ML20210C154

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Cycle 8 Reload Rept
ML20210C154
Person / Time
Site: Arkansas Nuclear Entergy icon.png
Issue date: 08/31/1986
From:
BABCOCK & WILCOX CO.
To:
Shared Package
ML20210C134 List:
References
BAW-1918, NUDOCS 8609180257
Download: ML20210C154 (65)
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Text

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BAW-1918 August 1986 s

ARIGNSAS NUCIEAR ONE, UNIT 1

- Cycle 8 Reload Report -

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! BABCOCK & WIIDOX l Nuclear Power Division P. O. Box 10935 Lynchburg, Virginia 24506-0935 8609180257 860910 ADOCK 05000313 Sabcock&Wilcox PDR PDR a McDermott company P

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INIRODDC1'ICN AND SG99JE . . . . . . . . . . . . . . . . . . . . . 1-1

2. OPERATING HISIU M

........................ 2-1

3. GENERAL IESCRIPTICH

....................... 3-1 4.

PUEL SYSIEM IESIGN . . . . . . . . . . . . . . . . . . . . . . . . 4-1 4.1. Ibal Assembly Mechanical Design .............. 4-1 4.2. Puel Bod Design . ..................... 4-1 4.2.1. MaMb ig Collapse 4.2.2. Cladding Stress

................. 4-2 4.2.3. cladding strain .........

.................. 4-2 4.3.

. . . . . . . . 4-2

'Ihermal Design . . . . . . . . . . . . . . ........ 4-2 4.4. Material Design ............... 4-3 4.5. Operating Experience . . . . . . . . . . . . ....... ........ 4-3 5.

NUCIDR IESIGN . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1 5.1. Physics Characteristics ................. 5-1 5.2. Analytical Irput . . . . . . . . . . . . . . . . . . . . . .

5.3. 5-1 Changes In Nuclear Designs . . . . . . . . . . . . . . . . 5-2 6.

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'IHERMMrHYIRAULIC IESIGN . . . . . . . . . . . . . . . . . . . . . 6-1

7. NY'TTMfr AND 'IRANSIINE ANM2 SIS ................. 7-1 7.1. General Safety Analysis .................. 7-1 7.2. Accident Evaluation .................... 7-2
8. PROFCSED KOIFICATICNS 'IO TECHNICAL SPECIFICATICNS . .

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...... 8-1 l 9. SIARIUP PROGRAM - RfYSICS 'IESI'ING ................ 9-1 9.1. Precritical Tests ..................... 9-1 9.1.1. Cbntrol Ibd Trip Test ............... 9-1 9.2. Zero Power Physi. Tests . . . . . . . . . . . . . . . . . . 9-1 9.2.1.

9.2.2.

Critical Boru1 Cs =iunticri . . . . . . . . . . . . 9-1 Terperature Reactivity Cbefficient . . . . . . . . . 9-2 9.2.3. Cbntrol Rod Group Reactivity Worth . . . . . . . . . 9-2 9.3. Power Escalation Tests . . . . . . . . . . . . . . . . . . . 9-3 9.3.1. Core Power Distribution Verificatical at s40 and 100% FP With Ncninal Ct:1 trol Rod Position . . . . . . . . . . . . . . . . . . . . 9-3 111 Sabcock &WIfcom a McDermott company

a 4 Contents (Cont'd)

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9.3.2. Incore Vs Excore Detector Imbalance Correlation Verification ............. 9-4 9.3.3. 'Ibmperature Reactivity Coefficient at %100% FP .................... 9-4 9.3.4. Power Doppler Reactivity Coefficient at N100% FP .................... 9-4 9.4. Procedure for Use if Acceptance Criteria Not Met . . . . . 9-5

10. REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . 10-1 List of Tables Table 4-1. Ebel Design Parameters and Dimensions ............. 4-4 4-2. Ebel 'Ihermal Analysis Parameters . . . . . . . . . . . . . . . . . 4-5 5-1. Ihysics Parameters for ANO-1, Cycles 7 and 8 . . . . . . . . . . 5-3 5-2. Shutdown Margin Calculations for ANO-1, Cycle 8 ........ 5-5 6-1. Maximum Design Concitions, Cycles 7 and 8 ........... 6-2 7-1. Comparison of FSAR and Cycle 8 Accident Doses ......... 7-4 7-2. Camparison of Key Parameters for Accident Analysis . . . . . . . .

7-5 7-3. Bounding Values for Allowable IDCA Peak Linear Heat Rates . . . 7-5 List of Ficures Figure 3-1. Core imHrg Diagram for ANO-1, Cycle 8 ............ 3-3 3-2. Enrichment and Burnup Distribution, ANO-1 Cycle 8 Off 425 EFPD Cycle 7 . . . . . . . . . . . . . . . . . . . . . . . . 3-4 l 3-3. Centrol Eod Iocations and Group Designations for l ANO-1, Cycle 8 . . . . . . . . . . . . . . . . . . . . . . . . . 3-5 3-4. IEP Enrid-aent and Distribution, ANO-1, Cycle 8 ... . . . . . 3-6 i 5-1. ANO-1, Cycle ~8, BOC Two-Dimensional Relative Power Distribution - Full Power Equilibrium Xenon, Normal l

Rod Positions . .. ... ................... 5-6 8-1. Boric Acid Addition Tank Volume and Concentration Vs RCS Average 'Ibnperature - ANO-1, Cycle 8 . . . . . . . . . . . 8-7 l

8-2. Rod Position Setpoints for Four-Pump Operation From 0 to 25+10/-0 EFFD - ANO-1, Cycle 8 ................ 8-8

! 8-3. Rod Position Setpoints for Four-Pump Operation Fram 25+10/-0 l

to 200 i 10 EFPD - ANO-1, Cycle 8 . . . . . . . . . . . . . . . 8-9 l 8-4. Rod Ibsition Setpoints for Four-PLmp' Operation From 200 i10 to 380 10 EFPD - ANO-1, Cycle 8 . ............ 8-10 l

8-5. Rod Position Setpoints for Four-Pump Operation After 380

! i 10 EFPD - ANO-1, Cycle 8 .................. 8-11 l

8-6. Rod Position Setpoints for 'Ihree-Pump Operation From 0 to 25

+10/-0 EFPD - ANO-1, Cycle 8 ................. 8-12 l

iv Babcock &Wilcox a McDermott company I _ _ . , - _ _ _ . _ , . - - . , _ _

s i a List of Fiaures (Cont'd) -

Figure Fage 8-7. Rod Position Setpoints for 'Ihree-Pump Operation From 25+10/-0 to 200 ! 10 EFPD - ANO-1, Cycle 8 . . . . . . . . . . . . ... 8-13 8-8. Rod Ibsition Setpoints for 'Ihree-Pump Operation From 200 i 10 to 380 i 10 EFPD - ANO-1, Cycle 8 ...~......... 8-14 8-9. Rod Position Setpoints for 'Ihree-Pump Operation After 380 .

i 10 EFPD - ANO-1, Cycle 8 . .................. 8-15 8-10. Rod Position Setpoints for 'CWo-Pump Operation From 0 to 25+10/-0 EFPD - ANO-1, Cycle 8 ................ 8-16 8-11. Rod Position Setpoints for '3m-Pump Operation Fron 25+10/-0 to 200 i 10 EFPD - ANO-1, Cycle 8 ................ 8-17 8-12. Rod Ibsition Setpoints for 'Bo-Pump Operation From 200 10 380 10' EFPD - ANO-1, Cycle 8 ................ 8-18 8-13. Rod Position Setpoints for 'IWo-Punp Operation After 380 10 EFPD - ANO-1, Cycle 8 . ................. 8-19 8-14. Operational Power Imbalance Setpoints for Operation From 0 to 25+10/-0 EFPD - ANO-1, Cycle 8 . . . ........... 8-20 8-15. Operational Power Imbalance Setpoints for Operation Frcm 25+10/-0 to 200 10 EFPD - ANO-1, Cycle 8 . . . .. . .... 8-21 8-16. Operational Power Imbalance Setpoints for Operation From 200 10 to 380 i 10 EFPD - ANO-1, Cycle 8 . . . ... ... . 8-22 8-17. Operational Power Imbalance Setpoints for Operation After 380 i 10 EFPD - ANO-1, Cycle 8 ................ 8-23 8-18. I.OCA Limited Maxinn Allowable Linear Heat Rate . ... .... 8-24 8-19. APSR Position Setpoints for Operation Frm 0 to 25+10/-0 EFPD - ANO-1, Cycle 8 . . . . . . . . . ............. 8-25 8-20. APSR Position Setpoints for Operation From 25+10/-0 to 200 10 EFPD - ANO-1, Cycle 8 ................ 8-26 8-21. APSR Position Setpoints for Operation Fr m 200 10 to 360 10 EFPD - ANO-1, Cycle 8 . ................. 8-27 8-22. APSR Position Setpoints for Operation After 380 10 EFPD -

ANO-1, Cycle 8 . . . . . . . . ................. 8-28 i

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1. INIRODUC1' ION AND

SUMMARY

s his report justifies the operation of the eighth cycle of Arkansas Nuclear One, Unit 1 (ANO-1) at the rated core power of 2568 NWt. Included are the required analyses as outlined in the USNRC dev m ant, " Guidance for Proposed Idcense AirmiaJumsits Relating to Refueling," June 1975.

To support cycle 8 operation of ANO-1, this report employs analytical techniques and design bases established in rewds that have been submitted to and accepted by the USNRC and its prad mas m r, the USAEC (see references).

Se cycle 7 and 8 reactor parameters related to power capability are summarized briefly in section 5 of this report. All of the accidents analyzed in the ESAR1 have been reviewed for cycle 8 operation. In those cases where cycle 8 characteristics were conservative cotrpared to those analyzed for previous cycles, no new accident analyses were performed.

Se Technical Specifications have been reviewed, and - the modifications required for cycle 8 operation are justified in this report.

Based on the analyses performed, which take into account the postulated effects of fuel densification and the Final Acceptance Criteria for Emergency Core Cooling Systems, it has been concluded that ANO-1 can be op e ted safely for cycle 8 at a rated power level of 2568 MNt.

. We cycle 8 core for ANO-1 will contain one thrice-burned lead test asseltbly l

(LTA) . h is assembly is part of a Department of Energy Extended Burnup Test ,

Program. 'Ihe LTA design is described in reference 2.

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2. OPERATING HIS'IORY

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'Ihe reference cycle for the nuclear and' thamal-hydraulic analyses of Arkansas Nuclear One, Unit 1 is the currently operating cycle 7. 'Ihis cycle 8 design is based on a design cycle 7 length of 425 effective full power days (EFPD).

No anomalies occurred during cycle 7 that would adversely affect fuel perfonnance during cycle 8.

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3. GENERAL DESG1PfION s

he ANO-1 reactor core is described in detail in section 3 of the Arkansas Nuclear One, Unit 1, Final Safety Analysis Report (FSAR).1 he cycle 8 core contains 177 fuel assemblies, each of which is a 15 by 15 array containing 208 fuel rods,16 wutwl red guide tubes, and one incere instrument guide tube. We fuel is ccraprised of dished-end, cylindrical pellets of uranium dioxide clad in cold-worked Zircalcy-4. We fuel asaamblies in all batches have an average nominal fuel loading of 463.6 kg of uranium, with the exception of one batch 7D LTA, which has a nominal loading of 440.0 kg uranium. S e undensified naminal 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 asaamblies except the LTA; the corresponding parameters for the IIIA are included in reference 2.

Figure 3-1 is the fuel shuffle diagram for ANO-1, cycle 8. We initial enrichments of batches 7D, 8B, 9 and 10 are 2.95, 3.21, - 3.30, and 3.35 wt%

U-235, respectively. All but one of the batch 7B assemblies and 28 of the twice-burned batch 8 ===amblies will be discharged at the end of cycle 7.

Se center location will contain the remaining batch 7 assembly (designated 7D) , and the remaining 44 batch 8 assemblies (designated 8B) will be shuffled to new locations, with 12 on the core periphery. Sixty of the 68 once-burned batch 9 ===amblies will be shuffled to new locations, primarily i

on or near the core periphery. We remaining 8 will surround the center assembly. We 64 fresh batch 10 assemblies will be loaded in a symmetric checkerboard pattern throughout the core. Figure 3-2 is an eighth-core map i

showing the assembly burnup and enrichment distribution at the beginning of cycle.

3-1 Babcock &Wilcon a McDermott company

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Reactivity is controlled by 60 full-length Ag-In-Od control rods, 64 burnable poison rod amamblies (BPRAs), and soluble boron shim. In addition to the full-length control rods, eight axial power shaping rods (APSRs) are provided for additional curikul of the axial power distribution. 'Ihe cycle 8 locations of the 68 coukul rods and the group designations are indicated i

in Figure 3-3. 'Ihe core locations and group designations of the total pattern (69 control rods) for cycle 8 are the same as those of the reference cycle 3 (69 c.vinhul rods) except for the center location. 'Ihe cycle 8 locations and enridmed.s of the BPRAs are shown in Figure 3-4.

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Figure 3-1. ~ Core Loading Diagram for ANO-1 C.ycle 8 X

l 9 9 88 9 9 A LOS NO3 G08 N13 L11 88 9 9 10 9 . 10 9 9 8B 8 LQ3 F EOS F L13 M12 P06 P10 M04 10 88 10 88 10 9 10 9 C w 9 10 9 F A06 F A10 F K10 F R08 HIS F K06 59 10 9 10 88 10 9 10 88 88 10 9- 10 0 B05 F 009 F B11 F 007 F LO2 L14 F K04 F 10 28 10 88 10 9 10 9 9 E 9 9 10 9 F H03 F C05 F C11 F C08 F F07 005 C11 F09 88 10 8B 10 88 10 9 9 9 9 10 88 10 89 10 g F K01 F E14 F CC6 [06 E10 C10 F E02 F P09 F R07 9 9 9 10 88 10 SS 10 9 g 9 10 88 10 88 10 F P09 B09 K02 F E13 F FIS F C04 C12 F F01 F E03 9 10 9 88 -Y 88 9 10 9 10 83 ' 9 70 9 88 10 w_g i K05 K14 G01 F K12 F H11 909 HOT N05 F G04 F KIS G02 i

10 88 IJ 9 9 9 10 8B 10 88 10 9 9 10 88 K F M03 F G14 P07 607 F M13 F L15 F 304 C12 F LOL i

88 38 10 88 10 1 9 9 9 10 88 10 98 10 10 L M10 810 F N02 F GIS F A09 F A07 F M14 F 306 MC6 1 10 39 10 93 10 9 10 9 9 9 9 10 M N11 LO9 F lt08 F 435 F d!! F H13 F LO7 NCS 9 10 S3 9 10 SB 10 9 10 83 N 88 10 l' N09 F PCS F 'I07 F P11 F G12 F F02 F14 F 10 9 10 83 10 SB 10 9 10 9 9

0 A08 F G06 F R06 F R'.0 F G10 F H01 88 9 9 10 9 10 9 9 SS p

810 E04 F03 F "08 F F13 E12 806 9 9 88 9 9 R

F05 003 K08 013 Fil I

Z 9 10 11 12 13 14 IS 4 5 6 7 ~8 t 2 3 BATCH Note: F Denotes Fresh Fuel Previous Core location l

3-3 matecock&WWIlcom a AacDermott company

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Figure 3-2. Enrichment and Burnup Distribution, ANO-1 Cycle 8 off 425 EFPD Cycle 7 8 9 10 11 12 13 14 15 2.95. 3.30 3.21 3.35 3.30 3.35 3.30 3.21 H

45792 15232 23972 0 16877 0 16806 31130 3.30 3.35 3.21 3.35 3.21 3.35 3.30 X

15233 0 26611 0 21394 0 12866 3.21 3.35 3.21 3.35 3.30 3.30 L ,

23982 0 22404 0 16528 16833 3.30 3.35 3.30 3.30 M

17455 0 17330 16865 3.30 3.35 3.21 N

16905 0 23087 3.30 0

10891 P

R x.xx Initial Enrichment, wt % U-235 xxxxx BOC Burnup mwd /mtU 3-4 Babcocira % g

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Figura 3-3. Control Rod Locations and Group Designations for ANO-1 Cycle 8 X

Fuel Transfer Canal )

A B 4 7 4 C 2 6 6 2 D 7 3 5 8 7 E 2 5 1 1 5 2 F 4 8 3 7 3 8 4 G 6 1 3 3 l1 6 H W- 7 5 7 l l7l l5 l !7 K 6 1 3 3l l1 l 6l l L~

l3 4 8 3 7 8 l 4 11 , 2 5 1 1l l5{ 2 N l 7 8 5 l8 l 7 0 l 2 6 6l 2{

P l l l 4 7 l4l R I I  !,

. Z 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 X Group Number Grouc No. of Rods Function 1 8 Safety 2 8 Sa fe ty 3 3 Sa fety 4 8 Safety 5 8 Control 6 8 Control 7 12 Control 8 8 APSRs 3-5 Babcock &WHcom a AACDermott company L_

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4 t' Figure 3-4. LBP Enrichment and Distribution, ANO-1 Cycle 8 .

8 9 10 11 12 13 14 15 H 1.35 1.35 K 1.21 1.18 0.20 l

L 1.21 -

1.24 0.20 I

M 1.35 1.24 1.24 4

N 1.18 1.24 0.20 ,

0 1.35 0.20 0.20 l

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x.xx LBP Concentration, wt % B4 C in A1230 l

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4. EUEL SYSITM DE: SIGN 4.1. Ebel A M lv Mechanical Desian 2n types of fuel aMlies and pertinent fuel paranaters for ANO-1 cycle 8 are M = = ==4 below and listed in Table 4-1.

All fuel aummhliss are identical in um ard are mechanically interchangeable.

Retainer ;

assemblies will be used on two fuel ama==hlies that contain the regenerative neutron sources, and on sixty-four fuel assemblies that contain Brms.

Justificaticm of the design and use of retainer assemblies is given in references 4 and 5. Se batch 7D fuel assembly has the highest burnup in the oors.

Se sixty four batch 10 fuel assemblies ircewate the design features of anti-straddle lower and fittings and annealed guik tubes. We anti-straddle end-fitting prevents mis-positioning fuel n e lles during refuelity operations and the annealed guide tubes reduce incere irradiation fuel assembly growth which allows for higher burnup capability.

We MK-BEB fuel assembly differs from the MK-B design in that it permits easy removal of a limited number of fuel rods. .In additten, windows are cut into the upper grid skirt to permit insy cbservatien of fuel red growth.

4.2. Ebel Rod Desian te MK-BEB fuel rod design differs frcm the MK-B fael rod in several areas.

We MK-BEB fuel rod cladding is thicker with a lcwer preprtsstne to achieve better high burnup perforrance. Anrular pellets, which are exp' x ted to improve high burnup perforrance are centained in scre of the MK-BEB rods.

We pin pre-pressure in the batch 10 fuel rods b.s been reduced 50 psi to improve fuel perforranco. Se reduced pre-pressure has been censidered in all nochanical and therm 1 analysis. %e results cf the rnchanical evaluations of the fuel rods are diccassed belcw.

4-1 Babcock &WHcom J MtDHmott LOWpeny

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4.2.1. M=MNrth11 -

The batch 7D fuel aM1y is more limiting than batches 8, 9, arti 10 hann an of its previous incore exposure time. %e batdt 7D power history -

Was aralyzed* to ensure that creep ovalization will not affect the fuel' '

perforranca during cycle .8. De creep collapse analysis is based on t

referunos 6.

he creep collapen analysis predicts a collapaa tima greater than 45000 effective full poker hours. (EFPH), which is longer than the =_v4== ,

expected residenot, time of 41000 EITH (Table 4-1).

j 4.2.2. M m M bus st m

%e ANI>1 cycle 8 stress persanters are enveloped by ocnearvative fuel red stress analysis. %e samm authod was used for analysis of cycle 8 that had I

! been used cat the previous cycle.

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4.2.3. clam 4'n Strain f

'Ihm fuel design criteria specify a 1.0% limit on cladding plastic tensile

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cirnanfarential atrain. Da gallet is designed to ensure that plastic '

j cladding strain is less than 1.0% at design local pellet burray and heat l generation rata.

ha design values are higher than the worst-case values the ant >-1 cycle 8 fuel is expected to sea. %e strain analysis is also based on the upper toleranos values for the fuel pellet diameter and density and the lower tolerance for the cladding inside diameter.

i j 4.3. 'Ihermal Desian i

4 All fuel in the cycle 8 core is thermally similar. %e design of the batch l.

70 lead tett assembly is such that the thermal perfomance of this fuel is equivalent tc or slightly better than the standarti Mark-B design used in the i

renaindar of the cere. All themal design analyses for cycle 8 fuel used the TA002 code, as described in Reference 7, for fuel temperature and fuel ,

{ red internal pressure prediction.

2n results of the thermal design evaluation of the cycle 8 core are

[ sumarized in Table 4-2. Cycle 8 core protection limits were based on a

{+ linear heat (UIR) to centerline fuel melt of 20.5 kW/ft as detamined by the 1 TA002 code. %e DiR to melt of the LTA fuel is greater than 20.5 kW/ft.

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%e maximum fuel a==*ily burr:up at DOC 8 is predicted to be .less than 4 l i

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41,600 MtVatif for the Mark-B fuel and less than 56,300 MtymtU for the IIDL fuel.

'Ibe fuel rod internal pressures have been evaluated with TACD2 for the highest burnup fuel rods ani are predicted to be less than the natinal reactor coolant pressure of 2200 psia.

4.4. Material Desian

'Iha- chemical ocupatibility of all possible fuelelaMig-coolant-assenably interactions for Batch 10 fuel assemblies is identical to that of the previous fuel batches.

4.5. Comratim %rience Rahmrt & Wilcox operating eqerience with the Mark B 15x15 fuel ammambly has verified the MMwy of its design. As of April 30, 1986, the following experience has been arv'= lated for eight B&W 177 fuel ahly plants using the Mark B fuel amihly; C11rrent nmailative not Reactor Max FA burnuo. (a) mwd /:rtU electric cycle Inocre Discharged outnut.(b) g g >

Oconee 1 10 33,710 50,598 62,028,968 Ooonee 2 8 38,100 37,326 55,785,115 Oconee 3 9 37,714 39,229 55,385,714 ,

'1hree Mile Island 5 28,440 32,400 25,105,483 Arkansas Nuclear '

One, Unit 1 -7 41,960 36,820 48,299,124 Rancho Seco 7 26,100 38,268 39,078,111 Crystal River 3 6 24,970 31,420 35,863,252 Davis-Besse 5 31,020 32,790 25,233,177 (a)As of April 30, 1986.

(D)As of January 31, 1986. t I .

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l Table 4-1. Fuel _Derian _33mg*cgg.ftdfzigt@j;?ili )

Batch 70 M MI ,pg g Q bip;Jq Ebel ascm bly '

type lE BEL MM ($ E4 W E4 i NL:mber of  ;

assenblies 1 44 f>8 64  ;

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, Ebel rod CD (nce) in 0.430 0.430 th.AA1 0.5 M .

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Fuel rod ID

, (nam) in~ 0.371 0.377 0.70 0.J7'i j Undensified active fuel length in 138.25 141.8 141 9 14 1. s Ebel pellet

, OD (mean) in 0.3635 0.3686 0.?C86 0.X00 l Ebel pellet initial density  !

(rm) % 'ID 95 95 95 SS Initial fuel enrichment ,

wt. % 23 % 2.95 3.21 3.30 3.M Average burnup, BCC, NWd/mtU 45800 24200 1590A 0

., Cladding '

collapse time, EFEH >45000 >35000 >35000 >35000 Estiratal residence
exposure time, i EFTH EOC 41000 30000 20000 10000 1

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s e Table 4-2. R:el ihemal Analysis Parameters t

Batch 7D .SgLtch 8B Batch 9 fatch_10 lio, of nsserblies 1(a) 44 68 64 Initial density,

% TD 95.0 95.0 95.0 95.0 k 'tsitial pellet 3

00, in 0.3635 0.3686 0.3686 0.3686 Tritial stack height in 138.25 141.80 141.80 141.80 E: ri.:: mat, 4 U-235 2.95 3.21 3.30 3.35 i'ottinal linear heat t

rate at 2558 Mit, M,t'ft(b) 5.89 5.74 5.74 5.74 TbCQ2-brexi Mediqigm T

J Averagc fuel tcrporature at nominal Um, F <1400 1400 1400 1400 MirJrm UR to mit, kW/ft 21.1 20.5 20.5 20.5 Core average UE = 5.74 M1/ft (a)12A analysis res'lts are reported in Reference 2.

(b)Eeed on a nominal stack height.

1 I 4-5 pj Babcock &Wilcox 1 g a uccermore company

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5. NJCIEAR DESIGT 5.1. Physics Charry;;paristics Table 5-1 lists tha core physics parameters of design cycles 7 and 8. 'Ihe values for cycle 7 wem generated using PDQ078 and the values for cycle 8 wem calculated with the NOCOIE cock.9 Figure 5-1 illustrates a representative relative power distributien for the beginning of cycle 8 at full power with equilibrium xenon and r-imi red positions.

Differences in feed enridu uit, EPRA loading, and shuffle pattern ruke it difficult to ramm the physics parameters of cycles 7 and 8. Calculated ajected rod worths and their adimrence to criteria are considered at all cimos in life and at all power levels in the develognent of the rod position limits presented in section 8. 'Ihe mv4== stuck rod worth for cycle 8 is less than that for the design cycle 7 at BOC, but greater at APSR pull and EOC. All safety criteria associated with those worths are mt. The adequacy of the shutdown rargin with e.:ycle 8 stuck rod worths is dini.-(=Lsated in Table 5-2. h following conse:.vatistos were applied for the shutdown emlm21stions:

1. Poison material depletion allowance.
2. 10% urca-Uninty on net red worth.
3. Flux redistribution.

Flux redistribution was accounted for since the shutdown analysis was calculated using a two-dimensional model. 'the reference fuel cycle shutdown margin is presented in the MD-1 cycle 7 Iuload report.3 5.2. Analvtical Inout h cycle 8 incore mcasurement calculation constants to be used for omputing core power distributions were prepared in the same ranner as these for the refererce cycle.

5-1 Babcock &Wilcos a McDermott company

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5.3. Channes in Ntaclear Desim core design changes for cycle 8 include the reaval of the center GA ard a change in the IRP length. '1he center GA will be replaced with a stand pipe and blind flange. Removal of the center GA will have a negligible effect on the nuclear parameters for cycle 8. 'Ihe IBP used in cycle 8 has a 4.5-inch longer poison stack than that used in cycle 7, i.e.,121.5 versus 117 inches of Al 23 0 -B4 C. 'Ihe top 4.5 inches of the poison stack are replaced by a Zircaloy tubular spacer. 'Ihis IEP design asymetrically positions the burnable poisen stack relative to the fuel column and alters the core axial '

power shape to create incraaamri " effective maneuvering recan" at the beginnity of the cycle.

As stated in section 5.1, the N00 DIE code was used to calculate the physics parametazs for cycle 8. 'Ibe N00 DIE =4=14ng of the tm t w- f--dzed '

fuel a===hly is the same as that used in PDQ07. However, the analytical i egression N00 DIE uses for the spatial flux solution provides more accurate '

results than the finite difference expression used in PDQ07 when there are few flux solution points per m e ly. Reference 9 illustrates the i calculational accuracy attainable with NOODIE in ccrparison to measured results for various physics paramters. E0Q07 results are ecmpared to measured data in references 10 and 11. 'Ihese careparisons show NOODIE to be as ammata as PDQ07.

+

As in cycle 7, the APSRs will be withdrawn near the end of cycle 8 (380 l EFFD). 'Ihe calculated stability index at 384 EFPD without APSPs is -0.022 ,

h-1 which d-edwates the axial stability of the core. 'Ihe calculational methods used to obtain the iqxartant nuclear design paramters for this cycle were the same as those used for the reference cycle. 'Ite operatirg '

limits (Technical Specifications changes) for the reload cycle are given in section 8.

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5-2 Babcock & WI8com a Mcoermott company

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1^ .1 Table 5-1. Ittvsics Parameter 1s for AND-1, (Neles 7 and 8(a) cvele 7(b) cvele 8(C)

Cycle length, EFPD 420 420 Cycle burrmp, IMS/ntU 13,158 13,147 Avg. core burnup, EOC, MWe/stU 2.4,238 25,522 Initial core loadirg, mtU 82.0 82.0 Critical bacon - BOC, ppm (No Xe)

HZP,(d) group 8 ins 1578 1644 HFP, group 8 ins 1346 1409 Critical boren - EDC, ppm HZP, group 8 out, no Xe 696 651 HFP, group 8 out, eq Xe 83 18 Control rod worths - HFP, BOC, % k/k Group 6 1.20 1.14 Group 7 1.E5 1.49 Group 8 0.39 0.39 Control red worths - HFP, IDC, % .k/k Group 7 1.53 1.52 Max ejected rod worth - HZP, % k/k(*)

BOC (N-12), group 8 ins 0.69 0.55 380 EFPD (N-12), group 8 ins 0.50 0.60 EOC (N-12), grcup 8 out 0.52 0.59 Max stuck rod worth - ICP, % k/k BOC (N-12), group 8 ins 1.71- 1.58 380 EFPD (N-12), group 8 ins . 1.73 1.86 EOC (H-12), group 8 out 1.29 1.63 Power deficit, HFP to IEP, % X/k BOC 1.60 1.56 BOC 2.35 2.34

~3 Babcock & WHees a AkDermott company

h cvele 7(b) gje3 g(c)

Doppler coeff - HFP,10-5 (ak/k/ F)

BOC (no Xe) -1.53 -1.54

. EOC (eg Xe) -1.80 -1.84-Moderator coeff - HFP,10-4 (A/V F)

BOC, (no Xe, crit ppra, group 8 ins) -0.69 -0.51 EOC, -(eg Xe, O ppu, group 8 out) -2.79 -2.78 Boron worth - HFP, m/% ak/k BOC 129 129 EOC 109 111 Xenon Worth - HFP, % (Jc/k BOC (4 EFPD) 2.55 2.55 EOC (equilibrium) 2.68 2.72 Effective delayed neutron fraction - HFP BOC 0.00G3 0.0062 EOC 0.0052 0.0052 (a) Cycle 8 data are for the conditions stated in this report. 'nie Cycle 7 core conditions are identified in Reference 3.

(D)maari on 400 ETTO at 2568 NWt, Cycle 6.

(C) Based on 425 EFPD at 2568 MWt, Cycle 7.

(d)HZP denotes hot zero power (532F T avg), HFP denotes hot full power (579 T avg)*

(e) Ejected rod worth for groups 5 through 7 3rm%, group 8 as stated.

5-4 gabcock &Wilcox a McDermott company

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NS)d a NaDNM D bhdhMbDd N bf 10 8 +

BOC 380 Ef?D 420 YHD 3_.ds/js  % &/k .  % ak/l i_

Ayaileble Prx1_ Wort))

Total,rtd terth, li2P 8.85 9.38 9.15 Wdeth redtetion due to poisort raterial burnup -0.10 -0.10 -0.1t>

Maxinna sttxi rod, IIZP -lt58 1,66 -L13 ,

Net worth 7.17 7.42 7.42 Isss 10% uncertainty -0.72 -A._7.3 -0.74 Total available worth 6.45 6.68 6.68 .

Reauired Rod Worth PcMur deficit, liFP to HZP 1.57 2.30 2.34 ulowable inserted rod ,

worth .50 .60 .65 Flux redistribution ,_d4 L29 L2Q Total required worth 2.91 4.10 4.19 Shutdown avgin (total available worth minus total reqh worth) 3.54 2.58 2.49 1[qt;g: 'Ihe required shutdcwn ergin is 1.00% &/k.

5-5 Babcock &Wilcom J McDermott company

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

H 0.70 1.17 1.08 1.27 1.28 1.29 1.02 0.42 K 1.16 1.24 1.29 1.05 1.28 1.12 1.15 0.54 L 1.08 1.29' 1.10 1.27 0.99 1.31 0.88 0.39 M 1.27 1.04 1.26 1.26 1.28 1.06 '0.62 N 1.27 1.28 0.99 1.28 1.18 1.01 0.36 0 1.28 1.12 1.31 1.06 1.02 0.59 P 1.02 1.15 0.88 0.62 0.37 l

l R 0.42 0.54 0.39 l

l x Inserted Rod group No.

l Relative Power Density

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5-6 Babcock &MW

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6. THERMAIrHYDRAULIC DESIGi Tha fresh batch 10 fuel is hydraulically and geometrically similar to the previously irradiated batches 8B and 9 fuel. h modified Mark B lower end fittire (IH) was found to have an insignificant impact on thermal-hydraulic results. The batch 7D In has been analy::ed to ensure that it is never the limiting assembly during cycle 8 operation. h results of the thermal-hydraulic analysis for the LTA are provided in reference 2.

The thermal-hydraulic design evaluation supporting cycle 8 operation is hacarl on methods and models described in references 12, 13, 14, and 15. The cycle 8 thermal-hydraulic design is identical to cycle 7. The thermal-hydraulic design conditions for cycles 7 and 8 are summarized in Table 6-1.

The reactor protection system (RPS) setpoints for the ENB-basal variable low pressure trip will remain the same for cycle 8. The 1.08 flux / flow setpoint remains applicable for cycle 8.

A red bow topical report (reference 16), which addresses the mechanisirs and resulting conditions of rod bow, has been submitted to and approved by the NRC. The topical report concludan that rod bow penalty is insignificant and is offset by the reduction in power production capabilitf- of the fuel accamblies with irradiation. Therefore, no departure from nucleate boiling ratio (ENER) reduction due to rod bow need be considered for cycle 8.

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Table 6-1. Maximum Desian Conditions. Cycles 7 and 8 Cycle 7 Cycle 8 Design power level, Mht 2568 2568 System pressure, psia 2200 2200 Reactor coolant flow, % design 106.5 106.5 Vessel inlet / outlet coolant temp at 100% power, F 555.6/602.4 555.6/602.4 INBR modeling Crossflow Crossflow Reference design radial-local power peaking factor 1.71 1.71 Reference design axial flux shape 1.65 cosine 1.65 cosine Hot channel factors Enthalpy rise 1.011 1.011 Heat flux 1.014 1.014 Flow area 0.98 0.98 Active fuel length, in. 141.8 141.8 Avg. heat flux at 100% power, 103 Btu /h-ft2 174 174 t Max. heat flux at 100% power, 103 Btu /h-ft2 492 492 OIF correlation B&W-2 B&W-2 Minimum [EBR at 112% pcrw'er 2.08 2.08-at 100% power 2.43 2.43 6-2 Babcock &Wilcox a McDermott company

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7. AccinFMr AND TRANSIENT ANALYSIS

~

3 7.1. General Safety Analysis ,

Each ESAR accident analysis has been examined with respect to changes in

! cycle 8 parameters to determine the effect of the cycle 8 reload and to ensure that thermal performance during hypothetical transients is not degraded.

'Ihe effect of fuel densification on the FSAR accident results have been evaluated and are reported in reference 17. Since batch 10 reload fuel anaamblies contain fuel rods whose theoretical density is higher than those considered in the reference 17 report, the conclusions in that reference are I

still valid.

'Ihe radiological- dose consequences of the accidents presented in Chapter 14

~

of the FSAR were re-evaluated for this reload Kr.pvit. 'Ihe raa=m for the re-evaluation is that, even though the ESAR dose analyses used 'a conservative basis for the amount of plutonium fissioning in the core,

. improvements in fuel manair.-r.ut techniques have increased the amount of erwrgy prr*M by fissioning plutonium. . Since plutonium-239 has different fission yields than uranium-235, the mixture of fission product nuclides in the core changes slightly as the plutonium-239 to uranium-235 fission ratio changes, i.e., plutonium fissions produce more of some nuclides and less of other nuclides. Since the radiological doses associated with each accident are inpacted to a different extent by each nuclide and by various mitigating factors and plant design features, the radiological consequences of the FSAR accidents were recalculated using the specific parameters applicable to cycle 8. The bases used in the dose calculation are identical to those ,

presented in the ESAR except for the following three differences:

1. 'Ihe fission yields and half-lives used in the new calculations are based on more current data.
2. Updated (lowered) whole body gama dose conversion factors.

7-1 gggg a McDermott company 2

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3. Se steam generator tube rupture accident evaluation considers the increased amount of steam released to the environment via the main steam relief and atmospheric dtmp valves because of the slower depressurization due to the r*ad heat transfer rate caused by tripping of the reactor coolant ptmps upon actuation of the high

' pressure injection (a post'IMI-2 modification).

A comparison of the radiological doses presented in the FSAR with those calculated specifically for cycle 8 (Table 7-1) show that some doses are slightly higher and scue are slightly lower than the FSAR values. However, with the exception of the mvb= hypothetical accident (MHA) all doses are bounded by the values represented in the FSAR or are a small fraction of the 10 CFR 100 limits, i.e., below 30 Rem to the thyroid or 2.5 Ren to the whole body. For the MHA the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> thyroid dose at the exclusion area boundary (EAB) is 157.3 Rem (53% of the 10 CFR 100 limit) and the 30 day thyroid dose at the low population zone (LPZ) is 73.0 Rem (24% of the 10 CFR 100 limit) .

Bus, the radiological impact of accidents during cycle 8 is not significantly different than that described in Chapter 14 of the FSAR.

7.2. Accident Evaluation h e key parameters that have the greatest effect on determining the outccme 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 red worths.

Core thermal properties used in the ESAR accident analysis were design operating values based on calculational values plus uncertainties. Thermal parameters for fuel batches 7D, 8, 9 and 10 are given in Table 4-2. The cycle 8 therral-hydraulic maxinm design conditions are cc:: pared with the previous cycle 7 values in Table 6-1. Wese parameters are cc=non to all the accidents censidered in this report. We key kinetics parameters from the FSAR and cycle 8 are cc:: pared in Table 7-2.

A generic IDCA analysis for a B&W 177-FA, lowered-loop NSS has been performed using the Final Acceptance Criteria ECCS Evaluation Model (reported in BAW-10103) .18 This analysis is generic since the limiting values of key parameters for all plants in this category were used.

Furthermore, the cambination of average fuel te::peratures as a function of 7-2 Babcock & Wilcox a McDermott company

. I IIR and lifetime pin pressure data used in the BAW-10103 IOCA limits

analysis is conservative ocupared to those m_1<,ilated for this reload. _

'Ihus, the analysis and the IOCA limits s w ried in BAW-10103 and substantiated by reference 19 provide conservative results for the operation of the reload cycle. Table 7-3 shows the bounding values for allowable IOCA peak IHR3 for ANO-1 cycle 8 fuel. 'Ihese IHR limits include the effects of NUREG 0630.

It is concluded frcan the examination of cycle 8 core thermal and kinetics psgdes, with re_ to acceptable previous cycle values, that this core reload will not adversely affect the ANO-1 plaru's ability to operate safely during cycle 8. Considering the previously accepted design basis used in the ESAR and siW=nt cycles, the transient evaluation of cycle 8 is considered to be bounded by previously accepted analyses. 'Ihe initial conditions for the transients in cycle 8 are bounded by the FSAR, the fuel densification report, ancyor sW=nt cycle analyses.

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7-3 Babcock &Wilcox a McDermott company

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, e Table 7-1. ComDarison of FSAR and Cycle 8 Accident Doses ESAR doses, Cycle 8 doses, Rem Rem Riel Handlim Accident hyroid dose at EAB (2 h) 0.92 1.15 Whole body dose at EAB (2 h) 0.54 0.21 Steam Line Break h yroid dose at EAB (2 h) 1.6 1.71 Nhole body dose at EAB (2 h) -

0.008 Steam Generator Tube Failure hyroid dose at EAB (2 h) 0.0087 6.14 Whole body dose at EAB (2 h) 0.16 0.52 Waste Gas Tan): Ruoture B yroid dose at EAB (2 h) 0.22 0.054 Whole body dose at EAB (2 h) -

1.53 Control Rod Eiection Accident hyroid dose at EAB (2 h) 11.4 12.2 Nhole body dose at EAB (2 h)' O.014 0.008 hyroid dose at LPZ (30 d) 8.3 9.09 Nhole body dose at LPZ (30 d) 0.0099 0.005 M

l h yroid dose at EAB (2 h) 3.6 4.02

Nhole body dose at EAB (2 h) 0.057 0.026 l

%yroid dose at LPZ (30 d) 1.66 2.05 Nhole body dose at LPZ (30 d) 0.043 0.018 Maximum Hvoethetical Accident .

'Ihyroid dose at EAB (2 h) 153 157.3 Whole body dose at EAB (2 h) 10 4.80 hyroid dose at LPZ (30 d) 64.1 73.0 Whole body dose at LPZ (30 d) 3.4 1.56 7-4 babcock & WilCOE a Mcoermott ccmpany

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1 Table 7-2. Ocznparison of Fey Parameters l for Accident Analysis l

FSAR and densification MD-1 Parameter levutt value cycle 8 Doppler coeff (BOC),10-5 3yyyjoF -1.17 -1.54 Doppler coeff (EOC),10-5 Ak/k/ F -1.30 -1.84 )

Moderator coeff (BOC),10-4 Ak/k/0F 0.0(a) -0.51 Moderator coeff (EOC),10-4 A k/k/0F -4.0 (D) -2.78 All-rod group worth (HZP), % Ak/k 12.9 8.85 Initial boron concentration, ppn 1150 1409 Doron reactivity worth (HFP), 100 129 pprV% Ak/k Max. ejected Itxi worth (HFP), % Ak/k O.65 0.34 Dug rod worth (HFP), % Ak/k O.65 <0.20 (a)+0.5 x 10-4 Ak/k/ F was used for the moderator dilution analysis.

(b)-3.0 x 10-4 Ak/k/0F was used for the steam line failure analysis.

Table 7-3. Bounding Values for Allowable IDCA Peak Linear Heat Rates Allowable Allowable Allowable Core peak IHR, peak IHR, peak IHR

' elevation, 0-1000 MNd/mtU, 1000-2600 MNd/mtU, after 2600 MNd/mtU, ft W/ft W/ft kW/ft i

2 13.5 15.0 15.5 4 16.1 '16.6 16.6 6 16.5 18.0 18.0 8 17.0 17.0 17.0 i

l 10 16.0 16.0 16.0 i

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7-5 Babcock & Wilcox a McDermott company

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8. PROPOSED MODIFICATIONS 'IO TECliNICAL SPECIFICATIONS 2e Technical- Specifications have been revised for cycle 8 operation to j account for changes in power peaking and control rod worths. As in cycle 7,
a very low leakage fuel cycle and crossflow analysis were inplemented in the fuel cycle design. Be IOCA linear heat rate limits used to develop the Technical Specification Limiting Conditions for Operation include the innae+

of NUREG-0630 claMing swell and rupture model. In addition, an analysis was cow +M to verify retoval of the power level cutoff hold requi1.eumius -

l of Technical Specification sections 3.5.2.4 and 3.5.2.5.

A cycle 8 specific analysis was cow +M to generate Technical Specification Limiting Conditions for Operation (rod index, APSR position, axial imbalance, quadrant . tilt). We analysis generated measurement-j irdri:perdait. 100 limits which were then error-adjusted to give alarm setpoints for power operation. Se Technical Specification IfD~ figures are presented as alarm setpoint figures. S e fuel cycle design allows for Axial Power Shaping Rod (APSR) withdrawal at 380 i 10 EFPD, and is reflected in the Iro figures. Figure 3.5.2-4 is also provided, which illustrates the burnup ,3ep-rdad. allowable IOCA linear heat rate limits used in the analysis. Se analysis also verified the 3.1% quadrant tilt setpoints referenad in Technical Specificatica 3.5.2.4.

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. Se following pages contain the revisions to previous Technical Specifications.

8-1 Babcock & Wilcox

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6. If a control rod in the regulating or axial power shaping groups is declared inoperable per Specification 4.7.1.2, operation above 60% of the thermal power allowable for the

' reactor coolant punp otrbination may continue provided the rods in the group are positioned such that the rod that was declared inoperable is contained within allowable group average position limits of Specification 4.7.1.2 and the withdrawal limits of Specification 3.5.2.5.3.

3.5.2.3. 'Ihe worth of single inserted control rods during criticality are limited by the restrictions of Specification 3.1.3.5 and the Control Rod Position Limits defined in Specification 3.5.2.5.

3.5.2.4. Quadrant tilt:

1. Dccept for physics tests, if quadrant tilt ex e 3.1%,

reduce power so as not to exceed the allowable power level for the existing reactor coolant purp cmbination less at least 2% for each 1% tilt in excess of 3.1%.

2. Within a period of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, the quadrant power tilt shall be r=+M to less than 3.1% except for physics tests, or the following adjust =ents in setpoints and limits shall be made:
a. 'Ihe protection systen maximum allcvable setpoints (Figure 2.3-2) shall be reduced 2% in power for each 1% tilt.
b. 'Ihe control rod group and APSR withdrawal limits shall be r=+M 2% in power for each 1% tilt in excess of 3.1%.
c. 'Ihe operational imbalance limits shall be reduced 2% i.n power for each 1% tilt in excess of 3.1%.

, 3. If quadrant tilt is in excess of 25%, except for physics tests or diagnostic testing, the reactor will 'be placed in the hot shutdown condition. Diagnostic testing during power operation with a quadrant power tilt is permitted provided

the therral power allowable for the reactor coolant purp cxrbination is restricted as stated in 3.5.2.4.1 above.

8-2 Babcock & WilcoE a McDermott company

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4. Quadrant tilt shall be monitored on a minimum frequency of once every two hours during power operation above 15% of rated power.

3.5.2.5. Control rod positions:

1. Technical Specification 3.1.3.5 (safety rod withdrawal) .does not prohibit the exercising of 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 red' group overlap shall be. 20% 5 between two sequential groups, except for physics tests.
3. Except for physics tests or exercising cvu'uvl rods, (a) the control rod withdrawal limits are specified on Figures 3.5.2-1, 3.5.2-2A and 3.5.2-2B for 4, 3 and 2 pump operation respectively; and (b) the axial power shaping control rod withdrawal limits are specified on Figures 3.5.2-4A and 3.5.2-4B. If any of these control rod position limits are WM , corrective measures shall be taken immMiately to achieve an acceptable control rod position. Acceptable wu'uvl rod positions shall be attained within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

3.5.2.6.. - Reactor Power Imbalance shall be monitored on a frequency not to a=M 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> during power operation above .40% rated power.

Except for physics tests, imbalance shall be maintained within the envelope defined by Figure 3.5.2-3. If the imbalance in not l

within the envelope defined by Figure 3.5.2-3, corrective measures shall be taken to achieve an acceptable imbalance. If

an acceptable imbalance is not achieved within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, reactor power shall be rMW until imbalance limits are met.

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

Bases

'Ihe power-imbalance envelope defined in Figure 3.5.2-3 is based on (1) IDCA analyses which have defined the maximum linear heat rate (see Figure 3.5.2-4), such that the maximum claMing temperature will not exceed the Final Acceptance Criteria and (2) the Protective System Paximum Allowable 8-3 Babcock &WHcom a McDermott company

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Setpoints (Figure 2.3-2) . Corrective measures will be taken immediately should the indicated quadrant tilt, rod position, or imbalance be outside their specified boundaries. Operation in a situation that would cause the Final Acceptance Criteria to be approached should a IOCA occur is highly iny1.duable because all of the power distribution parameters (quadrant tilt, rod position, and imbalance) must be at their limits while 8-4 Babcock & Wilcox a McDermott company

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9 4.7. REACIOR CatfrROL ROD SYSTD( TESTS 4.7.1. Control Rod Drive System Rinctional Tests Acolicability Applies to the surveillance of the wahul rod system.

Obiective To assure operability of the w ikvl rod system.

Soecification 4.7.1.1. 'Ihe wukul. red trip insertion time shall be maared for each wikul rod at either full flw or no f1w conditions follwing each refueling outage prior to return to pwer. 'Ihe maximum control rod trip insertion time for an operable wikvl rod drive mechanism, except for the Axial Power Shaping Rods (APSRs), from the fully withdrawn position to 3/4 insertion (104 inches travel) shall not exceed 1.66 seconds at reah coolant full f1w conditions or 1.40 seconds for no flw cr4x11tions. For the APSRs it shall be dermr= Lated' that loss of power will not cause red movement. If the trip insertion time above is not met, the red shall be declared inoperable.

4.7.1.2. If a control rod is misaligned with its group average by more than an indicated nine (9) inches, the rod shall be delaved inoperable and the limits of Specification 3.5.2.2 shall apply.

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'Ihe rod with the greatest misaliyamuit shall be evaluated first.

'Ihe position of a rod declared inoperable due to misalignment shall not be included in computing the average position of the group for. determining the operability of rods with lesser mistlignments.

4.7.1.3. If a control rod cannot be exercised, or if it cannot be located with absolute or relative position indications or in or out limit lights, the rod shall be declared to be inoperable.

8-5 Babcock &Wilcox a McDermott company

a Bases

'Ihe control red trip insertion time is the total elapsed time from power interruption at the control rod drive breakers until the control rod has cmpleted 104 inches of travel from tha fully withdrawn position. 'Ihe specified trip time is based upon the safety analysis in FSAR, Section 14.

Each C.u hul rod drive wchanism shall be exercised by a movement of approximately two (2) inches of travel every two (2) weeks. 'Ihis requirement shall apply to either a partial or fully withdrawn control rod at reactor operating conditions. Exercising the drive mechanism in this manner provides assurance of reliability of the mechanisms.

A rod is considered inoperable if it cannot be exercised, if the trip insertion time is greater than the specified allowable time, or if the rod deviates from its group average pccithn by more than nine (9) inches.

Conditions for operation with an inoperable rod are specified in Technical Specification 3.5.2.

REFERENCES (1) FSAR, Section 14 l

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84 Babcock & Wilcox a Mcoermott company

Figure 8-1. Boric Acid Addition Tank Volume and Concentration Vs RCS Average Temperature -- ANO-1 Cycle 8 8000 0_PERATION ABOVE AND TO THE LEFT OF THE CURVES IS ACCEPTABLE 7000 -

8700 ppm O ppm 6000 -

E 10000 ppm c7 5 5000 -

j 12000 ppn Iro

[ 4000. -

2 d

E 3000 -

E 2i-e 2000 -

1000 -

0 I I I 200 300 400 500 600 RCS Average Temperature, F Temp., F Required Volume, gal 8700 ppm 9500 ppm 10000 ppm 12000 ppm 579 7308 6657 6306 5200 532 6126 5580 5289 4355 500 5273 4802 4548 3749 400 2790 2543 2409 1984 300 1234 1129 1070 877 200 0 0 0 0 8-7 Babcock &WHas a McDermott company

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Figure 8-2. Rod Position Setpoints for 4-Pump Operation From 0 to 25+10/-0 EFPD -- ANO-1 Cycle 8 110 (300,102) 100 -

(19 'I ) (283.6,102) >

gg _.' (277.5,90)

SHUTDOWN MARGIN LIMIT 80 (262.5,78) g 70 OPERATION OPERATION IN RESTRICTED

$ 60 - THIS REGION IS NOT /

0 ALLOWED

[ 50 (130.5,48) (176.5,48) 40 -

E 30 -

PERMISSIBLE

- OPERATING 20 REGION ig _

(6/.5,13)

(0,6.9) 0 t i I e I I I I I I e i l i 0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 0 20 40 60 80 100 l i I I I I Group 7 0 20 40 60 80 100 t i I t t I Group 6 0 20 40 60 80 100 I I I I I I Group 5 Rod Index, % WD l

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Figure 8-3. Rod Position Setpoints-for 4-Pump Operation From 25+10/-0 to 200+ EFPD -- ANO-l' Cycle'8 110 (198,102) (266.5,102)(:300,102)  ::

1gg _

90 -

(266.5,90) p SHUTDOWN MARGIN LIMIT 80 -

(257.5,78)

~ OPERATION y 70 -

RESTRICTED x

m 0PERATION IN 60 THIS REGION IS NOT ALLOWED 50 -

(130.5,48) (176.5,48)

$ 40 -

5

' PERMISSIBLE

- OPERATING 30 REGION 20 -

1g _

(67.5,13)

(0.6.9) 0 t ' ' ' ' ' ' ' ' ' ' ' ' '

0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 0 20 40 60 80 100 i t i I I t Group 7 0 20 40 60 80 100 t I i I t i Group 6 0 20 40 60 80 100 Rod Index, % WD Group 5 8-9 Babcock & Wilcox a McDermott company L

t i Figure 8-4. Rod Position Setpoints for 4-Pump Operation From 200110 to 380110 EFPD -- ANO-1 Cycle 8 110 (266.5,102)(300,102) 100 - (232.2,102) , ; i 90 - (266.5,90) ,

80 ( 53.5,78.1 PERATION ESTRICTED 70 E

co OPERATION IN THIS 60 -

REGION IS NOT e ALLOWED 50 (159.5,48) (176.5,48)

E -

g 40 O

PERMISSIBLE 30 -

OPERATING REGION 20 -

r

. 10 f <

l o (0,5.8)i i i i i i i i i i i i i l 0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 0 20 40 60 80 100 i i i i i i Group 7 0 20 40 60 80 100 l i i i i 1 0 20 40 60 80 100 Rod Index, % WD Group 5 l

l l

8-10 Babcock & Wilcox a McDermott company

Y Figure 8-5. Rod Position Setpoints for 4-Pump Operation After 380+10 EFPD -- ANO-1 Cycle 8 110 ---

(266.7,102)(300,102)

(234.3,102) 100 -

90 - qb(253.5,90)

SHUTD0'4N MARGIN 'LINIT 80 - OPERATICM J(253.5,78)

ESTRICTEL a 70 -

E OPERATION IN THIS g REGION IS NOT -

g 60 ALLOWED 50 -

(156.2,48) l

[ (176.5,48) 2 f 40 -

30 PERMISSIBLE OPERATINC 20 -

REGION 10 -

' (0,5.9) 0 e i i i i i i i ., i e , 1 i l 0 20 40 60 80 100 120 140 160 180 200 220 E40 263 280 300 0 20 40 60 80 100 i f i H 1 0 20 40 60 80 100 i a e i i _

1 Group 6 0 20 40

,60 ,80 q0 Rod Index, % **D Group 5 I

l l

8-11 Babcock & Wilcox l a McDermctt cempany

.J

t /

Figure 8-6. Rod Position Setpoints for 3-Pump Operation from 0 to 25+10/-0 EFPD -- ANO-1 Cycle 8 110 100 -

90 -

(300,77) 80 (198.8,77) (283.9,77) I a

~

E 70 (277.5,67)

SHUTDOWN MARGIN LIMIT

$ 60 -

(262.5,58)

% OPERATION 50 RESTRICTED

- OPERATION IN THIS

. REGION IS NOT ALLOWED u

E 40 - .

2 ( 0.5,36)

(176.5,35.5) 30 -

~

FERMISSIBLE OPERATING 10 - (0,5.1) 67.5,9,75) REGION 1

0 i i i t 1 i e i i i I i ' '

O 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 0 20 40 60 80 100 l t i t I i Group 7 0 20 40 60 80 100

, i i i i I Group 6 0 20 40 60 80 100

, , , , , , Rod Index, % WD Gruop 5 8-12 Babcock & Wilcox a McDermott company

i i c Figure 8-7. Rod Position Setpoints for 3-Pump Operation From 25+10/-0 to 200_+10 EFPD -- ANO-1 Cycle 8 110 i

i 100 -

~

90 -

80 - (266.5,77)(300,77)

(198.8,77)  ;  :

70 - i i SHUTDOWN MARGIN LIMIT (266.5,67) >

E 60 -

OPERATION IN THIS OPERATION D 50 -

REGION IS NOT RESTRICTED

>< ALLOWED ,

a 5 40 -

[ (130.5,36) (176.5,35.5) 1 30 -

20 -

PERMISSIBLE OPERATING 10 - (0,5.1) GION (67.5,9.75) i 0 i i i e i i i i i i 0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 0 20 40 60 80 100 I I f i e i Group 7 0 20 40 60 80 100 e i i e i i 0 20 40 60 80 100 1 I I i I i Group 5 Rod Index, " WD i

I 1

8-13 Babcock & Wilcox a McDermott company

q i (

Figure 8-8. Rod Position Setpoints for 3-Pump Operation From 200110 to 380+10 EFFD -- ANO-1 Cycle 8 110 l

100 -

90 -

gg _ (2c6.5,77)(300,77)

(233.1,77) 7  :-

~

$ 70 (266,5,67) '

m SiluTDOWN MA'4 GIN LIMIT E 40

- (253.5,58)

" OPERATION IN THIS PERATION REGION IS NOT . ESTRICTE'

, 50 ALL0biED 2

c.' -

40 (1 9,5,36) c (176.5,35.5) 30

~

PERMISSIBLE 20 OPEPATIi4G f REGION 10 -

(83.9,75) 0 i i e i i , i i , ,

i t i 0 20 40 60 80 100 120 140 160 189 200 220 240 260 280 300 0 70 40 60 80 100 i , ,1 v i Group 7 0 20 40 60 80 '00 i 1 'l i r _I Group 6 0 20 40 60 80 100 l

Red Index, % WD Group 5 l

8-F .1 Babrock & WNcox a McDermott company

,.s J Ffgure 8-9. Red Position Setpoints for 3-Pump Operation Af ter 38Q10 EFPD -- ANO-1 Cycle 8

~-

110 ~ ~ ~

100 -

9fI ,,,

80' _ (267.3,77) (300,77)

^

(235.8,77) 70' - <(253.5,67)

]r SutfrDOWN HARGIN LIMIT N

60 -

OPERATION 4(253.5,58)

OPERATION IN THIS RESTRICTED o -

REGION 15 110T 50 ALLOWED d .

  • U ~ l (155.2,36)

(176.5,35.5) 30 w  ;

PERMISSIBLE 20 ' -

< OPE. RATING REGION 13

- (0,4. 4 ) (86.5,9.75) t 0 3 ' ' ' 1 I ' I 8 8

_ _t i ' '

0 20 40 60 00 100 120 140 160 180 200 220 240 260 280 300 0 20 40 60 80 100 e f I f f I Group 7 0 20 40 60 80 100 e f i f f I Ge up 6 O 20 40 60 80 100 8 I - ' ' ' ' Pod Index, % WD Group 5 8-15 Babcock &Wilcox a McDermott company e

i <

Figure 8-10. Rod Position Setpoints for 2-Pump Operation From 0 to 25+10/-0 EFPD -- ANO-1 Cycle 8 110 100 -

90 -

~

80 h 70 e

" 60 -

(300,52)

(200.5,52) (284.3,52) >

50 _

w

- SHUTOOWN MARGIN LIMIT (277.5,44) b 40 - OPERATION IN THIS (262.5,38) d REGION IS NOT OPERATION ALLOWED RESTRICTED 30 -

(130.5.24)

(1 6.5,23) .

20 PERMISSIBLE 10 -(0,3.4) CPERATING REGION (67.5,6.5)

O< t I i 1 i i t t t t i i  !

~

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

0 20 40 60 80 100 I t I i i I Group 7 -

0 20 40 60 80 100 I I f I i 1 0 20 40 60 80 100

' I I ' ' ' Rod Index, % WD Group 5 l

l l

l 8-16 Babcock & Wilcox a McDermott company

Figure 8-11. Rod Position Setpoints for 2-Pump Operation From 25+10/-0 to 200+10 EFPD -- ANO-1 Cycle 8 -

110 100 -

~

90 80

~

h 70 60 (300,52) o (200.5,52) - (256.5,52)" .e

  • 50 OPERATION IN THIS SHUTDOWN 5-REGION IS NOT MARGIN (266.5,44) b E 40 -

ALLOWED LIMIT OPERATION (257.5,38) c' '

RESTRICTED 30 -

~

(130.5,24)

(176.5,23) 20 - '

PERMISSIBLE 10 - OPEPATING -

(0,3.4) - (67.5,6.5) 0 I t i t i i i i i i i t 9 i 0 20 40 60 80 100 120 140 160 180 200 2.20 240 250 280 300 0 20 40 60 80 100 .

1 P

  • t t t Group 7 -

0 .20 40 60 80 100 I I

  • 1 1 1

! Group 6 0 20 40 60 80 100 -

i , e . i Rod Index, 7, WD t

Group 5 l

l l 8-17 Babcock &Wilcox J MCDerme,t: Company

~

n

/ .

Figure 8-12. Rod Position Setpoints for 2-Pump Operation Frcm 200110 to 380110 EFPD -- ANO-1 Cycle 8 110 -

100 - .

1 90 80 -

y 70 -

co

$ 60 -

% (266.5,52)(300,52)

[, 50 _ (234.9,52) , ;  : ,

SHUTDOWN MARGIN s."

LIMIT (266.5,44) >

? 40 - OPERATION IN THIS '

y REGION IS NOT 0PERATION dESTRICTED (253.5,38)

ALLOWED 30 -

(159.5,24) ,

20

- (176.5,23)

PERMISSIBLE 10

- (0,2.9) OPERATING

- (83,6.5) REGION 0' r , , , , , , , , , , , , , ,

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

0 20 40 60 80 100 l I f f f f 0 20 40 60 80 100 t I i f f f Group 6 0 23 40 60 80 100 $

i t i  ! t i Rod Index, % MD ,

Group 5 I

I 8-18 Babcock & Wilcom a M:Dermott company

Figure 8-13. Rod Position Setpoints for 2-Pump Operation After 380+10 EFPD -- ANO-1 Cycle 8 110 l s i

100 -

90 -

80 ~

y 70 -

ce

$ 60 -

[ (268.2,52)(300,52)

(237.7,52) P 50 -

SHUTDOWN MARGIN LIMIT

40 OPERATION IN THIS OPERATION g REGION IS NOT RESTRICTE b(253.5,38)

ALLOWED '

30 -

I ' ' 4) (176.5,2h) 20 -

PERMISSIBLE OPERATING 10 - (0,2,9) REGION m

(80.5,6.5) 0 e i I I i e i i i i i i 0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 0 20 40 60 80 100 1 4. I t i I Group 7 0 20 40 60 80 100 t i I t i f i

Group 6 ,

0 20 j0' 60 80 100 Rod Index, % WD Group 5 i

is ,

\

8-19 Babcock &Wilcox l J McDermott company

t .-

Figure 8-14. Operational Power Imbalance Setpoints for Operation From 0 to 25+10/-0 EFPD -- ANO-1, Cycle 8 ff i

t'

% T

--110

\

(-10.1,'102 ) -

- -1001,(10.8,102)

(-20.2,92) ,,

(12.2,92) l

(-25.9,80) .

80 (16.8,80)

PERMISSIBLE OPERATING REGION

--70

--60 RESTRICTED RESTRICTED REGION o . 50 REGION E

m--40

$ s g--30 w

.--20

-' 5 3 .10

. o-a t t i I t  ! I I i

.h. 40 20 -10 0 10 20 30 40 50 Axial,' Power Imbalance, .%

i e

i t

i e

} 8-20 Babcock &Wilcox a McDermott company

x .-

Figure 8-15. Operational Power Imbalance Setpoints for Operation From 25+10/-0 to 200t10 EFPD -- ANO-1, Cycle 8 s

-- 110

(-18.0,102) e q

-- 100 9(14.6,102)

(-18.2,92)

- 90

(-24.7,80) .- 80 6(16.2,80)

PERMISSIBLE OPERATING REGION 70 60 RESTRICTED RESTRICTED REGION . REGION 50 E

g- - 40

.. . 30 a O

. , -- 20 i

E y- 10 e t I f I l i f f I 40 20 -10 0 10 20 30 40 50 Axial Power Imbalance, %

l l

l l

8-21 Babcock & Wilcox l a McDermott company l

i .'

Figure 8-16. Operational Power Imbalance Setpoints for Operation From 200110 to 380110 EFPD -- ANO-1, Cycle 8

-- 110

(-22.5,102) (13.0,102)

-- 100 1,

(-22.6,92) **

90

(-25.9,80) -

80 b(16.2,80)

PERMISSIBLE OPERATING ~" 70 REGION 60 RESTRICTED RESTRICTED REGION y. 50 REGION

=c

$~

N 40

%- 30 e

J- 20 2-- 10 m t i e i t t t t g 40 20 -10 0 10 20 30 40 50 Axial Power Imbalance, %

8-22 Babcock &Wilcox a McDermott company

a ,

i Figure 8-17. Operational Power Imbaiance Setpoints for Operation After 380_+10 EFPD -- ANO-1, Cycle 8

-- 110

(-23.6.102) (12.0,102) 100

(-26.3,92) l( .0,92)

- 90

(-26.7,80)0 -- 80 A(16.2,80)

PERMISSIBLE OPf. RATING -

70 REGION

- 60 RESTRICTED y RESTRICTED REGION M REGION

- 50 8

$-- 40 t

a -- 30 2

f- - 20

-- 10 I I I I I I I I I I

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

8-23 Babcock & Wilcox a McGermott company e....... ..

a .-

Figure 8-18. LOCA Limited Maximum Allowable Linear Heat Rate 20 i i i i i i i i i i i w

18 _

C sa - _

~

B E 16 -

/'f', -

% / s S _ / p'

? '

/

ia

/

5 14 - ~

=<

/

B - -

.5 M _ __ 0-1000 mwd /mtU

, ~ ~

--- 1000-2600 mwd /mtU After 2600 mwd /mtU -

10 i I i i i i i i i i i 0 2 '4 6 8 10 12 Axial Location of Peak Power From Bottom of Core, ft 8-24 Babcock & Wilcox a McDermott company

Figure 8-19. APSR Position Setpoints for Operation From .0 to 25+10/-0 EFPD -- ANO-1, Cycle 8 110 (9.5,102) (33.3,102)

^

100 -

RESTRICTED REGION 90

- ' (9.5,90) (43.5,90) 80

-(0,75.8) (48.5,78) y <

70 8

E 60 -

t u 50 - PERMISSIBLE

., OPERATING (100,47.1) I g 40 -

ON 2

30 20 .-

10 -

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

% Withdrawn l

l 8-25 Babcock & Wilcox a McDermott company

s : .-

Figure 8-20. APSR Position Setpoints for Operation From 25+10/-0 to 200+10 EFPD -- ANO-1, Cycle 8 110 (9.5,102) _33.3,102)

(

_ RESTRICTED 100 REGION gg _ ,(9.5,90) (43.5,90) 80 (48.5,78) 70 -

E 60 -

$ PERMISSIBLE N OPERATING 50 -

REGION o (100,47.1) I w -

. 40 5

~

h 30 20 -

10 -

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

% Withdrawn l

l l

8-26 Babcock & Wilcox a McDermott Company

a - .

Figure 8-21. APSR Position Setpoints For Operation From 200+10 to 380+10 EFPD -- ANO-1, Cycle 8 (9.5,102) (33.3,102) 100 -

RESTRICTED REGION 90 -- ' (9.5,90) '

i (0,75.8) '

(100,77.7) h 70 -

co 8 60 -

PERMISSIBLE

[ OPERATING 50 - REGION a

E 40 30 -

20 -

10- -

0 ' ' ' ' ' ' ' ' '

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

% Withdrawn 8-27 Babcock & Wilcox a McDermott Company

.e . .-

Figure 8-22. APSR Position Setpoints for Operation After

~

380 + 10 EFPD -- ANO-1, Cycle 8 110 100 90 80 -

70 -

APSR INSERTION NOT ALLOWED IN THIS. TIME INTERVAL 5 60 -

8 50 -

E de 40 -

C I 30 -

2 20 -

10 -

, 0 i ' ' ' ' ' ' '

0 10 20 30 40 50 60 70 80 90 100

% Withdrawn i

I 8-28 E E"*

a MCDermott con 9any

- - - _ . . - . . . _ _ _ . _ _ _ _ _ - _ . . _ . _ _ . . ,~... -._-_ _- _ _____.,,_.- . _ . _ . . . - . . , - - - - - _ _ _ . - - _ _ _ . - . _ - -

+ v..

l

9. STARIUP PROGRAM - PHYSICS TESTDG

'Ihe planned startup test pr % am associated with core performance is outlined below. 'Ihese tests verify that core performance is within the a = = tions of the safety analysis and provide information for continued safe operation of the unit. '

9.1. Precritical Tests 9.1.1. Control Rod Trio Test Precritical control rod drop times are recorded for all control rods at hot full-flow conditions before zero power physics testing begins. Acceptance criteria state that the rod drop time frm fully withdrawn to 75% inserted shall be less than 1.66 seconds at the conditions above.

It should be noted that safety analysis calculations are based on a rod drop frm fully withdrawn to two-thirds inserted. Since the most accurate position indication is obtained from the zone reference switch at the 75%-inserted position, this position is used instead of the two-thirds inserted position for data gathering.

9.2. Zero Power M1vsics Tests 9.2.1. Critical Boron Concentration once initial criticality is achieved, equilibrium baron is obtained and the

' critical boron concentration determined. 'Ihe critical boron concentration is calculated by correcting for any red withdrawal required to achieve equilibrium boren. 'Ihe acceptance criterion placed on critical boron concentration is that the actual baron concentration must be within i 100 ppn boron of th9t prMb+M value.

9-1 Babcock &Wilcox a McDermo;t company

4

. v..

9.2.2. Temperature Reactivity Coefficient he isothermal HZP temperature coefficient is measured at approximately the all-rods-cut configuration. During changes in temperature, reactivity feedback may be compensated by control red movement. The change in reactivity is then calculated by the summation of reactivity (obtained from a reactivity calculator strip chart recorder) associated with the temperature change. Acceptance criteria state that the measured value shall not differ from the predicted value by more than 0.4x10-4 W F.

We moderator coefficient of reactivity is calculated in conjunction with the temperature coefficient measurement. After the temperature coefficient has been measured, a predicted value of fuel Doppler coefficient of reactivity is added to obtain the moderator coefficient. 'Ihis value nust not be in excess of the acx:eptance criteria limit of +0.5x10-4 W F.

9.2.3. Control Rod Group Reactivity Worth Control rod group reactivity worths (groups 5, 6, and 7) are measured at hot zero power conditions using the boren/ rod swap method.. This technique consists of establishing a deboration rate in the reactor coolant system and ocupensating for the reactivity changes from this deboration by inserting control rod groups 7, 6, and 5 in incremental steps. We reactivity changes that occur during these measurements are calculated based on reactimeter data, and differential rod worths are obtained frcm the measured reactivity worth versus the change in red group position. W e differential rod worths of each of the controlling groups are then sn wi to obtain integral rod group worths. We acceptance criteria for the control bank group worths are as follows:

1. Individual bank 5, 6, 7 worth:

predicted value - measured value M ue x 100 5 15

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

Dredicted value - meamired value measured value x 100 5 10 l

9-2 Babcock & Wilcox 4 McDermott company

4 C* O*

9.3. Ibwer Escalation Tests 9.3.1. Core Power Distribution Verification at s40 arid 100% FP With Newnirial Control Rod Position Core power distribution tests are performed at 40 and 100% full power (FP).

'Ihe test at 40% FP is esimid.ially a check on power distribution in the core to identify any abnormalties before escalating to the 100% FP plateau.

Peaking factor criteria are applied to the 40% FP core power distribution results to determine if 75% FP tests are required prior to 100% FP operation. If these criteria are met, the 75% FP tests are not required.

'Ihe follwing acmpi an r criteria are placed on the 40% FP tests:

1. 'Ihe worst-case mavi== IHR aust be less than the IOCA limit.
2. 'Ihe minimum INBR mst be greater than 1.30.
3. 'Ihe value obtained frtan extrapolation of the mininann WBR to the next power plateau overpower trip.setpoint must be greater than 1.30,
or the extrapolated value of imhalance mst fall outside the RPS power / imbalance /flw trip envelope.

j 4. 'Ihe value obtained from extrapolation of the worst-case maximum-IHR to the next power plateau overpower trip setpoint nust be less than the fuel melt limit, or the extrapolated value of imbalance must fall outside the RPS power / imbalance /flw trip envelope.-

5. 'Ihe quadrant power tilt shall not exceed the limits specified in the

! Technical Specifications.

6. 'Ihe highest measured and predicted radial peaks shall be within the follwing limits:

i credicted value - measured value x 100 more positive than -8 measured value

7. 'Ihe highest measured and predicted total peaks shall be within the follwing limits:-

credicted value - measured value measured value x 100 more positive than -12

'Ihe power distribution test performed at 100% FP is identical to the 40% FP test except that core equilibrium xenon is established prior to the 100% FP 9-3 "" ""

a McDermott company

__ _ - - ,,__,-,,r _-..,,.-,,.,____.__..-...._,.,____..-.___._,-,,m.,. . . , _ _ . . . . . , , , , , ~ , . . , _ . . , _ _ _ , _ _

( e ,. , .

I test. Accordingly, the 100% FP measured peak acceptance criteria are as follows:

1. S e highest m a=tred and predicted radial peaks shall be within the followirx] limits:

credicted value - measured value x 100 more positive than -5 measured value

2. Se highest wamired and predicted total peaks shall be within the following limits:

n ue - w h wlue measured value x 100 more positive than -7.5 9.3.2. Incore Vs. Excore Detector Imbalance Correlation Verification Imbalances, set up in the core by control red positioning, are read simultaneously on the incore detectors and excore power range detectors.

Se excore detector offset versus incore detector offset slope must be greatcr than 0.96. If this criterion is not met, gain amplifiers on the excore detector signal processing equignent are adjusted to provide the required gain.

9.3.3. Temoerature Reactivity Coefficient at N100% FP 2e average reactor coolant temperature is decreased and then increased by about SF at constant reactor power. Se 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. Se temperature reactivity coefficient is calculated from the measured changes in reactivity and temperature.

Acceptance criteria state that the moderator temperature coefficient shall be negative.

9.3.4. Power Doooler Reactivity Coefficient at N100% FP Reactor power is decreased and then increased by about 5% FP. Se reactivity change is obtained from the change in controlling rod group position. Control rod group worth is measured using the fast.

insert / withdraw method. Reactivity corrections are made for changes in xenon and reactor coolant temperature that occur during the measurement.

9-4 Babcock &Wilcox a McDermott company

6 . .. .

2e power Doppler reactivity coefficient is calculated from the m==tred -

reactivity change, adjusted as stated above, and the m= =1 red power change.

Se ===Lred fuel Doppler coefficient must be more negative than the acceptance criteria limit of -0.90 x 10-5 ggo, p

9.4. Procedure for Use if Accentance Criteria Not Met If the acceptance criteria for any test are not met, an evaluation is performed before the test pr w ma is continued. He results of all tests will be reviewed by the plant's nuclear engineering group. If the acceptance criteria of the startup physics tests are not met, an evaluation will be performed by the plant's nuclear engineering group with assistance fran general office personnel, Middle South Services, and the fuel vendor, as needed. Se results of this evaluation will be presented to the On-site Plant Safety Committee. Resolution will be required prior to power escalation. If a safety question is involved, the Off-site Safety Review Committee would review the situation, and the NRC would be notified if an unreviewed safety question exists.

4 9-5 Babcock & Wilcom a McDermott company

s s ,w

10. REFERDICES

~

1. Arkansas Nuclear One. Unit 1 - Final Safety Analysis Recort, Docket 50-313, Arkansas Power & Light.
2. T. A. Coleman and J. T. Willse, Extended Burnup Isad Test Assembly Irradiation Frwsmu, BAW-1626, mhk & Wilcox, Lynchburg, Virginia, October 1980.
3. Arkansas Nuclear One Unit 1, Cycle 7 Reload Report, BAW-1840, Babcock &

Wilcox, Lynchburg, Virginia, August 1984.

4. BERA Retainer Design Report, BAW-1496, mWk & Wilcox, Lynchburg, Virginia, May 1978.
5. J. H. Taylor (B&W) to - S. A. Varga (NRC), I.etter, "BFRA Retainer Reinsertion," January 14, 1980.
6. Fr@ tam to Determine In-Reactor Performance of B&W Fuels - Cladding Creep Collapse, BAW-10084A. Rev. 2, mhk & Wilcox, Lynchburg, Virginia, October 1978.
7. Y. H. Hsii, et al. , TACO 2-Fuel Pin Performance Analysis, BAW-10141P-A, Rev.1, m W k & Wilcox, Lynchburg, Virginia, June 1983.
8.
  • W k & Wilcox Version of PDQ User's Manual, BAW-10117P-A, mhk &

Wilcox, Lynchburg, Virginia, January 1977.

9. NOODIE - A Multi-Dimensional 'IVo-Group Reactor Simulator, BAW-10152, mWk & Wilcox, Lynchburg, Virginia, September 1984.

~

10. Comparison of Core Physics calculations with Measurements, BAW-10120, m W k & Wilcox, Lynchburg, Virginia,-June 1978.
11. Power Peaking Nuclear Reliability Factors, BAW-10119, Babcock & Wilcox, Lynchburg, Virginia, November 1977.

10-1 Babcock & Wilcox

, a McDermott company

h* W .

12. B. R. Hao arxi J. M. Alcorn, LYNX 1: Reactor Fuel Assembly 'Iherral Hydraulic Analysis Ccde, BAW-10129-A, hhk & Wilcox, Lyncuburg, Virginia, July 1985.
13. LYNX 2: Subchannel 'Ihermal-Hydraulic Anlaysis F1Oguun, BAW-10130-A, h W k & Wilcox, Lynchburg, Virginia, July 1985. .
14. J. H. Jones, et al. , LYNXT - Core Transient 'Ihermal-Hydraulic Program, BAW-10156, hWk & Wilcox, Lynchburg, Virginia, February 1984.
15. R. L. Harne and J. H. Jones, 'Ihermal-Hydraulic Crossflow Applications, PAW-1829, hWk & Wilcox, Lynchburg, Virginia, May 1984,
16. Fuel Rod Bowing in h W k & Wilcox Fuel Designs, BAW-10147P-A. Rev. 1, h W k & Wilcox, Lynchburg, Virginia, May 1983.
17. Arkansas Nuclear One, Unit 1-Fbel Densification Report, BAW-1391, h W k & Wilcox, Lynchburg, Virginia, June 1973.
18. ECCS Analysis of B&W's 177-FA Iowered-Irop NSS, BAW-10103. Rev. 1, h W k & Wilcox, Lynchburg, Virginia, September 1975.
19. J. H. Taylor (B&W Licensing) to'R. L. Baer (Reactor Safety Branch, USNRC), Intter, July 8,1977.

10-2 Babcock &Wilcox a McDermott company

. _ . .

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