuck rod, HZP -1.56 -1.68 -1.49 ,

Net worth 7.11 7.30 7.22 ;

Less 10% uncertainty -0.71 -0.73 -0.72 Total available worth 6.40 6.57 6.50 ,

Required Rod Worth Power deficit , HFP to HZP 1.33 2.42 2.40 ,

Allowable inserted rod 0.39 0.57 0.50 worth Flux redistribution 0.57 1.19 1.20 Total required worth 2.29 4.18 4.10 Shutdown margin (total 4.11 2.39 2.40 available worth minus total required worth)

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

F 5-4 Babcock & \Milcox

- Figure 5-1. ANO-1 Cycle 5, BOC (4 EFPD) Two-Dimensional Relative Power Distribution - Full Power Equilibrium Xenon, Normal Rod Positions 8 9 10 11 12 13 14 15 1.03 1.17 1.13 1.21 1.15 1.24 0.99 0.72 H

i K 1.21 1.22 1.16 1.23 1.25 1.10 0.60 f .

[. 1.14 1.20 1.05 1.15 0.95 0.43 L

[

1.08 1.19 1.06 0.67 ,

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1.16 0.91 0.43 N

0 0.53

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\X.' Inserted Rod b ,% Group No.

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XXX Relative Power Density

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6. THERMAL-HYDRAULIC DESIGN R The fresh batch 7 fuel is hydraulically and geometrically similar to the pre-M:

viously irradiated batch SB and 6 fuel. The four batch 7 LTAs have been ana-p:.

g lyzed to ensure that they are never the limiting assemblies during cycle 5 opera tion. The results of the thermal-hydraulic analysis of the LTAs are in-l cluded in reference 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-f

'" loved a reduction of the design radial-local peak from 1.78 to 1.71. As a re- ,

l::

sult of this peaking reduction, the steady-state design overpower minimum DNBR I5 increased from 1.88 to 2.05. Table 6-1 summarizes the cycle 4 and 5 maximum design conditions.

p

[' ' A rod bow DNBR penalty has been calculated for cycle 5 according to procedures

n. approved by reference 9. The burnup used to calculate the penalty is conservn-m-

Er tively chosen as the highest assembly burnup in the cycle 5 core of 35,309 mwd /mtU. The resultant rod bow penalty is 3.8%, which includes 1% credit for use of a flow area reduction factor in the DNBR analyses.

[I:[

A flux / flow setpoint of 1.07 has been established for cycle 5 operation. This f 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.

E, 5

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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 p.; ,

Reactor coolant flow, % design 106.5 106.5 E"i Vessel inlet / outlet coolant temp 555.6/602.4 555.6/602.4 a:: 100% power, F Reference design radial-local 1.78 1.71 power peaking factor Reference 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 s,

=

Active fuel length, in. 140.2 140.2 Avgheatfluxa 100% power, 175 175 10 Btu /h-ft2 (a Max heat flux at 100% power, 468 449 l 103 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 i

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

(b) Based on average heat flux with reference peaking.

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Babcock & Wilcox 6-2 ;_ .

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7. ACCIDENT AND TRANSIENT ANALYSIS F 7.1. General Safety Analysis l'ach FSAR accident analysis has been examined with respect to changes in p

I cycle 5 parameters to determine tha 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-usted and are reported in reference 8. Since batch 7 reload fuel assemblies contain fuel rods whose theoretical density is higher than those considered in the reference 8 report, the conclusions in that reference are still valid.

?:.;

h 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 If 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,

ll which represents 28% of the 10 CFR 100 limits.

7.2. Accident Evaluation y[r The key parameters that have the greatest ef fect on determining the outcome of a transient can typically be classiffed 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-

f.f 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 coinpared to the previous cycle 4 values in Table 6-1. These parameters are common to all the accidents considered in this report. The key kinetics parameters from the FSAR and 6 cycle 5 are compared in Table 7-2.

Babcock & Wilcox

..,~ 7-1

A generic LOCA analysis f or a B&W 177-FA, lowered-loop NSS has been performei l using the Final Acceptance Criteria ECCS Evaluation Model (reported in BAW-10103).18 This analysis is generic since the limiting values of key parame-ters for all plants' in this category were used. Furthermore, the combination of average fuel temperatures as a function of LHR and lifetime pin pressure b

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 c values for allowable LOCA peak LHRs for ANO-1 cycic 5 fuel. The basis for two sets of LOCA limits is provided in reference 12.

It is concluded f rom 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 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.

Table 7-1. Founding Values for Allowable TOCA Peak Linear Heat Rates

' Allowable Allowable Core peak LHR, peak LHR, elevation first 50 EFPD, balance of cycle, ft kW/ft kW/ft 2 14.5 INS 4 16.1 16.6 "i

6 17.5 18.0 l 8 17.0 17.0 y 16.0 ;

. 10 16.0 *

, y

! Babcock &Wilcox 7-2 _

Table 7-2. Comparison of Key Parameters

~.

'for Accident Analysis E

FSAR and j

densification ANO-1 Parameter report value c/cle 5 LE 2 Doppler coef f (BOC),10-5 Ak/k/*F -1.17 -1.52 i Doppler coeff (EOC),10-s Ak/k/*F -1.30 -1.78 Moderator coeff (BOC), 10-" Ak/k/*F 0.0(*) -0.47 c.

j Moderator coeff (EOC), 10-4 Ak/k/*F -4.0 -2.87 All-rod group worth (HZP), % Ak/k 12.9 9.09 Initial boron concentration, ppm 1150 1412 Boron reactivity worth (HFP), 100 123

$[:; ppm /% Ak/k

e. Max ejected rod worth (HFP), % Ak/k 0.65 0.30 Dropped rod worth (HFP), % Ak/k O.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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'ri US 8. PROPOSED MODIFICATIONS TO TECHNICAL SPECIFICATIONS t

The Technical Specifications have been revised for cycle 5 operation to ac-IV count for changes in power peaking and control rod worths inherent with the .

transition to 18-month LBP fuel cycles. The cycle 'i flux / flow setpoint is increased from 1.057 to 1.07. This is the result of a flatter radial power distribution.

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-t tain the revisions to previous Technical Specifications. These revisions pro-

~

vide protection for the severe overpower condition that could occur during an i~ overcooling transient because of nuclear instrumentation errors.

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2 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS g 2.1 SAFEYY LIMITS, REACTOR CORE P..

Applicability E.': 7 Applies to reactor thermal power, reactor power imbalance, reactor coolant E 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 the core expressed as a percentage of the rated power) shall not exceed 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 operating conditions. This is accomplished by operating within the nucleate "~

boiling regime of heat transfer, wherein the heat transfer coefficient is 1arge enough so that the cladding surface temperature is only slightly greater

_1 than the coolant temperature. The upper boundary of the nucleate boiling re-gine is termed departure from nucicate boiling (DNB). At this po' int there is a sharp reduction of the heat transfer coef ficient, 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 DSB through the use of the BAW-2 correlation.(1) The BAW-2 corre- - p.

lation has been developed to predict DNB and the location of DNB for axially y 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 DNB. _E The minimum value of the 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 probability at a 95 percent confidence is level that DNB will not occur; this is considered a conservative margin to ,

DNB for all operating conditions. The difference between the actual core outlet pressure and the indicated reactor coolant system pressure has been 8-2

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 r-e assumed in reducing the pressure trip setpoints to correspond to the ele-g vated location where the pressure is actually measured.

The curve presented in Figure 2.1-1 represents the conditions at which a minimum DNBR of 1.3 is predicted. The curve is the most restrictive com-bination of 3 and 4 pump curves, and is based upon the maximum possible thermal power at 106.5% design flow per applicable pump status. This curve is based on the following nuclear power peaking factors (2) with potential

[9 d fuci densification effects:

!? F = 2.56; q F"AH = 1.71; F = 1.50.

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These design limit power peaking factors are the most restr1ctive calculated y at full power for the range from all control rods fully withdrawn to maximum

it allowable control rod insertion, and for the core DNBR design basis, g 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:

e. 1. The 1.3 DNBR limit produced by a nuclear power peaking factor T? of FN = 2.56 or the combination of the radial peak, axial peak, l h and!hepositionoftheaxialpeakthatyieldsnolessthan 1.3 DNBR.

i h 2. The combination of radial and axial peaks that prevents central fuel melting at the hot spot. The limit is 20.1 kk'/f t.

g" 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-pected minimum flow rates with four pumps, three pumps, and one pump in a cach loop, respectively.

The curve of Figure 2.1-1 is the most restrictive of all possible reactor 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 maximum possible thermal power for the number of reactor coolant pumps in operation or the local quality at the point of min-

~

imum DNBR is equal to 22 percent (1), whichever condition is more restrictive.

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 h:: quality at the exit is higher than the quality at the point of minimum DNBR. .

5 p:.

Babcock s \Milcox J- 8-3

,=

ir r.

r 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::

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. ~"!i The maximum thermal power for other reactor coolant pump conditions is pro-duced in a'similar manner.

E:

For each curve of Figure 2.1-3, a pressure-temperature point above and to the E 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 .

coolant pump situation. Curves 1 and 2 of Figure 2.1-3 are the most re-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) Corr.919 tion of Critical Heat Flux in a Bundle Cooled by ricssurized Water, PAW-10000A, May 1976.

(2) FSAR, Section 3.2.3.1.1.c.

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8-4 Babcock & Wilcox

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2.2 SAFETY LIMITS - REACTOR SYSTEM PRESSURE p Applicability b

Applies to the limit on reactor coolant system pressure.

b p

Objective 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 execcd 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-D cordance with the ASME Boiler and Pressure Vessel Code,Section III, Article 9, Summer 1968.

. Bases

.. The reactor coolant system (1) serves as a barrier to prevent radionuclides n 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 of fission products. Eastablishing a system pressure limit helps to assure y

g the integrity of the reactor coolant system. The maximum transient pressure F allowable in the reactor coolant system pressure vessel under the ASME Code, Sectica III, is 110 percent of design pressure.(2) The maximum transient 17 pressure allowable in the reactor coolant system piping, valves, and fittings (f.

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 r::-

established.(2) The settings for the reactor high-pressure trip (2300 psig) i and the pressurizer code safety valves (2500 psig 11%)(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-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 2450 psig. (4)

REFERENCES

$ (1) FSAR, Section 4.

(2) FSAR, Section 4.3.10.1.

h (3) FSAR, Section 4.2.4. .

g.3 (4) FSAR, Table 4-1.

M 8-5 Babcock & Wilcox

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2.3 LIMITING SAFETY SYSTEM SETTINGS, PROTECTIVE INSTRUMENTATION Applicability Applies to instrumants 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 an; combination of process variables from exceeding a safety limit.

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

Bases The reactor protection system consists of four instrument channels to monitor .

each of several selected plant conditions which will cause a reactor trip if any one of these conditions deviates from a preselected operating range to the 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 C instrumentation trip setpoints plus calibration and instrumentation errors.

Nuclear Overpower A reactor trip at high power level (neutron flux) is provided to prevent dam-age to the fuel cladding from reactivity excursions too rapid to be detected 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 would be actuated could be 112%, which is the value used in the safety analysia.

A. Overpower Trip Based on Flow and Imbalance The power 1cvel trip setpoint produced by the reactor coolant system flow is based on a power-to-flow ratio which has been established to accommodate the most severe thermal transient considered in the de-sign, the loss-of-coolant-flow accident from high power. Analysis has demonstrated that the specified power-to-flow ratio is adequate to prevent a DNBR of less than 1.3 should a low flow condition exist due to any cicctrical malfunction.

- Babcock s)Vilcox 8-6

The power level trip setpoint produced by the power-to-flow ratio provides both high power level and low flow protection in the event 3 the reactor power level increases or the reactor coolant flow rate g decreases. The power level trip setpoint produced by the power-to-flow ratio provides overpower DNB protection for all modes of pump operation. For every flow rate there is a maximum permissible power level, and for every power level there is a minimum permissible low flow rate. Typical power level and low flow rate combinations for the pump situations of Table 2.3-1 are as follows:

t;:

p r! 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 p" rate is 93.45 percent and power level is 100 percent.

l E.

2. Trip would occur when three reactor coolant pumps 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 h each loop (total of two pumps operating) if the power is 52.64 t- percent and reactor flow rate is 49.2 percent or flow rate is 45.79 percent and the power level is 49.0 percent.

t

}:1 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.

l y ..

Ne 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 thercal limits from being exceeded. These thermal limits are either power peaking kW/f t limits or DNBR limits. The reactor power imbalances (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 associated with reactor power-to-reactor power imbalance boundaries by 1.07 l

! percent for a 1 percent flow reduction.

[ B. Pump Monitors

F .

" In conjunction with the power imbalance / flow trip, the pump monitors prevent the minimum core DNBR f rom decreasing below 1.3 by tripping R, the reactor due to the loss of reactor coolant pump (s). The pump b monitors also restrict the power level for the number of pumps in

operation.

q.,

7: C. RCS Pressure

.~

During a startup accident from low power or a slow rod withdrawal from high power, the system high-pressure trip setpoint is reached r before the nuclear overpower trip setpoint. The trip setting limit Babcock &)Milcox 8-7 h5 ~

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shown in Figure 2.3-1 for high RCS pressure (2300 psig) has been es- , [2l tablished to maintain the system pressure below the safety limit i (2750 psig) for any design transient.(2) 7, i :U The low-pressure (1800 psig) and variable low-pressure (11.75 Tout

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

O.. l l Due to the calibration and instrumentation errors, the safety analy- li!

i sis used a variable low reactor coolant system pressure trip value of L

, (11.75 Tout - 5143). rz!

j D. Coolant Outlet Temperature _ hh

, The high reactor coolant outlet temperature trip setting limit (619F) .. a i shown in Figure 2.3-1 has been established to prevent excessive core ha r coolant temperatures in the operating range. Due to calibration and ", '

7 instrureentation errors, the safety analysis.used a trip setpoint of ,

620F. ..

E-i

E. Reactor Building Pressure The high reactor building pressure trip setting limit (4 psig) pro- .l'[

g vides positive assurance that a reactor trip will occur in the un- !

i likely event of a secam line failure in the reactor. building or a ,

i loss-of-colant accident, even in the absence of a low reactor coolant L,.

l system pressure trip.

f F. Shutdown Bypass 8

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

Two conditions are' imposed when the bypass is used: _

c

1. A nuclear overpower trip setpoint of 55.0 percent-of' rated power 1

is automatically imposed during reactor. shutdown. .

I

! 2. A high reactor' coolant system pressure trip.setpoint of-1720 psig .

.is automatically imposed.-

The purpose of the 1720 psig high-prussure 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 : p~

-initiated. The overpower; trip setpoint ofL$5.0 prevents any signifi-si cant. reactor power from being produced when performing the physics tests.- Suf ficient natural circulation (5) would be available to remove . ET

.5.0. percent of. rated powerlif none of the reactor coolant pumps wereL ((

' operating.

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Babcock & Wilcox 8 ,

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REFERENCES

. (1) FSAR Section 14.1.2.3.

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'l (2) FSAR, Section 14.1.2.2.

(3) FSAR, Section 14.1.2.7.

(4) FSAR, Section 14.1.2.8.

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L (5) FSAR, Section 14.1.2.6.

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3.2 MAKEUP AND CHEMICAL ADDITION SYSTEMS 0 Applicability Applies to the operational status of the makeup and the chemical addition systems.

i Objective To provide for adequate boration under all operating conditions to assure t ability to bring the reactor to a cold shutdown condition. i Specification 3.2.1 The reactor shall not be heated or maintained above 200F 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 1 cast 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 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1C After the 24 '

hour 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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Bases The makeup and chemical addition systems provide control of the reactor cool-ant s -tem boron concentration. (1) This is normally accomplished by using any .

of the three makeup pumps in series with a boric acid pump associated with the boric acid addition tank. The alternate method o' ooration 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% sub:ritical y margin in the cold condition (200F) at the worst time in core life with a stuck p control rod assembly and after xenon decay.

n Minimum volumes (including a 10% safety factor) as specified by Figure 3.2-1 'r for the boric acid addition tank or 44,549 gallons of 2270 ppm boron as boric l acid solution in the borated water storage tank (3) will each satisfy this re-quirement. The specification ensures that adequate supplies are available :.

1.

8-10 Babcock & Wilcox i

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whenever the reactor is heated above 200F so that a single f ailure 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, g.

w:

The pricipal method of adding boron to the primary system is to pump the m

concentrated boric acid solution (8700 ppm boron, minimum) into the makeup tank using the 25 gpm boric acid pumps.

The altcrnate 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 p- 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-c ature for the concentration present. Once in the makeup system, the concen-y

rate is sufficiently well mixed and diluted so that normal system tempera-P cares ensure boric acid solubility.

1 REFERENCES k?.

(1) FSAR, Sections 9.1, 9.2.

If (2) FSAR, Figure 6-2.

(3) FSAR, Sectica 3.3.

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3.5.2 Control Rod Group and Power Distribution Limits _ f Applicability ,

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

1 Objective To ensure an acceptable core power distribution during power operation, to -

set a limit on potential reactivity insertion from a hypothetical control rod l ejection, and to ensure core subcriticality 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.

p 3.5.2.2 Operation With Inoperable Rods:

1. Operation with more ".aan cae inoperable rod, as defined in Speci- _

fications 4.7.1 and 4.7.2.3, in the safety or regulating rod 'J 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 r immediately to verify the existence of 1% Lk/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, it is not determined that a 1% Ak/k available shutdown margin exists combining the worth of the in- er 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 be exercised within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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 i...

'T the thermal power allowable for the reactor coolant pump combi-nation. .

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Babcock a Wilcox 8-12 3

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[. 6. If a control rod in the regulating or axisi power shaping groups

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is declared inoperable per Specification 4.7.1.2, operation above 60% of the thermal power allowable for the reactor coolant

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pump combination may continue provided the 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 f?.

A 3.5.2.5.3.

The worth of single inserted control rods during criticality is lim-p 3.5.2.3

c. ited by the restrictions of Specification 3.1.3.5 and the Control Rod Position Limits defined in specification 3.5.2.5.

p g 3.5.2.4 Quadrant Tilt:

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

.. shall be reduced immediately to below the power level cutof f (see bF Figures 3.5.2-1A through 3.5.2-1D) . Moreover, the power level cutoff value shall be reduced 2% for each 1% of tilt in excess r: of 4.92% tilt. For less than four-pump operation, thermal power y stall be reduced 2% of the thermal power allowable for the re-actor coolant pump combination for each 1% tilt in excess of 4.92%.

c:

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

~

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

p.

g b. The control rod group and APSR withdrawal limits shall be

~~

reduced 2% in power for each 1% tilt in excess of 4.92%.

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

in 3. If quadrant tilt is in exess of 25%, except for physics tests or P diagnostic testing, the reactor will be placed in the hot shutdown condition. Diagnostic testing during power operation with a quad-e rant power tilt is permitted provided the thermal power allowable for the reactor coolant pump combination is restricted as stated in 3.5.2.4.1 above.

h 4. Quadrant tilt shall be monitored on a ninimum frequency of once h every 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months during power operation above 15% of rated power.

3.5.2.5' Control Rod Positions: ,

l9i 1.

Technical Specification 3.1.3.5 (Safety Rod Withdrawal) does not -

prohibit the exercising cf individual safety rods as required by Table 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% 15 between two sequen-tial groups, except for physics tests.

Babcock &)Milcox

8-13

f b.

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3. Except for physics tests or exercising control rods, (a) the con-trol rod withdrawal limits are specified on Figures 3.5.2-1A, 3.5.2-1B, 3.5.2-1C, and 3.5.2-ID for four-pump operation, on Fig- f ures 3.5.1-2A, 3.5.2-2B, 3.5.2-20, and 3.5.2-2D f or three-pump op-eration, and on Figures 3.5.2-2E, 3.5.2-2F, 3.5.2-2G, and 3.5.2-2H for two-pump operation; and (b) the axial power shaping control y-rod withdrawal limits are specified on Figures 3.5.2-4A, 3.5.2-4B. :~

3.5.2-4C, and 3.5.2-4D. If any of these control rod position lim-its are exceeded, corrective measures shall be taken immediately p to achieve an acceptable control rod position. Acceptable control .

rod positions shall be attained within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

4 Except for physica tests, power shall not be increased above the [

power leve.1 cut-off of 92% of the maximum allowable power level F unless one of the following conditions is satisfied:

a. Xenon reactivity is within 10% of the. equilibrium value for ,

operation at the maximum allowable power icvel and asymptot-ically approaching stability.

b. Except fcr xenon free startup, when 3.5.2.5.4a applies, the ,

reactor has operated within a range of 87 to 92% of the maxi-mum allowable power for a period exceeding 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

3.5.2.6 Reactor Power Imbalance shall be monitored on a frequency not to ex- r cced 2 hcurs during power operation above 40% rated power. Except b for physics tests, imbalance shall be maintained within the envelopes defined by Figures 3.5.2-3A, 3.5.2-3B, 3.5.2-3C, and 3.5.2-3D. If the imbalance is not within the envelopes defined by Figures 3.5.2-3A, 3.5.2-3B, 3.5.2-3C, and 3.5.2-3D, corrective measures shall be taken to achieve an acceptable imbalance. If an acceptable imbalance is not achieved within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , reactor power shall be reduced until imbal- '

ance 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 power-imbalance envelopes defined in Figures 3.5.2-3A, 3.5.2-3B, 3.5.2-3C, E and 3.5.2-3D 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 tempera- [:;

ture will not exceed the Final Acceptance Criteria and (2) the Protective Sys- p tem Maximum Allowable Setpoints (Figure 2.3-2). Corrective measures will be taken immediately should the indicated quadrant tilt, rod position, or imbal- 3 ance be outside their specified boundaries. Operation in a situation that h" would cause the Final Acceptance Criteris to be approached should a LOCA occur is highly improbable because all of ~ the power distribution parameters (quad-rant tilt, rod position, and imbalance) must be at their limits while ,,E t

I 14 Babcock & Wilcox h

H simultaneously all other engineering and uncertainty factors are also at their limits.* Conservatism is introduced by application of:

7

. [.

a. Nuclear uncertainty factors.

b. Thermal calibration.

frn c. Fuel densification effects.

d. Hot rod manufacturing tolerance factors.

e. Fuel rod bowing.

The 20 15% overlap between successive control rod groups is allowed since the wor:h of a rod is lower at the upper and lower parts of the stroke. Control r rod.s are arranged in groups or banks defined as follows:

R, ,

Group Function U

{ 1 Safety 2 Safety 3 Safety 4 Safety 5 Regulating f p~

6 Regulating g 7 Regulating

$- 8 APSR (axial power shaping bank) i The rod position limits are based on the most limiting of the following three f; 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 w?,th the rod position limits, provides for achieving hot shutdown by

= reactor trip at any time, assuming the highest worth control rod that is wit-C 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 L, 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 h

single inserted control rod worth of 1.0% Ak/k is allowed by the rod position E. , limits at hot zero power. A single inserted control rod worth of 1.0% Ak/k I 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.

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

Actual operating limits depend on whether or not incore or excore detectors -

lp are used and their respective instrument and calibration. errors. The method

{lj used to define the operating limits . is defined 'in plant operating procedures.

Babcock & Wilcox m

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The quadrant power tilt limits set forth in Specification 3.5.2.4 have been -

established within the thermal analysis design base using the definition of ..

j quadrant power tilt given in Technical Specifications, Section 1.6. These p'

~

limits, in conjunction with the control rod position limits in Specification 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. i~.

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

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

During the physics testing program, the high flux trip setpoints are administra-tively set as follows to ensure that an additional safety margin is provided: g 9

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F-8-16 Babcock & WilCOX $

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

Four RC pumps operating Three RC pumps operating One RC pump operating in (nominal operating (nominal operating each loop (nominal Shutdown power, 100%) power, 75%) operating power, 49%) bypass Nuclear power, I 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 imbalance, % reduction due to im- reduction due to in- reduction due to im-of rated, aux balance (s) balance (s) balance (s)

Nuclear power based on NA NA 55 Bypassed pump monitors, % of rated, maxc High RC system pressure, 2300 2300 2300 1720*

Psig, max

?

- Low RC system Rressure, 1800 1800 1800 Bypassed

" psig, min d d d Variable low RC system 11.75 T,,g - M 3 11.M T out 03 11.M T,,t - 23 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, asx

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

@ Reactor coolant system flow.

k "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 8

F two-pump operation.

d y T,,g is ghen in degrees FahrenMt M.

I

=

O O

M

Figure 8-1. Core Protection Safety Limits .

(Tech Spec Figure 2.1-1) 2600 r:.

i r.

2400 E

g 2200 5 ACCEPTABLE

$ OPERATION ,

n.

O E 2000 o <

a, j UNACCEPTABLE OPERATION 1800 g

,500 560 580 600 620 640 660 Reactor Outlet Temperature, F F

8-18 Babcock & Wilcox -

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

POWER, 4 , %

R

(-26.88,112) 112 _ 120 (22 J0,112) y2 = -0.77 Ng = 0.66 ACCEPTABLE l4PuwP l 100 (38,100)

I-I (-45,100) l OPERATION

- - 1 l I 1 l

I I i i I

( ! 86.42 I i

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l ACritTABLE

_ _ 80 l l l e l 3 & 4 PUME l

(-45,74.42) I (38,74.42)

.. l OPERATION ! l l l I I

. g i '

60 I l 59.74 l

I ACCEPTABLE l y

' 1 2, 3, & 4 PUNP 1 8 f

(-45,47.14) l OPERATION (38,47,14) 8 I l l

, ~ ~ I I l 40 l s I l t

l i 1 i l i 1 i

i l I i

' I l UNACCEPTABLE l -~

UNACCEPTABLE i i

.l 20

, I I OPERAil0N OPERATION i ,

?! 9 = l =l I in I n ', 88 I 88

.- l ; *l l si l ls i i i i i li

-60 -40 -20 0 20 40 60 h..

REACTOR POWER IN8ALANCE beg%

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d Figure 8-3. Core Protection Safety Limits (Tech Spec Figure 2.1-3) lT 2600 2400

.?

E f 2200

t. 2 f

o 2000 ,

3 a /

1800 f

/f 1600 560 580 600 620 340 660

! Reactor Outlet Temperature, 'F CURVE GPM POWER PUMPS OPERATING (TYPE OF LIMIT) 1 374,880 (100'4)* 112'; FOUR PUMPS (DNBR LIMIT) 2 280,035 (74.7'4) 86.7'4 THREE PUMPS (DNBR LIMIT) 3 184,441 (49.2'4) 59.0'4 ONE PUMP IN EACH LOOP (QUALITY LIMIT)

106.5'4 0F DESIGN FLOW ;

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8-20 Babcock & Wilcox

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Figure S-4. Protective System ?!aximum Allottable Setpoints (Tech Spec Figure 2.3-1) 2500

[:":: P = 2300 PSIG T = 619'F 2300 !

l ACCEPTABLE f-( .5 OPERATION a

f-2l00 a

p P = (11.75T00T - 5103) PSIG

i. ~

=

j t

T: 0 1900 I UNACCEPTABLE 2o j M OPERATION tr e P = 1800 PSIG'

<=

.: 1700 1500 580 600 620 640' 660 560 Reactor Outlet Temperature. *F O

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8-21 Babcock &Wilcox

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i Figure'6-5. Protective System Maxicium Allowable Setpoints ,

(Tech Spec Figure 2.3-2) '

E9 POWER, EA 9 !"

. . 120 (11,107) _

Mg = 0.83

(-15.5.107) 107 M2 -0.92 lACCEPTABL!- 100 ,I

,4 PUMP l (24,95)

(-30,95) g l, OPERATION l l 1 g I

l 79.92 80 ! -

I I ACCEPTABLE I I l 8 I (3&4 PUMP I

(-30,67.92) 10pERATicN l 'U D

I

' 1 i j 60 , l I I 52 G4 4 !

l lACCEPTABLE g i I I 8

12, 3 & 4 I

(-30,40.64) I - - (24,40.64)

I PUMP 40 l '

i I 1

l0PERATION l

! ! I UNACCEPTABLE OPERATION UNACCEPTABLE l e ! l OPERATION o I i - - 20 l o, 8

I IC ,i =s, in I 8

" " i si l r ~ i le !

! ii i t 60 -40 -20 0 20 40 60 -

l POWER IMBALANCE, %

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l 8-22 Babcock & Wilcox

i Figure 8-6. Boric Acid Addition Tank Volume and Requirements Vs RCS Average f Temperature j (Tech Spec Figure 3.2-1) 6000 j gg 8700 PPMB t- OPERATION ABOVE AND TO

. 3500 PPMJ in THE LEFT OF THE CURVES

- Is ACCEPTABLE 5000 L

b !kbGAL .

}- :d

= . 12000 PPMB F

m 4000 -

l9l.jggt .

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/!b!bGAL t .

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0 3000 - /

G

500F 287 3 GAL g /

M /

S hhhhGAL /

$ 2000 -

/

/ 3 GAL

- /

1000 - 300F /

974 GAL /300F

/

/ 698 gal.

/

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l f f 200 300 400 500 600 U

RCS AVERAGE TEMPERATURE (F) h p

8-23 Babcock & Wilcox

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U Figure 8-7. Rod Position Limits for Four-Pump Operation From 0 to 60 EFPD - ANO-1, Cycle 5 (Tech Spec Figure 3.5.2-I A) t 110 (271,102)i' 100 POWER LEVEL SHUTOOWN U OFF MARGIN (271.92) 90 LIMIT 80 - (258,80)

OPERATION IN 70

. THIS REGION O RESTRICTED IS NOT REGION

$ ALLOWED g 60 -

~

I IBM E 50 (67.50 (175 50) OPERATING ae REGION g 40 -

E 30 20 -

, ( 0.13 )

10 -

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

,0 2,0 4,0 60 g010p GROUP 7 0

30 4p 6p 8Q 10Q GROUP G 0 20 40 80 100 t i . i GROUP 5 Rod index. % WO 8-24 Babcock & Wilcox

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-S' Figure 8-8. Rod Position Limits for Four-Pump Operation From 50 to 200 1 10 EFPD - ANO-1, Cycle 5 6 (Tech Spec Figure 3.5.2-1B) 110

( }"

100 SHUT 00nN POWER LEVEL (271.92) >

MARGIN

[-

~~

90 CUT 0FF LIMIT 80 - (258.80) p:

p OPERATION IN THIS REGION 70 RESTRICTED

- IS NOT

' E t.LLOWED REGION

- PERMISSIBLE b = 60 OPERATING E REGION i (67,50) (175.50;

50 -

ne

40 t;

d' 30 -

(.:. 20 -

i

( 0,13 )

E 10 -

t 0i , , , , , ,

(0.0)0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300

,0 29 40 6,0 80 10p i

GROUP 7 0 20 40, 60 8,0 100

! ! f f GROUP 6 0 20 40 80 100 f i t e i

GROUP 5 Roc Index, % WD f,.,

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8-25 Babcock & Wilcox

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Figure 8-9. Rod Position Limits for Four-Pump Operation From . I 200 10 to 400 1 10 EFPD - ANO-1, Cycle 5 (Tech Spec Figure 3.5.2-1C) ..

E 110 (271,102) e (215.102) gp ::

100 -

r R

POWER LEVEL CUT 0FF >

i 90 - SHUTOOWN

' N 1,92:

RESTRICTED 0

lN REGION (258,80)

OPERATION IN

. THIS REGION 70 -

IS NOT (240.70) y ALLOWED

= 60 -

~

50 -

(156.50) (175,50) PERMISSIBLE a OPERATING

- REGION

_ 40 _

E' 30 -

20 -

10 , (83.15)

(' ' }

0 ' f f f f f I '

1 f f I f

( 20 40 50 60 100 120 140 160 180 200 220 240 260 280 300

,0 JO 4,0 q0 8.0 10,0 GROUP 7 Q 2p 49 90 8,0 1p0 0 20 40 80 100 GROUP 6 t t

f e GROUP 5 Rod Index', % WD E

, . t r;;

1

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Figure 8-10. Rod Position Limits for Four-Pump Operation From 400 t 10 to 435 t 10 EFPD - ANO-1, Cycle 5 (Tech Spec Figure 3.5.2-ID) 110 (220,102) (283.102) 100 .

SHUTDOWN RESTRICTE0 Jm 90 -

279,92 REGION POWER LEVEL MARGIN f .~

"~ CUT 0FF LINIT 80 -

(238,80) i .

70 - CT2 RAT 10N IN g THis REGION

= 15 NOT

" 60

(

ALLOWED a

ta (175,50)

50 - (156,50 PERWIS$18tE g

OPERATING i

40 I

REGION y '

30 -

20 -

(83.15)

(0,8.9) 10 ,

(80.0)

{.. 0 i ' ' ' ' ' ' ' ' ' ' ' '

O 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 9 30 j0 90 q0 190 GROUP 7 9 29 4p q0 q0 igg GROUP 6 p 70 40 SO IQ0 L

t, GROUP 5 ood index, % 50 j

l i

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8-27 -

Babcock s.Wilcox i _ _ .

Figure 8-11. Rod Position Limits for Three-Ptep Operation - I:J From 0 to 60 EFPD - ANO-1, Cycle 5 (Tech Spec Figure 3.5.2-2A) 110 100 -

90 -

80 -

SHUTOOWN (134,77)

MARGIN (250,77) 70 - OPERATION IN LIMIT THIS REGION RESTRICTED b 60 - IS NOT REGION g ALLOWE0 N 50 -

(175,50)

=

PERMISSIBLE

" 4 0 (67,38) OPERATING g REGION

$ 30 -

20 -

10 C ' ' ' ' ' ' ' ' ' ' ' ' ' ' '

0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 0 2p 4,0 6,0 8p IQO GROUP 7 0 2p 40 6,0 8,0 10,0 GROUP 6 0 2,0 4,0 8,0 10,0 GROUP 5 l

Rod index , % WD 8-28 Babcock & Wilc0X

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Figure 8-12. Rod Position Limits for Three-Pump Operation From 50 to 200 1 10 EFPD - ANO-1. Cycle 5 (Tech Spec Figure 3.5.2-2B) 110 100 -

90 -

u

~

SHUTDOWN (134,77) (250,77) h.:.

L MARGIN 70 -

OPERATION IN LIMIT THIS REGION RESTRICTED E E REGION lj = 60 - IS NOT

~'.

ALLOWED l @

f ': ; 50 - (175,50)

. ow PERMISSIBLE

OPERATING 40 -

(67,38)

% REGION E -

30 f

20

[.

10 4 (0,10) 0, , , , , , , , , , , ,

i i ,

0 20 40 60 80 100 120 140 160 100 200 220 240 260 280 300 l

0 2,0 40 6,0 8p ig0 GROUP 7 0 2,0 40 6,0 8,0 10,0 GROUP 6 0 2,0 40 10,0 i i i.

GROUP 5 Rod index, % WD ,

a h

1 1

if.

l f . _ - - - - .

8-29 Babcock & Wilcox

Figure 8-13. Rod Position Limits for Three-Pump Operation From 200 t 10 to 400 2 10 EFPD - ANO-1, Cycle 5 .

(Tech Spec Figure 3.5.2-2C) ,

110 E

100 -

90 Restricted Region 80 -

OPERATION IN r 70 -

THit REGION SHUTDOWN (240,70)

' iS NOT MARGIN

$ 60 .

ALLOWED LIMIT M

m

50 -

at PERMISSIBLE J 40 -

(156,38) OPERATING g REGION 30 -

20 -

(83,11 )

10 , ,-

(0.7) ,

0 ' ' ' ' ' ' ' ' ' ' ' ' ' '

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

0 20 t

40 60 t

80 f 100 1

GROUP 7 0, z0 4Q 6,0 8,0 1p0 GROUP 6 0 20 4,0 100 t i GROUP 5 Rod Index, % WO

g. :

8-30 Babcock & Wilcox n l

I

Figure 8-14. Rod Position Limits for Three-Pump Operation From 400 ! 10 to 435 10 EFPD - ANO-1, Cycle 5 (Tech Spec Figure 3.5.2-2D) 110 100 -

i i 90 -

Restricted

'" Region (220.77) p (232.77) 70 SHUTDOWN y -

P WARGIN l

0 llMli

I 60 _

i: OPERATION IN I THIS REGION 50 IS NOT R

l ALLOWED EE 40 -

PFRMISSIBLE f OPERATING

- 30 _

REGION 20 _

( .. (0,7)

(83.11 )

10 -

r 0 , , ,

0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 0, 20 4,0 6,0 8,0 igg Rod Index, % WO GROUP 7 20 40 6,0 80 100 0, I f f f GROUP 6 60 80 100

,0 2,0 4,0 GROUP 5 a-8-31 -

Babcock & Wilcox

r.

p:

i.

Figure 8-15. Rod Position Limits for Two-Punp Operation Frota 0 to 60 EFPD - ANO-1, Cycle 5 (Tech Spec Figure 3.5.2-2E) .g.;

i::

110 c:.

100 - h 90 _

80 -

70 -

60 ~

$ OPERATION IN SHUT 00NN (134.52)

IN WN (182.52) 50 -

l te ALLOREO

%g 30 -

PERMISSIBLE (67.26)

OPERATING

.-(0,7) REGION i 20 10 -

. . i i , , i ,

0.

i 0 20 40 60 60 100 120 140 160 180 200 220 240 260 280 300 0, 2,0 40 6,0 8,0 10,0 GROUP 7 p 20 4,0 6,0 8,0 10,0 ,.

GROUP 6 0 2,0 4,0 60 80 10p GROUP 5 Roa Index. % WO i

=

e.

8-32 Babcock & Wilcox ,

? ..

[*

~

Figure 8-16. Rod Position Limits for Two-Pump Operation From 50 to 200 10 EFPD - ANO-1, Cycle 5

-)) (Tech Spec Figure 3.5.2-2F)

.i 110

'f:.::

1-- --

100 h, 90 -

t 80 -

. .g g

6 70

~

?:: g

= 60

" (134'52) (182,52)

-- OPERATION IN SHUTDOWN 50 _

THIS REGION MARGIN

' " IS NOT LIMIT f 40 -

ALLOWED E

30 - PERMISSIBLE (67,26) OPERATING (I 20 -

REGION im 10 -

i (0.7)

( 0,0 ) 0' ' ' ' ' '

, 0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 O, 20 4,0 6,0 89 1,00 GROUP 7 0, 2,0 40 6Q 8,0 10,0 GROUP 6 0 20 40 80 100 f f f f n GROUP 5 Rod index, % WD U

E l

E kh'.

I j !.:

=;

8-33 Babcock & Wilcox s

,h.i.

Figure 8-17. Rod Position Limits for Two-Pump Operation From 200 ! 10 to 400 t 10 EFPD - ANO-1, Cycle 5 (Tech Spec Figure 3.5.2-2G) ( ..

l :.

110 p

7 100 90 -

80 70 -

E

= 60 -

0 OPERATION IN SHUTOOWN (215,52)

E 50

- THIS EEGION NARGIN de IS NOT tigit ALL0tED t 40 E PERMISSIBLE OPERATING 30 (156,26) REGION 20 - l .

10 -(0.5) _ (83,8) 4 0 , .

0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 Q 2.0 4p qu 40 100 ...

O 20 40 60 80 100 I f f 8 1 I l 100 GROUP 6 0 20 40 80 I e f i l GROUP 5 Rod Index, % N0 t:.

. IS

,Il.-

[-

8-34 Babcock & Wilcox '

?

L

es i .; '

V.

/!i.

'- Rod Position Limits for Two-Pump Operation From Figure 8-18.

400 10 to 435 10 EFPD - ANO-1, Cycle 5 E (Tech Spec Figure 3.5.2-2H) bi 110 100 -

-j. 90 -

80 -

G j 70 -

. ,7 2 60 -

"o

<< 50 -

SHUTOOWN MARGIN 40 OPERATION IN LIMIT .

THIS REGION IS NOT 30 (156,26)

ALLOWE0 PERMISSIBLE

( 20 -

OPERATING 10

- (0,5) _

+

0 i ! ' ' I i i i i 1 ' i i i 0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300

,0 2,0 40 69 $0 1q0 GROUP 7 p 2p 49 90 g0 10,0 r GROUP G-g 20 4p 1,00 GROUP 5 Rod index, % WO l

l b^

y i .

i :;

i 1 ,f.

8-35 Babcock a.Wilcox f.f

?

p Figure 8-19. Operational Power Imbalance Envelope for Operation From 0 to 60 EFPD - A50-1, Cycle 5 (Tech Spec Figure 3.5.2-3A) '.

. 110

~..

!S c

(-18.102) rp . igg o (25.102)

(-19,92) (25,92)

. . 90

(-26,80) . .

80 (32,80)

PERMISSIBLE RESTRICTED RESTRICTED OPERATING REGION REGION REGION

. 60

. . 50 40

. m "w

g- - 30 si ae - - 20 5

$- - 10

, i e i , , .

l -40 -30 -20 -10 0 10 20 30 40 Axial Power Imualance, %

i C.1 l

l l

8-36 Babcock & WilCOX i

_ _ _ . - _ . _ . . _ . = - .-

'=

cp r::

fie "

. Figure 8-20. Operational Power Imbalance Envelope for Operation From 50.to 200 10 EFPD - ANO-1, Cycle 5 '

f (Tech Spec Figure 3.5.2-3B) gf;

- 110 et:

d (-24,102) c ' '

100

[, h (-24,92) (25,92)

. . 90

(-34.80) . . 80 (32.80)

PERMISSIBLE OPF. RATING.

' t.?: RESTRICTED RESTRICTED REGION

= REGION REGION

-- 60

.=

- 50 -

40 E

i .

- - 30 n

=-

g. 20

.. x

.._ $- 10 E

o.

I f f I I I I i 40 -30 -20 -10 0 10 20- 30 40 Axial Power imoalance % ;

lii lt::

I h.'

r::

C p

5

.- g 1

.- 8-37 Babcock & Wilcox

(.

i.

E"

'i:

b Figure 8-21. Operational Power Imbalance Envelope for Operation From 200 i 10 to 400 ! 10 EFPD - ANO-1, Cycle 5 (Tech Spec Figure 3.5.2-3C) [}

. 110 f

(-19,102) 0 ( '

100 p

(-23,92) , ,

(25.92)

(-32,80' - -

80 (32,80)

PERMISSIBLE RESTRICTED OPERATING RESTRICTED REGION REGION REGION 60 a:

~ ~

50

- 40 E

m

=- - 30 N

Ei - - 20 a

10 e

f I I I I f 1 e

-40 -30 -20 -10 0 10 20 30 40 Axial Power Imualance %

f

(

E

?01 l

j,2 p

8-38 Babcock a.Wilcox j:l

r Operational Power Imbalance Envelope for Operation Figure 8-22.

From 400 !_10 to 435,Jt 10 EFPD - ANO-1, Cycle 5 E (Tech Spec Figure 3.5.2-3D) bll

-- 110

%l I C

(-16.102) -

- 100 lii (16,92)

(-16,92) . - 90 i

n

(-32,80) - -

80 (24,80) iE PERMISSI Bl.E RESTRICTED e RESTRICTE0 OPERATING REGION REGION REGION i- -

60

~

50 40 E

=

3. g- 30

i. N

$ 20

= -

5- .

10 a

m

, , , , , . i i

, -40 -30 -20 -10 0 10 20 30 40

" Axial Power Imbalance %-

t=

b G

G i

E 8-39 Babcock & Wilcox l.

APSR Position Limits for Operation From i-Figure 8-23.

O to 60 EFFD - ANO-1, Cycle 5 (Tech Spec Figure 3.5.2-4A)

U 110 (2,102) in e

m (41,102)

100 -

RESTRICTE0 E

J, (2,92) 90 (41,80) 80 < > (0,80) <

70 -

= PERMISSIBLE g 60 -

OPERATit;G N REG 10t1 50 (100.50)

E 40 -

2 30 20 -

10 -

l

. . . . . e i i 0 i 0 10 20 30 40 50 60 70 80 90 100

% Withdrawn l

l

=

c 8-40 Babcock & Wilcox

's, APSR Position Limits for Operation From 50 Figure 8-24.

to 200 ! 10 EFPD - ANO-1, Cycle 5 (Tech Spec Figure 3.5.2-4B) 110 (9.102) f~ '

(41, 02) 100 -

RESTRICTED i (9,92) 90 - REGION II: 80 (0,80) 4 (41.80) r.

E. 70 -

C m

= 60 -

PERMISSIBLE g 0/ERATING ,

- REGION 50 - '

(100.50) p- a y 40 -

F- cE l

30 -

(. 20 - )

Li 10 -

b 0 , , , , , , , , ,

20 30 40 50 60 70 80 90 100 c 0 10

% Withdrawn 1 l

[_:

L 8-41 Babcock s,Wilcox

_n-ve m ---,,,,w ,- e , e- a - e-

r:

I Figure 8-25. APSR Position Limits for Operation From 200

10 to 400 1 10 EFPD - ANO-1, Cycle 5 (Tech Spec Figure 3.5.2-4c) ;..

110 (13.5,102) p.

100 _

' ' O( '

RESTRICTED (13.5,92) REGION i 90 -

i (0,80) ,

80 70 - p G

60 _

PERMISSIBLE E OPERATING N REGION g 50 -

a t 40 -

5 l 30 20 10 -

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

l l % Withdrawn I

E

=

8-42 -Babcock 8.Wilcox

- Figure 8-26. APSR Position Limits for Operation From 400 t 10 to 435 10 EFPD - ANO-1, Cycle 5 (Tech Spec Figure 3.5.2-4D) 110 f :.

,~

100 -

90 _

b.

80 -

70 - APSR INSERTION NOT ALLOWED E IN THIS TIME INTERVAL

[ 60 -

m n

50 -

at d 40 -

E a_

30 -

1 20 -

I 10 -

0 , ,

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

,c fi^

i=

t ..

j=.4

[$.

D h

8-43 Babcock & Wilcox

_, m_,, __ . , - - - - -- - - , ,-.y-m-- ..w.-- .---,.e w .-c--g- -.

l 1::

p l Figure 8-27. LOCA Limited Maximum Allowable Linear lleat Rate ,

l; (Tech Spec Figure 3.'5.2.4) i:!:

21 20 p::

.~_ 19 s,

e.ac

.; 18 BALANCE OF CYCLE 3 17 I, \ ,

~

16 \

Z

//- /- FIRST 50 EFPD E / .

E 14 13 12 0 2 4 6 8 10 12 Axial location of Peak Power from Bottom of Core, ft p:

cs

[.;

8-44 Babcock 8.Wilcox .:

. 'j l

)

1 p -

J

,; 9. STARTUP Ph0 GRAM - PHYSICS TESTING The planned startup test program associated with core perfornance is outlined below. These tests verify that core performance is within the assumptions of g the safety analysis and provide courirmation for continued safe operation of N the unit.

s 9.1. Precritical Tests u

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. Acceptable g criteria state that the rod drop time from fully withdrawn to 75% inserted t

shall b'e less than 1.46 seconds at the conditions above.

I

{ It should be noted that safety analysis calculations are based on a rod drop from fully withdrawn to two-thirds inserted. Since the cost accurate position

[i indication is obtained from the zone reference switch at the 75%-inserted b~

position, this position is used instead of the two-thirds inserted position r for data gathering.

E b.:

f F:

R l

E!

b:

p qs -

s 8:

9-1 Babcock & Wilcox

t. 4

r h.

9.2. Zero Power Physics Tests .

E 9.2.1. Critical Boron Concentration Criticality is 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 E correcting for any rod withdrawal required to achieve equilibrium boron. The ...

p acceptance criterion placed on critical boron concentration is that the ac- {.

teal boron concentraiton must be within 1100 ppm boron of the predicted value.

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

Acceptance criteria state that the measured value shall not differ from the [

predicted value by more than 20.4 x 10-" (ak/k)/*F (predicted value obtained '

from Physics Test Manual curves).

The moderator coefficient of reactivity is calculated in conjunction with the I temperature coefficient measurement. After the temperature coefficient has been measured, a predicted value of fuel Doppler coefficient of reactivity is l added to obtain moderator coefficient. This value must not be in excess of the =

f Technical Specification limit of +0.5 x 10-" (Ak/k)/*F.

l 9.2.3. Control Rod Group l

Reactivity Worth L Control bank group reactivity worths (groups 5, 6, and 7) are measured at hot zero power conditions 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 r eactimeter 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 p 9-2 Babcock & Wiledx 5

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:

predicted value - measured value x 100 5 15 measured value

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

4 predicted value - measured valu x 100 5 10 measured value 9.2.4. Ejected Control Rod Reactivity Worth After CRA groups 7, 6, and 5 have been positioned near the minimum rod inser-tion limit, the ejected rod is borated to 100% withdrawn and the worth ob-

. tained by adding the incremental changes in reactivity. .

- 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.

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

1, lpredictedvalue-measuredvalue measured value x 100l520

2. Measured value (error-adjusted) 5 1.0% Ak/k i

j.. 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 N40, 75, and 100% FP With Nominal Contrcl Rod Position

. Core power distribution tests are performed at 40, 75, and 100% full power in- (FP). The test at 40% FP is essentially a check on power distribution in the n

core to identify any abnormalities before escalating to the 75% FP plateau.

L Rod 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 imbalance corresponding' to the imbalance at which the core

[ power distribution calculations were performed.

N y 9-3 Babcock & Wilcox

E e

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

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

2. The minimum DNBR must be greater than 1.30.

u"'

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, p.

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 trip 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 valu x 100 5 8.

measured value

7. The highest measured and predicted total peaks shall be ,

within the following limits:

predicted value - measured valu x 100 5 12.

, measured value Itecs 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.

Itecs 3 and 4 establish the criteria whereby escalation to the next power ,

plateau may be accomplished without exceeding the safety limits specified by the safety analysis with regard to DNBR and linear heat rate.

C The power distribution tests performed at 75 and 100% FP are identical to the E

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 ;s within the following limits: l predicted value - measured value x 100 ~5 < "

measured vclue 1 Babcock & Wilcox -

9-4 F4

l

\/

i 4Ai

$++/ ,... - -

. TEST TARG,T (MT-3)

'*%'s4 1.0 g m L24 g m gn cm l,l p

bb l]l.8

' ~- '

l.25 1.4 1.6 l __

= 6" S

/+4 %

$f>p////

o '4M)+[

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

predicted value - measured valu measured value x 100 5 7.5 ca 9.3.2. Incore Vs Excore Detector Imbalance Correlation Verification at %40% FP

'= 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.

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

G"~

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 coderator temperature coefficient shall be negative.

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

- activity corrections are made for changes in xenon and reactor coolant tem-perature that occur during the measurement. The power Doppler reactivity coefficient is calculated from the measured reactivity change, adjusted as f( stated above, and the measured power change. The predicted value of the power Doppler reactivity coefficient is given in the Physics Test Manual, t.:

L. Acceptance criteria state that the-measured value shall be more negative than c

-0.55 x'10 " (ak/k)/% FP.

-9

.f.,

i M, .- s a

p. 9-5 Babcock & VVilcox 4

[

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 cont inued. The result s of all tests will If the acceptance be reviewed by the plant's nuclear engineering group.

criteria of the startup physics tests are not met, an evaluation will be per-formed by the plant's nuclear engineering group with assistance f rom general of fice personnel, Middle South Services, and the fuel vendor, as needed. The results of this evaluation will be presented to the On-site Plant Safety Com-mittee. Resolution will be required prior to power escalation. If a safety question is involved, the Of f-site Safety Review Committee would review the situation, and the NRC would be notified if an unreviewed safety questien exists.

t 9-6 Babcock & Wilcox

5 n

r,.

=

REFERENCES

~

l Arkansas Nuclear One, Unit 1 -- Final Safety Analysis Report, Docket 50-313, Arkansas Power & Light.

2 T. A. Coleman and J. T. Willse, Extended Burnup Lead Test Assembly Irradi-ation Program, BAW -1626, Babcock & Wilcox, October 1980.

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

Wilcox, October 1978.

J. H. Taylor to S. A. Varga , Letter, "BPRA Retainer Reinsertion," January 14, 19P 0.

=

s

Program to Determine In-Reactor Performance of B&W Fuels - Cladding Creep

~

Collapse, BAW-10084, Rev. 1, Babcock & Wilcox, November 1976.

6 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.

I L. S. Rubenstein (NRC) to J. H. Taylor (B&W), Letter, " Evaluation of In-

, terim Procedure for Calculating DNBR Reduction Due to Rod Bow," October i

18, 1979.

"i 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.

12 J. H. Taylor (B&W Licensing) to L. S. Rubenstein (USNRC), Letter, September 5, 1980.

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

Babcock & Wilcox A-1

ML19341B537

Text

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