ML19326B716

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Cycle 3 Reload Rept.
ML19326B716
Person / Time
Site: Arkansas Nuclear Entergy icon.png
Issue date: 11/30/1977
From:
BABCOCK & WILCOX CO.
To:
References
BAW-1471, NUDOCS 8004170552
Download: ML19326B716 (44)
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ARKANSAS NUCLEAR ONE, UNIT 1

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- Cycle 3 Reload Report - '

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BAW-1471 November 1977 ARKANSAS NUCLEAR ONE, UNIT 1

- Cycle 3 Reload Report -

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BABCOCK & WILCOX Power Generation Group l'> Nuclear Power Generation Division

! - P. O. Box 1260 l- Lynchburg, Virginia 24505 i

Babcock & Wilcox

CONTENTS Page

1. INTRODUCTION AND

SUMMARY

. . . . . . . . . . . . . . . . . .... 1

2. OPERATING HISTORY . . . . . . . . . . . . . . . . . . . . .... 2
3. GENERAL DESCRIPTION . . . . . . . . . . . . . . . . . . .... 3 4
4. FUEL SYSTEM DESIGN . . . . . . . . . . . . . . . . . . . . .... 8

- 4.1. Fuel Assembly Mechanical Design . . . . . . . . . . .... 8 4.2. Fuel Rod Design . . .. . . . . . .. .. . . . . .. ... 8 4.2.1. Cladding Collapse . . . . . . . . . . . . . .... 8 4.2.2. Cladding Stress . . . . . . . . . . . . .. .... 9 4.2.3. Cladding Strain . . . . . . . . . . . . . . .... 9 4.3. Thermal Design . . . . . . . . . . . . . . . . . . . .... 9 4.4. Material Design . . . . . . . . . . . . . . . . . .. ... 9 i

4.5. Operating Experience . . . . . . . . . . . . . . .... 9

-l 5. NUCLEAR DESIGN . . . . . . . . . . . . . . . . . . . . . . .... 13 5.1. Physics Characteristics . . . . . . . . . . . . . . .... 13 5.2. Analytical Input . . . . . . . . . . . .. . . . . . .... 14 5.3. Changes in Nuclear Design . . .. . . . . . . . . .. ... 14

6. THERMAL-HYDRAULIC DESIGN . . . . . . . . . . . . . . . . . .... 19
7. ACCIDENT AND TRANSIENT ANALYSIS . . . . . . . . . . . . . .... 21 7.1. General Safety Analysis . .. . . . . . . . . . . . .... 21 7.2. Accident Evaluation . . . . . . .. . . . . . . . . .... 21
8. PROPOSED MODIFICATIONS TO TECHNICAL SPECIFICATIONS . . . . .... 24
9. . STARTUP PROGRAM -- PHYSICS TESTING . . . . . . . . . . . . .... 39
10. REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . .... 40

- 111 Babcock & Wilcox b

n List of Tables

' Table Page 4-1. Fuel Design Parameters and Dimensions . ... . .. . .. . .. .

11 4-2. Fuel Thermal Analysis Parameters . . . . . . . . . . . . . . . .

12 5-1. ANO-1, Cycle 2 and 3 Physics Parameters . . . . . . .. . . . .. 15 5-2. Shutdown Margin Calculation, ANO-1,_ Cycle 3........... 17 6-1. Cycle 2 and 3 Maximum Design conditions .

7-1. Bounding Values for Allowable LOCA Peak Linear . .. . . . .. . . .. 20 Heat Rates . . . . 22 7-2. . Comparison of Key Parameters for Accident Analysis . . . . . . . 23 List of Figures Figure 3-1. Core Loading Diagram for ANO-1, Cycle 3 .. . . . . .. . . . . 5 3-2. Enrichment and Burnup Listribution for ANO-1, Cycle 3 . . . .. 6 3-3. Control Rod Locations for ANO-1, Cycle 3 . . . .. . . . . . . . 7 5-1. BOC 3 _(4 EFPD) Two-Dimensional Relative Power Distribution --

Full Power, Equilibrium Xenon, Normal Rod Positions . . . . 18 8-1. ANO-1 Cycle 3_ Core Protection Safety Limits . . . . . . . . . .

25 8-2. ANO-1 Cycle 3 Protective System Maximum Allowable Setpoints . . 26 8-3. Rod Position Limits for Four-Pump Operation From 0 to 100 i 10 EFPD -- ANO-1, Cycle 3 . . . . .. . ... . . . . . . . .. 27 8-l . Rod Position Limits for Four-Pump Operation From 100 10 to 250 10 EFPD - ANO-1, Cycle 3 . . .. .. . . . . . . . . . 28 8-5. Rod Position Limits for Four-Pump Operation Af ter 250 10 EFPD - ANO-1, Cycle 3 ............ . . . . . . . .. 29 8-6. Rod Position Limits for Two- and Three-Pump Operation From 0 to 100 10 EFPD - ANO-1, Cycle 3 . . .. . . . . .. . . .. 30 8-7. ~

Rod Position Limits for Two- and Three-Pump Operation From

. 100 t 10 to 250 10 EFPD - ANO-1, Cycle 3 . . .. . . . . . . . 31 8-8. Rod Position Limits for Two- and Three-Pump Operation Af ter 250 10 EFPD -- ANO-1, Cycle 3 . . . . . . . . . . . .. . . . . 32 8-9. Operational Power Imbalance Envelope for Operation From 0 to 100 10 EFPD -- ANO-1, Cycle 3 33 8-10. Operatinnal Power Imbalance Envelope .for. Operation From

,100 10 to 250 1 10 EFPD -- ANO-1, Cycle 3 . . . .. . .. . . . 34 8-11. Operational Power Imbalance -Envelope for Operation After 250 t 10 EFPD'- ANO-1, Cycle 3 . . . . . . . . . . . . . . j

. . . 35 7 8-12. APSR Position Limits for.0peration From 0 to 100 10 E FP D -- ANO-1, Cy c le 3 . . . . . . . . . . . . . . . . '. . . . . . 36 8-13. APSR Position Limits for Operation From 100 1 10 to 250 t 10 EFPD -- ANO-1, cycle 3 . . . . . . . . . . . . . . . .. 37 8-14. APSR Position Limits for Operation Af ter 250 1 10 EFPD -- ANO-1. Cycle 3 ... . . . . .. .. . . . . . . . . .. 38

- iv - Babcock & \Milcox

1. INTRODUCTION AND

SUMMARY

This report justifies the operation of the third cycle of Arkansas Nuclear One, Unit 1 (ANO-1) at the rated core power of 2568 MWt. Included are the required analyses outlined in the USNRC document, " Guidance for Proposed License Amendments Relating to Refueling," dated June 1975. Supporting cycle 3 operation of ANO-1, this report employs analytical techniques and design bases established in reports that have been accepted by the USNRC (see refer-ences).

Cycle 2 and 3 ceactor parameters related to power capability are summarized briefly in section 5. All accidents analyzed in the FStR have been reviewed for cycle 3 operation. In cases where cycle 3 characte-istics were conserva-tive compared to those analyzed for previous cycles, no new accident analyses were performed.

The Technical Specifications have been reviewed, and the modifications required for cycle 3 operation are justified in this report. Based on the analyses and considering the postulated effects of fuel densification and the Final Acceptance Criteria for Emergency Core Cooling Systems, AN0-1 can be operated safely during cycle 3 at the rated core power level of 2568 MWt.

Babcock a.Wilcox

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2. OPERATING HISTORY 4

The reference fuel cycle for.the nuclear and thermal-hydraulic analyses of the third cycle of ANO-1 is the currently operating cycle 2. Cycle 2 achieved initial criticality on April 2, 1977, and power escalation began on April 5, 1977. The 100% power level of 2568 MWt was reached on April 9, 1977. No

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operating anomalies have occurred during cycle 2 operation that would adverse-ly affect the fuel performance in cycle 3 during the design length of 294 10 EFPD. No control rod interchanges are planned for cycle 3. Control rod group 7 will be withdrawn at 250 110 EFPD of operation.

Babcock & Wilcox

3. GENERAL DESCRIPTION The Arkansas Nuclear One, Unit 1 (ANO-1), reactor core is described in detail in Chapter 3 of the FSAR.1 The cycle 3 core comprisee 177 fuel assemblies, each of which is a 15 by 15 array containing 208 fuel rods,16 control rod guide tubes, and one incore instrument guide tube. The fuel rod cladding is cold-worked Zircaloy-4 with an OD of 0.430 inch and a wall thickness of 0.0265 inch. The fuel is dished-end, chamfered, cylindrical pellets of uranium diox-ide which are approximately 0.370 inch in diameter, 0.700 inch in length for batches 1 and 3, and 0.600 inch in length for batches 4 and 5. (See Tables 4-1 and 4-2 for additional data.) All fuel assemblies in cycle 3 maintain a nominal fuel loading of 463.6 kg of uranium. The undensified nominal active fuel lengths and theoretical densities vary slightly between batches. Specific values are given in Tables 4-1 and 4-2.

Figure 3-1 is the core loading diagram for ANO-1, cycle 3. All the batch 2 assemblies will be discharged at the end of cycle 2. Five once-burned batch 1 assemblies with an initial enrichment of 2.06 wt % 235U will be reloaded into the central portion of the core.

Reuse of these batch 1 fuel assemblies reduces feed batch size and spent fuel storage requirements and results in a more ef-ficient fuel cycle. Batches 3 and 4, with initial enrichments of 3.05 and 2.64 wt % 235 U , respectively, will be shuffled to new locations.

Batch 5, with an initial enrichment of 3.01 wt % 235 U, will occupy the core periphery and eight interior locations. Figure 3-2 is an eighth-core map showing the assembly burnup and enrichment distribution at the beginning of cycle 3.

Reactivity control is supplied by 61 full-length Ag-In-Cd control rods and coluble boron shim. In addition to the full-length control rods, eight parcial-length axial power shaping rods (APSRs) are provided for additional control of exial power distribution. The cycle 3 locations of the 69 control rods and the group designations are indicated in Figure 3-3. The core locations of the total pattern (69 control rods) for cycle 3 are identical to those of the reference cycle indicated in Chapter 3 of the FSAR.I However, the group Babcock a. Wilcox

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designations-differ between cycle 3 and the reference cycle to minimize power l peaking. Neither control rod. interchange nor burnable poison rods are neces- l sary for cycle 3.

1 The nominal system pressure is 2200 psia, and the core average densified nomi-nal heat rate is 5.79 kW/ft at the rated core power of 2568 MWt.

i Babcock & Wilcox

a Figure 3-1. Core Loading Diagram for ANO-1, Cycle 3 n ZL TRMSFER CANAL >

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5 5 5 5 5 0-11 N-8 0-5 5 5 5 3 3 3 5 5 5 P-12 N-10 0-12 N-9 0-8 N-7 O-4 N-6 P-4 5 4 3 3 4 3 4 3 3 4 5 N-14 0-13 P-11 P-10 L-9 L-7 P-6 P-5 0-3 N-2 3

5 4 4 4 4 3 5 3 4 4 4 4 5 L-12 M-14 H-15 R-10 R-9 K-7 R-7 R-6 R-8 M-2 L-4 E

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5 3 4 4 4 4- 3 4 4 4 4 3 5 N-13 L-14 L-15 N-12 N-ll M-5 N-5 N-4 L-1 L-2 N-3 F (Cyl) (Cyl) 5 5 3 4 4 1 3 3 3 1 4 4 3 5 5 M-13 K-12 K-10 ,K-15 M-12 P-8 M-4 K-1 K-6 K-4 M-3 C

5 3 4 3 4 3 5 3 5 3 4 3 4 3 5

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, H-12 H-13 K-9 M-ll B-14 M-13 H-2 E-5 G-7 H-3 E-4 E (Cyl) T E 5 3 3 5 3 3 3 1 3 3 3 5 3 3 5 E-13 i G-12 G-10 G-15 E-12 B-8 E-4 G-1 G-6 G-4 E-3 K

5 3 4 3 4 3 5 3 5 3 4 3 4 3 5

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D-13 F-14 F D-12 D-11 E-ll D-3 D-4 F-1 T-2 D-3 L (Cyl) (Cyl) 5 5 3 4 4 1 3 3 3 1 4 4 3 5 5

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F-12 E-14 A-8 A-10 A-9 G-9 A-7 A-6 H-1 E-2 F-4 M .

5 3 4 4 4 4 3 4 4 4 4 3 5 D-1= C-13 B-11 B-10 7-9 F-7 B-6 B-5 C-3 D-2

,5 4 4 4 4 3 5 3 4 4 4 4 5 B-12 >10 C-12 D-9 C-8 D-7 C-4 D-6 B-4 0

5 4 3 3 4 3 4 3 3 4 5 C-11 D-8 C-5 P

5 5 5 3 3 3 5 5 5 R

5 5 5 5 5 I I l Z l l 2 3 4 5 6 7 4 9 10 11 12 13 14 15

- Previous Core Location Batch (Cyl): Cycle 1 1

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Babcock & Wilcox

Figure 3-2. Enrichment and Burnup Distribution for ANO-1, Cycle 3 8 9 10 11 12 13 14 15 2.06 3.05 3.05 3.05 3.01 3.05 3.05 3.01 H

14854 20658 20130 23325 0 24737 23508 0 3.01 3.05 2.64 3.05 2.64 3.05 3.01 0 22422 6647 24222 10781 19895 0 2.06 2.64 2.64 3.05 3.01 3.01 L

13084 5719 9524 18414 0 0 t

2.64 2.64 3.05 3.01 M

6825 8713 22026 0 2.64 2.64 3.01 N

6629 6181 0 3.01 0

0 1

P R

I.XI Initial Enrichment IIIIX BOC Burnup, mwd /mtU Babcock & Wilcox

Figure 3-3. Control Rod Locations for ANO-1, Cycle 3 I

I A

3 4 7 4 '

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

7 8 4 i c 5 2 3 3 2 l

5 V- 7 3 6 5 6 -Y 3 7 K 5 2 3 3 2 5 l

L 4 8 7 6 7 8 4 1 g 1 3 2 2 3 1 i N 6 8 3 8 6 i

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o 1 5 5 1 y 4 7 4 1

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1 2 3 4 5 6 '7 8 10 11 9 12 13 14 15 Group Number of Rods Function 1 8 Safety I Group Number 2 8 Safety 3 12 Safety 4 8 Safety 5 9 Control 6 8 Control 7 8 Control

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8 _,8_ APSRs Total 69 Babcock & Wilcox

4. FUEL SYSTEM DESIGN 4.1. Fuel Assembly Mechanical Design The types of fuel assemblies and pertinent fuel design parameters for ANO-1, cycle 3 are listed in Table 4-1. All fuel assemblies are identical in concept and are mechanically interchangeable. The reload fuel assemblies incorporate minor design modifications to the spacer grid corner cells which reduce spacer grid interaction during handling. In addition, improved test methods (dynamic impact testing) show that the spacer grids have a higher seismic capability and thus an increased safety margin over the values reported in reference 2.

All results, references, and identified conservatisms presented in section 4.1 of the previous ANO-1 reload report 3 are applicable to the cycle 3 reload core.

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

4.2.1. Cladding Collapse Creep collapse analyses were performed for three-cycle assembly power histories for ANO-1. The batch 3 fuel is more limiting than the other batches because of its previous incore exposure time. The batch 3 assembly power histories were analyzed and the most limiting assembly determined.

The power history for the most limiting assembly was used to calculate the fast neutron flux level for the energy range above 1 MeV. 'The collapse time for the most limiting assembly was conservatively determined to be more than 30,000 EFPH (effective full power hours), which is longer than the maximum projected three-cycle exposure life of 25,584 hours0.00676 days <br />0.162 hours <br />9.656085e-4 weeks <br />2.22212e-4 months <br /> (see Table 4-1). The creep collapse analysis was performed based on the conditions set forth in references 3 and 4.

Babcock & )Milcox

4.2.2. ' Cladding Strese The batch 1B reinserted fuel and the batch 3 fuel are the most limiting for cladding stress. The results presented in the ANO-1 fuel densification re-ports are applicable.

, 4.2.3. Cladding Strain The f.uel design criteria specify a limit of 1.0% on cladding plastic circum-ferential strain. The pellet design is established for a plastic cladding strain of less than 1% at values of maximum design local pellet burnup and heat generation rate, which are considerably higher than the values the ANO-1

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fuel is expected to see. This will result in an even greater margin than the analysis demonstrated. The strain analysis is also based on the maximum al-lowable value for the fuel pellet diameter and density and the lowest permitted tolerance for the cladding ID.

4. 3. Thermal Design All fuel assemblies in this core are thermally similar. The fresh batch 5 fuel inserted for cycle 3 operation introduces no significant differences in fuel thermal performance relative to the other fuel remaining in the core.

The design linear heat rate (LHR) capability and average fuel temperature for each fuel batch in cycle 3 are shown in Table 4-2. LHR capabilities are based on centerline fuel melt and were established using the TAFY6 code with fuel densification to 96.5% of theoretical density. As described in the cycle 2 reload report 3, a design LHR limit of 19.4 kW/ft was used in cycle 3; there-fore, the selective fuel loading requirements are the same as for cycle 2.

The densification power spike model for cycle 3 is the same as that described in the cycle 2 reload report3 using the conservative combination of initial density and enrichment to calculate the spike factor.

4.4. Material Design The chemical compatibility of all possible fuel-cladding-coolant-assembly interactions for the batch 5 fuel assemblies is identical to that of the present fuel.

4.5. Operating Experience Babcock & Wilcox operating experience with the Mark B 15 by 15 fuel assembly has verified the adequacy of its design. As of June 30, 1977, the following Babcock & )Milcox

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

Current cycle Cumulative Current max i ably net electrical Reactor cycle burnr _ 'd /mtU output. MWe ,

Oconee 1 3 26,300 17,559,861 Oconee 2 2 26,800 13,461,166

-Oconee 3 2 24,900 13,015,418

.Three Mile Island Unit 1 3 24,500 14,120,220 ANO Unit 1 2 22,500 10,898,144 Rancho Seco 1 17,200 7,306,304 -

Crystal River 3 1 2,300 1,257,177 I

i Babcock & iMilcox

Table 4-1. Fuel Design Parameters and Dimensions Twice-burned *** ""

assemblies, Once-burned assemblies assemblies, Batch 3 Batch IB Batch 4 Batch 5 Fuel assembly type Mark B3 Mark B3 Mark B4 Mark B4 No. of assemblies 60 5 56 56 Fuel rod OD (nom.), in. 0.430 0.430 0.430 0.430 Fuel rod ID (nom.), in. 0.377 0.377 0.377 0.377 Flexible spacers, type Corrugated / Corrugated / spring Spring Spring spring Rigid spacers, type Zr02 Zroz Zr-4 Zr-4 Undensified active fuel 141 144 142.6 142.25

, length (nom.), in.

p -Fuel pellet OD (mean 0.3670 0.3700 0.3700 0.3695

, specified), in.

Fuel pellet initial density, 96.0 92.5 93.5 94.0

% TD (nom.)

Initial fuel enrichment, 3.05 2.06 2.64 .3.G1 wt %.235U Burnup, BOC, mwd /m.:U 21,748 13,438 7,752 0 Cladding collapse t ime, >30,000 >30,000 >30,000 >30,000 EFPH

,, Estimated residence time, 25,584 19,056 20,544 20,976 m EFPH 8

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h ble 4-2. -Fuel Thermal Analysis Parameters Batch 1 Batch 3 Batch 4 Batch 5 No. of assemblies 5 60 56 56 Initial density. % TD 92.5 96.0 (a) 93.5 94.0 .

Pellet diameter, in.- 0.370 0.367(a) 0.370 0.3695 Stack height, in. 144 141(a) 142.6 142.25 Densified Fuel Parameters Pellet diameter, in. 0.3632 0.3646 0.3645 0.3646 Fuel stack height, in. 141.1 140.65 140.49 140.5 2

_ Nominal linear heat rate 5.77 5.79 5.80 5.80 at 2568 MWt, kW/ft Average fuel temperature 1330 1330(b) 1315 1320 at nom. LHR, F Linear heat rate capability 20.1(c) 19.40(d) 20.15 (c) 20.15(c)

(centerline fuel melt), kW/ft

("}Nopinal values af ter resintering.

(b) Based on batch 1 fuel, which is more limiting than batch 3.

Minimum capability based on fuel specification.

(d) Design limit. -The LHR capability of each fuel assembly in batch 3 was determined from resinter data.7 Babcock & Wilcox

5. NUCLEAR DESIGN 5.1. Physics Characteristics Table 5-1 compares the core physics parameters of cycles 2 and 3; the values for both cycles were generated using PDQ07. Since the core has not yet reached an equilibrium cycle, differences in core physics parameters are to be expected between the cycles. The longer cycle 3 will produce a larger cycle differen-tial burnup than that for cycle 2. The accumulated average core burnup will be higher in cycle 3 than in cycle 2 because of the presence of once- and twice-burned f uel. Figure 5-1 illustrates a representative relative power dis-tribution for the beginning of the third cycle at full power with equilibrium xenon and normal rod positions.

The critical boron concentrations for cycle 3 are given in Table 5-1. As in-dicated in Table 5-2, the control rod wort.hs are sufficient to maintain the required shutdown margin. The BOC hot, full-power control rod worths are com-parable to those of cycle 2. The cycle 3 ejected rod worths are slightly higher than those in cycle 2 for the same number of regulating banks inserted.

It is difficult to compare values between cycles or between rod patterns since neither the rod patterns from which the CRA is assumed to be ejected nor the isotopic distributions are identical. Calculated ejected rod worths and their adherence to criteria are considered at all times in life and at all power levels in the development of the rod insertion limits presented in section 8.

The maximum stuck rod worths for cycle 3 are greater than those in cycle 2.

The adequacy of the shutdown margin using cycle 3 stuck rod worths is demon-strated in Table 5-2. The following conservatisms were applied for the shut-down calculations:

1. Poison material depletion allowance.
2. 10% uncertainty on net rod worth.
3. Flux redistribution penalty.

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! Babcock & )Milcox

Flux redistribution was accounted for since the shutdown an,alysis was calcu-laced using a two-dimensional model. The shutdewn calculation at the end of cycis '

la analyzed at approximately 250 EFPD. This is the latest time (110 days) in core life at which the transient bank is nearly fully inserted. After 250 EFPD, the transient bank will be almost fully withdrawn, thus increasing the available shutdown margin.

The cycle 3 power deficits from hot zero power to hot full power are similar to but slightly lower than those for cycle 2. Doppler coefficients, moderator coefficients, and xenon worths are similar for the two cycles. The differen-tial boron worths for cycle 3 are lower than those for cycle 2 due to depletion of the fuel and the associated buildup of fission products. The effective de-layed neutron fractions for both cycles decrease with burnup.

5.2. Analytical Input The cycle 3 incore measurement calculation constants used for computing core power distributionr were prepared in the same manner as for the reference cycle.

5.3. Chantes in Nuclear Design There were no relevant changes in core design between the reterence and reload cycles. The same calculational methods and design information were used to ob-tain the important nuclear design parameters. The only significant operational procedure change from the reference cycle is the specification of APSR position limits in addition to the usual regulating control rod and imbalance limits for ECCS. The APSR position limits will provide additional control of power peak-ing and assurance that LOCA kW/ft limits are not exceeded. The operational limits (Technical Specification changes) for the reload cycle are given in sec-tion 8.

Babcock 8.Wilcox

Table 5-1. ANO-1. Cycle 2 and 3 Physics Parameters Cycle 2(* Cycle 3 b' Cycle length, EFPD 272 294 Cycle'burnup, mwd /mtU 8512 9200 Average core burnup, EOC, mwd /mtU 18615 19405 Initial core loading, mtU 82.1 82.1 Critical boron, BOC (no Xe), ppm HZP,("} group 8 37.5% wd 1315 1318 HZP,' groups 7 and 8 inserted 1195 1248 HFP, groups 7 and 8 inserted 1000 1065 Critical boron, EOC (eq Xe), ppm HZP, group 8 37.5% wd, eq Xe 410 302 HFP, group 8 37.5% wd, eq Xe 200 36 Control rod worths, HFP( BOC, %Ak/k Group 6 1.05 1.07 Group 7 0.92 0.77 Group 8 (37.5% wd) 0.46 0.42 Control rod worths, HFP (250 EFPD), %Ak/k Group 1 1.20 1.05 Group 6 -(37.5% wd) 0.47 0.45 Max ejected rod worth, HZP, %Ak/k BOC 0.64 250 EFPD 0.

0.5354d)((d) 0.57 Max stuck rod worth, HZP, %Ak/k BOC 1.94 2.50 250 EFPD 1.92 2.22 Power deficit, HZP to HFP, %Ak/k BOC 1.60 1.27 EOC 2.20 2.03 Doppler coeff, BOC, 10-5 Ak/k/*1!

100% power (no Xe) -1.47 -1.47 Doppler coeff, EOC, 10-5 Ak/k*F 100% power (eq Xe) -1.51 -1.61 Moderator coeff, HFP,10-4 Ak/k'F BOC-(0 Xe, 1150 ppm, group 8 ins) -1.09 -0.66 EOC (eq Xe,17 ppm, group 8 ins) -2.66 -2.62 Boron worth, HFP, ppm /%Ak/k BOC (1000 ppm) 110 107 EOC (17 ppm) 99 97 Babcock & Wilcox

i Table 5-1. (Cont'd)

Cycle 2("} Cycle 3( )

Xenon worth, HFP, %4k/k BOC (4 days) ' 2.61 2.66 -

EOC (equilibrium) 2.64 2.75 Effective delayed neutron fraction, HFP BOC 0.035814 0.005909 EOC 0.005280 0.005239

("} Based oa a cycle 1 length of 490 EFPD.

( ) Cycle 3 data are for the condition stated; cycle 2 data may not -

be at the same conditions as cycle 3 (see reference 3 for cycle 2 core conditions).

~ " HZP: hot zero power; HFP: hot full power.

(

Ejected rod value for groups 5, 6, 7, and 8 inserted.

l 1

l Babcock & Wilcox '

Table 5-2. Shutdown Margin Calculation, ANO-1, Cycle 3 Available Rod Worth BOC, %Ak/k EOC(* %Ak/k Total rod worth, HZP(b) 8.83 8.66 Worth reduction due to burnupLof -0.30 -0.35 poison material Maximum stuck rod, HZP -2.50 -2.22 Net worth 6.03 6.09 Less 10% uncertainty -0.61 -0.61

. Total available worth 5.42 5.48 Required Rod Worth Power deficit, HFP to HZP 1.27 2.03 Maximum allowable inserted rod 0.99 0.96 worth Flux redistribution 0.40 0.85 Total required worth 2.66 3.84 Shutdown Margin Total available worth minus 2.76 1.64 total required worth Note: Required shutdown margin is 1.00% ok/k

(* For shutdown ma_ gin calculations, this is defined as about 2501:FPD, the latest time in core life at which the transient bank is nearly full-in.

HZP: hot zero power; HFP: hot full power.

Babcock & Wilcox l: . _ _ _ _ _ _ _ _ _ - - _ _

Figure 5-1. BOC 3 (4 EFPD) Two-Dimensional Relative Power Distribution - Full Power, Equilibrium Xenon, Normal Rod Positions 8 9 10 11 12 13 14 15 4

H 0.95 1.08 0.99 1.02 1.39 0.83 0.44 0.66 1.08 1.40 0.94 1.14 0.99 0.97 0.80 0.75 a

7 8 b

0.99 0.94 0.58 1.17 1.02 1.02 1.25 0.73 M

, 1.03 1.14 1.17 1.32 1.25 1.03 1.09 N 1.40 1.00 1.03 1.26 1.33 1.11 0.77 0

5 0.84 0.97 1.02 1.03 1.11 0.87 0.44 0.81 1.25 1.09 0.77 i_

r 0.67 0.76 0.73 Rod groups 7 and 8 inserted.

X Inserted Rod Group Number

-I.XX Relative Power Density Babcock & Wilcox

6. THERMAL-HYDRAULIC DESIGN The incoming batch 5 fuel is hydraulically and geometrically similar to batch 4 fuel. The cycle 2 and 3 maximum design conditions and significant parameters are shown in Table 6-1. The only change in the thermal-hydraulic evaluation for cycle 3 is the removal of the densification power spike from DNBR calcula-tions.8 This results in an increase in the minimum calculated steady-state DNBR from 1.84 for cycle 2 to 1.90 for cycle 3.

The potential effect of fuel rod bov on DNBR was considered by incorporating suitable margins into DNB-limited core safety limits and RPS setpoints. The maximum rod bow magnitude was calculated using the NRC interim model, AC/C, =

0.065 + 0.001449/BU, where AC is the rod bow magnitude (mils), C, is the ini-tial gap, and BU is the assembly burnup (mwd /mtU). The resultant DNBR penalty, based on the maximum predicted assembly burnup at EOC 3, regardless of batch, is 11.2%. The variable low-pressure trip and the flux / flow trip setpoint of 1.060 provides the thermal margin required for the 11.2% DNBR rod bow penalty.

Babcock 8.Wilcox

Table 6-1. Cycle 2 and 3 Maximum Design Conditions Cyclel Cycle 3 Design power level, MWt 2568 2568 System pressure, psia 2200 2200 RC flow, % design flow 106.5 106.5 Vessel inlet coolant temperature 555.6 555.6 at 100% power, F Vessel outlet' coolant temperature 602.4 602.4 at 100% power, F Reference design axial flux shape 1.5 cosine 1.5 cosine Hot channel factors -

Enthalpy rise _1.011 1.011 Heat flux 1.014 1.014 Flow area 0.98 0.98 Average heat flux at 100% power, 175,205 175,427 Btu /h-ft2(a)

CHF correlation B&W-2 B&W-2 Minimum DNBR (% power)I ) 1.84 (112) 1.90 (112)

(*) Cycle 2 heat flux was based on batch 3 densified length; cycle 3 uses batch 5 densified length (located in hottest core loca-tion).

(

Cycle 2 DNBR included effects of densification power spike; )

cycle 3 does not, l l

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Babcock & Wilcox J J

7. ACCIDENT AND TRANSIENT ANALYSIS l

7.1. General Safety Analysis Each FSARI accident analysis has been examined with. respect to changes in cycle 3 parameters to determine the effect of the cycle 3 reload and to ensure that thermal performance during hypothetical transients is not degraded.

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

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

Core thermal properties used in the FSAR accident analysis were design operat-ing values based on calculational values plus uncertainties. First-core values (FSAR values) of core thermal parameters and subseq;ent fuel batches are com-pared to parameters rised in cycle 3 analyses in Table 4-2. The cycle 3 thermal-hydraulic maximum design conditions are compared to the previous cycle 2 values in Table 6-1. These parameters are common to all the accidents considered in this report. A comparison of the key kinetics parameters from the FSAR and cycle 3 is provided in Table 7-2.

A generic LOCA analysis for a B&W 177-FA, lowered-loop NSS has been performed using the Final Acceptance Criteria ECCS Evaluation Model (reported in BAW-101039). The analysis in BAW-10103 is generic since the limiting values of key parameters for all plants in this category were used. Furthermore, the combination of average fuel temperature as a function of LHR and lifetime pin pressure data used in the BAW-10103 LOCA limits analysis is conservative Babcock & Wilcox

compared to those calculated for this reload. Thus, the analysis and the LOCA limits reported in BAW-10103 and substantiated by reference 10 provide conser-vative results for the operation of the reload cycle.

Table 7-1 shows the bounding values for allowable LOCA peak LHRs for ANO-1 cycle 3 fuel.

It is concluded from the examination of cycle 3 core thermal and kinetics prop-erties, with respect to acceptable previous cycle values, that this core re-load will not adversely affect the ANO-1 plant's ability to operate safely during cycle 3. Considering the previously accepted design basis used in the FSAR and subsequent cycles, thc transient evaluation of cycle 3 is considered to be bounded by previously accepted analyses. The initial conditions for the -

transients in cycle 3 are bounded by the FSAR I , the fuel densification reports, and/or subsequent cycle analyses.

Table 7-1. Bounding Values for Allowable LOCA Peak Linear Heat Rates Core Allowable peak linear elevation, ft heat rate, kW/ft 2 15.5 4 16.6 6 18.0 -

8 17.0 10 16.0 Babcock & \Nilcox

Table'7-2. Comparison of Key Parameters for Accident Analysis FSAR and densification Predicted Parameter report value value Doppler coeff (BOC), Ak/k/*F -1.17 x 10

~ ~

-1. 4 7 x 10 '

Doppler coeff (EOC), Ak/k/*F -5

-1.30 x 10 -1.61 x 10~

Moderator coeff (BOC), Ak/k/*F 0.0(*) -

-0.66 x 10 "

Moderator coeff (EOC), ak/k/*F -4.0 x 10- (b) ~

-2.62 x 10 All-rod group worth, % ok/k 12.9 8.83 Initial boron conc, ppm 1150 1070 Boron reactivity worth (HFP), 100 107 pps/1% Ak/k Max ejected rod worth, % Ak/k 0.65 0.47 Dropped rod worth (HFP), % ak/k 0.65 0.20

~

(*)+0.5 x 10 " ak/k/*F was used for the moderator dilution analysis.

( ~

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

Babcock & Wilcox

N

8. PROPOSED MODIFICATIONS TO TECHNICAL SPECIFICATIOiiS The Technical Specifications have been revised for cycle 3 operation. Changes were the results of.the following:
1. Specifying APSR. position limits in addition to the usual regulating control rod and imbalance limits for ECCS. The APSR position limits will provide additional control of power peaking and assurance that LOCA kW/ft limits are not exceeded.
2. The Technical Specification limits based on DNBR and LHR criteria include appropriate allowances for projected fuel rod bow penalties, i.e., poten-tial reduction in DNBR and increase in power peaks. A statistical combi-nation of the nuclear uncertainty factor, engineering hot channel factor, and rod bow peaking penalty was oced in evaluating LHR criteria, as ap-proved in reference 11. s
3. -Per reference 8, the power spike penalty due to fuel densification was not used in setting the DNBR- and ECCS-dependent Technical Specification limits.
4. The allowable quadrant tilt limit for cycle 3 is 4.92%. This is a result of the lower power peaks associated with the third cycle of operation ver-aus those of the second cycle and the use of the procedures discussed in references 8 and 11.

Based on the-Technical Specifications derived from the analyses presented in this report, the Final Acceptance Criteria ECCS limits will not be exceeded, nor will the thermal design criteria be violated. Figures 8-1 through 8-11 illustrate revisions to previous Technical Specification limits; Figures 8-12 through 8-14 illustrate limits not previously included in the Technical Speci-fications.

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l Babcock s. Wilcox

a Figure 8-1. ANO-1 Cycle 3 Core Protection Safety Limits Thermal Power Level, ".

UNACCEPTABLE OPERATION I****)

I ""0 '_s i 2 ) 112 _

1 ACCEPTABLE

[4PUNP OPERATION _ J 00 85.6

(-70,80) (-40,85.6) (26,85.6 2 --80 (u,80)

ACCEPTABLE 3 & 4 PUMP

(-64,64)

_ 58.6 - 60 _

f(-40,58. 6) (26,58.6 (55,55)

(-48,48)

ACCEPTABLE 2,3 & 4 PUMP OPERATION . . 40 I'3'"3) 2Q i

l 1

-80 -60 40 -20 0 20 40 60 80 l Power Imaalance, 5 l

l CURVE REACTOR COOLANT FLOW (GPM) l 1 374,880 ,

2 280,035 3 184,441 l

Babcock & Wilcox

Figure 8-2. ANO-1 Cycle 3 Protective System Maximum Allowable Setpoints Thermal Power Level, ",

120 -

,0 (-19,106) 1 06 (35.308)

+', 100- -

+

(.26,99) ACCEPTABLE 4 y PUMP OPERATION *#

4

(-19,79.1) 80_.7g.j (42,80)

.(ig,73,i) ,

(-25,72.1) e ACCEPTABLE 314 PUMP L'P E R A T I O N 60.

(-19,52.1) 52.1 ,

(14,52.1) ,(42,53.2)

(-26,45.1),

40._

ACCAPTABLE 2,3

&4 PUMP OPERATION p(42,26.I) 20 -

2

  • I t i i

+ +

11 11 gg gg

=

~

E  :

, i . , , , , , , ,

50 -40 -30 -20 -10 10 0 20 30 40 50 Power imoalance, ",

- 26 Babcock 8. Wilcox

Figure 8-3. Rod Position Limits for Four-Pun.c Operation From 0 to 100 1 10 EFPD - AliO-1, Gycle 3 110 (113,102) (185.102) o ,(217,102)

(185,92) , (217,92) POWER LEVEL 90 -

NOT A!. LOWED CUT 0FF (92.05) 80 -

(l52,80) '(262,80) g RESTRICTED

=

g 70 -

(300,70) >

~

en-60 S

50 -

(33,50) (95,50) 40 3, _ PERMISSIBLE 20

< (0,15) 10 . /

{ ...

0 ' ' '

O 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 Rod index 5 10 , , , , ,

0 25 50 75 100 Group 7 0 25 50 75 100 Group 6 0 25 50 75 100 Group 5 l

l i

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Figure 8-4. Rod Position Limits for Four-Pet.p Operation From 100 1 10 to 250 10 EFPD - ANO-1, Cycle 3 110

, (185,102) 100 - (183,102) ' o(217,102) g ,

NOT ALLOWE0 (185,92y (217,92)f.0,WEMElEL CUTOFF RESTRICTED (92,05) 80 -

(152,80) (262,80) 4 0 -

g (130,88) (300,70)

M

[O 60 -

4%

! 50 -

(101,50) 2 2 '

PERMISSI8LE 40 -

30 -

20 -

( 'I )

(0.9) 10 1,31)

RESTRICTED #

0 ' ' ' ' - - ' . . .

i .

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

. Rod Index 5 10 0 25 50 75 150

, , e , , Grcup 7 ,

0 25 50 75 100 Group 6 e . . . . 1 0 25 50 75 100 Group 5  ;

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v.

Figure 8-5. Rod Position Limits for Four-Pump Operation After 250 10 EFPD - ANO-1, Cycle 3 7'

110 100 - (1 5,102) og

- ~

. POWE[I LEVEL ,(260,92)

CUT 0FF(92.0%)

0 80 -

,, , NOT ALLOWED (226,80)

U 70 h I ?E *

'g n .

j- 60 ; -

'l g 50 w (117,53)

(14'.30)

)<2 ,

[ 40 h (108,38) 30 p PERMISSISLE I

20 -

(0,8) (92,15) jg RECTRICfE0 h J 20 40 00 80 100 120 140 160 180 200 220 240 260 280 3C0

, Rod index % WD 0 25 50 15 100 Group 7 0 25 50 75 100 Group 6 t_ i f i a

0 23 50

,. 75 100 Group 5

('

l Babcock a, Wilcox

Figure 8-6. Rod Position Limits for Two- and Three-Pump Operation From 0 to 100 t 10 EFPD - ANO-1, Cycle 3 lie ,

municne (2H.le!) f (183,1e2) (152,102)

NOT Allegtg te - (3es,sg)

' (87,es)-

E 1e -

=

. Se -

a se

-( 33,5e) 3 h pgnsp; sing 8

4e -

(45,39) ,

3e 8 2.

(0.15) < /

s' -

RI$iRICilO FOR 2 & 3 ruar . . . . . . . . . . , , ,

s .

9 20 40 68 80 III 129 148 ISO 180 200 220 240 280 200 300 Rod lades (5 WO) a e 9 25 50 15 100 Ereup T e 25 la il 'i os 8"8 8 0 25 50 15 100 stem, 5 Babcock a.Wilcox

Figure 8-7. Rod Position Limits for Two- and Three-Pump Operation From 100 10 to 250 10 EFFD -

ANO-1, Cycle 3 lio (1 ,102) 100 .

90 - NOT ALLOWE0

80 -

3 2

2 70 -

2

, 60 -

! 50 -

(101,50)

PERhllSSIBl.E 40 -

30 -

20 (20,11) (72,15) 10 , -

(0,9 r RESTRICTED p) ' '

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

0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 Rod index (5 WD) e t t t n 0 25 53 75 100 Group 7 0 25, 50 75 100 Group 6 0 25 50 75 - 150 Group 5 Babcock & Wilcox

? __.

1 Figure 8-8. Rod Position Limits for Two- and Three-Pump Operation After 250 10 EFPD - ANO-1, Cycle 3 110 NOT ALLOWE0 (195,102) (228,102) 100 -

90 -

n 80 a

2 5

70 -

~

% 141,64) PERNIS$1BLE

  • 60.

E g 50 -

(117,50) 40 -

RESTRICTED FOR 2 & 3 30 -

PUNP OPERATION 0

(100,19)

(70,13 (92,15) 10 ,-

(0,8) 0 . . , , . .

0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 Rod Index (5 WO) 0 25 50 75 100 Group 7 0 25 50 75 100 Group 6 0 25 50 75 100 Group 5 l

l l

Ba' bcock & Wilcox

l Figure 8-9. Operational Power Imbalance Envelope for Operation '

From 0 to 100 10 EFPD - ANO-1, Cycle 3 POWER (5 0F 2568 MWt)

-20.1,102  ; q +7. 8,102 100 - -

-21.0,92 .1,92 90 - -

-26.7,80 '

80 - -

> +10. 7. 80 70 - -

60-, -

PERMIISIBLE OPERATING REGION RESTRICTED RESTRICTED 4 -

REGION REGION

, 30 - -

20 - -

10 - -

i i I t i 1. I I I I

-50 -40 -30 -20 -10 0 10 20 30 40 50 Axial Pawer Imaalance (%)

i c

- 33 , Babcock & Wilcox

Figure 8-10. Operational P .sr Imbalance Envelope for Operation From 100 2 ,, so 250 1 10 EFPD - ANO-1, Cycle 3 POWER (5 of 2568 U t)

-16.9,102 - '

100 --

-23.7,92 ,+9.1,92 g _,

-26.7,80 < 80 - - ,

, +10.7,80 70 - -

60 - -

l RESTRICTED PERMISSIBLE REG l0N OPERATING RESTRICTED REGION REGION 40 --

I l

i l

30 --

0.0 --

n -

t I I I I t I I l l

-50 -40 -30 -20 -10 0 10 20 30 40 50 Axial Power imbalance 1

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1 Babcock & Wilcox l

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Figure 8-11. Operational Power Imbalance Envelope for Operation After 250 i 10 EFPD - ANO-1, Cycle 3 POWER (5 of 2568 MWt) 1

-27.3,102 - +13.4.102

-33.3,92

-- 90 4

-36.3,80 -- 80 ,+16.6,80

-- 70

+ 60 PERNISSIBLE OPERATING RESTRICTED RESTRICTED REGION REGION REGION 40 30

-- 20

-- 10 I i t i t i I e i I

-50 -40 -30 -20 -10 0 10 20 30 40 50 Axial Power Imaalance (5) l I

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a Figure 8-12. APSR Position Limi:s for Operation From 0 to 100 10 EFPD - ANO-1, Cycle 3

, p 35.0,102 RESTRICTED REGION 90 - 35.5,92 43.5,80 C

m 70 -

100,70

~

S0 a

w

,- 50 -

E 40 -

30 20 PERMISSIBl.E OPERATING 10 REGION 0 ' ' ' ' ' ' ' ' '

O 10 20 30 '40 50 60 70 80 90 100 APSR, ", Wi tnd tawn l

l Babcock & Wilcox

Figure 8-13. APSR Position Limits for Operation From 100 10 to 250 10 EFPD - ANO-1, Cycle 3 29.0,102 RESTRICTED 100 -

REGION 33.0,92 90 -

43.5,80 G

as 1

== 70 -

l 100,70 E 60 -

w j 50 -

2 40 30 -

20 . PERMISSIBLE OPERATING REGION 10 -

0 , , , , , , , , ,

0 10 20 30 40 50 60 70 80 90 100 APSR, ", Witndrawn l

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Figure 8-14. APSR Position Limits for Operation After 250 10 EFPD - ANO-1, Cycle 3 4.0,102 100 - RESTRICTED REGION 90 -

61.0,80 80 -

G

=

= 70 -

W 100,70 a 60 -

m

~

g 50 -

PERillSSIBl.E OPERATING REGION 40 -

30 -

2C -

10 O i e i i i , e i ,

0 10 20 30 40 50 60 70 80 90 100 APSR, 5 Withdrawn

(

Babcock & Wilcox

0. STARTUP PROGRAM - PHYSICS TESTING The planned startup test program associated with core performance is outlined below. These tests verify that uore performance is within the assumptions of the safety analysis and provide the necessary data for continued safe operation.

Precritical Tests

1. Control rod trip test.

Zero Power Physics Tests

1. Critical boron concentration.
2. Temperature reactivity coefficient.
a. All regulating groups out. .
b. Regulating groups inserted.
3. Regulating control rod group reactivity worth.
4. Ejected control rod reactivity worth.

Power Tests

1. Core power distribution verification at approximately 40, 75, and 100%

full power with normal control group configuration.

2.

Incore versus out-of-core detector imbalance correlation verification at approximately 75% full power.

3. Power Doppler reactivity coefficient at epproximately 100% full power.
4. Temperature reactivity coefficient at approximately 100% full power.

- Babcock 8.Wilcox

10. REFERENCES 1

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

2 Fuel Assembly Stress and Deflection Analysis for Loss-of-Coolant Accident and Seismic Excitation, BAW-10035, Babcock & Wilcox, Lynchburg, Virginia, June 1970.

3 Arkansas Nuclear One, Unit 1 - Cycle 2 Reload Report, BAW-1433, Babcock &

Wilcox, Lynchburg, Virginia, November 1976.

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

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

& Wilcox, Lynchburg, Virginia, June 1973.

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

7 B&W Report, " Proprietary Letter Report on Classification and Selective Loading of Fuel for Arkansas Nuclear One, Unit one," Babcock & Wilcox, Lynchburg, Virginia, November 1973.

8 K. E. Suhrke (B&W) to S. A. Varga (USNRC), Letter, "Densification Power Spike," December 6, 1976.

9

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

& Wilcox, Lynchburg, Virginia, September 1975.

10 J. H. Taylor (B&W Licensing) to R. L. Baer, (Reactor _ Safety Branch, NRC)

Letter, June 8,1977.

11 S. A. Varga (NRC) to J. H. Taylor (B&W), Letter, " Comments on B&W's - Sub-4 mittal on Combination of Peaking Factors," May 13, 1977.

{

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Babcock s.Wilcox  !

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