ML20153F633

Proposed Tech Specs Re Variable Low Pressure Trip Setpoint of Reactor Protection Sys

ML20153F633
Person / Time
Site:	Arkansas Nuclear
Issue date:	08/31/1988
From:	ARKANSAS POWER & LIGHT CO.
To:
Shared Package
ML20153F626	List: 1CAN088803, Application for Amend to License DPR-51,revising Variable Low Pressure Trip Setpoint for Reactor Protection Sys.Fee Paid ML20153F633
References
NUDOCS 8809070379
Download: ML20153F633 (3)

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pump (s). The pump monitors also restrict the power level for the number of pumps in operation.

C. RCS Pressure During a startup accident from low power or a slow rod withdrawal from high power, the system high pressure trip is reached before the nuclear overpower trip setpoint. The trip setting limit shown in Figure 2.3-1 for high reactor coolant system pressure (2355 psig) has been established to maintain the system pressure below the safety limit (2750 psig) for any design transient.(2)

The low pressure (1800 psig) and variable low pressure (13.89T

-6766)tripsetpointshowninFigure2.3-1havebeenestablishNg to maintain the DNB ratio greater than or equal to the minimum allowable DNB ratio for those design accidents that result in a pressure reduction.(2,s)

To account for the calibration and instrumentation errors, the accident analysis used the safety limit of Figure 2.1-1.

D. Coolant Outlet Temperature The high reactor coolant outlet temperature trip setting limit (618F) shown in Figure 2.3-1 has been established to prevent excessive core coolant temperatures in the operating range. Due to calibration and instrumentation errors, the safety analysis used a trip setpoint nf 620F, E. Reactor Building Pressure The high reactor building pressure trip setting limit (4 psig) provides positive assurance that a reactor trip will occur in the unlikely event of a steam line f.ilure in the reactor building or a loss-of-coolant accident, even in the absence of a low reactor coolant system pressure trip.

F. Shutdown Bypass In order to provide for control rod drive tests, zero power physics testing, and startup procedures, there is provision for bypassing certain segments of the reactor protection system. The reactor protection system segments which can be bypassed are shown in Table 2.3-1. Two conditions are imposed when the bypass is used:

1. A nuclear overpower trip setpoint of 55.0 percent of rated power is automatically imposed during reactor shutdown.

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

Amendment No. 2, 2I, H , 67, N A 13 0809070379 esos31 PDR ADOCK 05000313 P PDC 1 - _ _ - _

2500 P:2355 PSIG T=618'F 2355 2000

$a ACCEPTABLE OPERATION Y

a

$ 2100 E

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O 1900 Pa (13A9T g -6766) PSG - UNACCEPTABLE

$ OPERATON N P=1800 CSIG 1700 1500 560 580 600 620 640 660 REACTOR OUTLET TEMPERATURE, *F PROTECTIVE SYSTEM MAXIMUM ALLOWABLE SETPOINT Figure 2.3-1 Amendment No. 27, 49, 67, IDA 14a e

Table 2.3-1 -

Reacter Prctection System Trip Setting Linits .

One Reactor Coolant Pump Four Reactor Coolant Pumps Three Reactor Coolant Pumps Operating in Each Loop Operating (Nominal Operating (Nominal (Nominal Operating Shutdown Operating Power - 100%) Operating Power - 75%) Power - 49%) Bypass touclear power, % of 104.9 104.9 104.9 5.0(8) rsted, max NucigarPowerbasedon 1.07 times flow minus 1.07 times flow minus LO7 times flow minus Bypassed flow and imbalance, reduction due to reduction due to reduction due to

% of rated, max imbalance (s) imbalance (s) imbalance (s)

Nuc1 car Power based on NA NA 55 8ypassed pumpmonitops,%of rated, max 8

High RC system 2355 2355 2355 1720 pressure, psig, max Low RC system 1800 1800 1800 8ypassed pressure, psig, min d d d Variable low RC 13.89 T "

-6766 13.89 T "

-6766 13.89 T "

-6766 Bypassed l system pressure.

psig, min RC temp, F, max 618 618 618 618 High reactor building 4(18.7 psia) 4(18.7 psia) 4(18.7 psia) 4(18.7 pressure, psig, max psia)

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

b R2 actor coolant system flow.

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

during two pump operation.

d T

od is given in degrees Fahrenheit (F).

Amendment No. 2, 21, A3, 49, 52, 87, 92, 104 15 l

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BAW-2027 June 1988 ARKANSAS NUCLEAR ONE, UNIT 1

- Cycle 9 Reload Report -

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AREN IUCHAR CtE, UtGT 1

- Cycle 9 Reload Report -

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L IRIGCK & WIILOX Nuclear Ptuer Divisico b" P. O. Box 10935 Lyrddurg, ViIginia 24506-0935 r - - - - _ - - - - - - ,

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1. DTITOCUCTIQI NO SQtRRY . . . . . . . . . . . . . . . . . . . . . 1-1

2. OPD&TIIG IIISICRY . .... . .................. 2-1

3. GDEAL DESCRIPTIOi . ... . .................. 3-1

4. DJEL SYSIIN DESIQi . . ... . .................. 4-1 4.1. Fuel Assebly 1%chanical Eesign .............. 4-1 4.2. Fuel Rcd Ard Gray APSR Design ............... 4-2 4.2.1. Clackiin7 Collapso ................. 4-2 4.2.2. ClaMirg Stress .................. 4-3 4.2.3. ClaMirg Strain .................. 4-3 l 4-4 4.3. 'Iherral Design . . . . . . . . . . . . . . . . . . . . . . .

4.4. K'Aterial Design .. . ................... 4-4 4.5. Operatirg Experience . . . . . . . . . . . . . . . . . . . . 4'

5. IUCIIAR DESIQi . . . . ... ................... 5-1 f 5-1 5.1. Ihysics duracteristics ..................

5.2. Aralytical Input . . . . . . . . . . . . . . . . . . . . . . 5-1

%.3. Owncs In liuclear Decign. . . . . . . . . . . . . . . . . 5-1

6. 'IIIDMIrliYEFAULIC DESIGi . . ................... 6-1

7. ACCIDDTI NO 'IPRGIDir N&IXSIS ................. 7-1 7.1. General Safety Arnlysis .................. 7-1 7.2. Accident Evaltuticri . . ..... ............. 7-2 8-1

( 8. ITOICSID FDDITICATIOG 'IO 'I1X3CTICAL SIECIFICATIQG . .......

9. STARIUP ITOGPAM - UfYSICS 'IESTDG ............. ... 9-1 r

( 9.1. Procritical 'Ibsts . . ................... 9-1 9.1.1. Cbntrol Rod Trip Test ............... 9-1 9.1.2. RC Flck ... ................... 9-1 Zero Pcuer Ihysics 'Ibsts . . . . . . . . . . . . . . . . . . 9-1

( 9.2.

9.2.1. Critical Boron Cbroentration . . . . . . . . . . . . 9-1 9.2.2. Teperature Reactivity Ctefficient . . ....... 9-2 r 9.2.3. Control Rod Grmp/Doron Reactivity Worti. . ..... 9-2

( 9.3. Power re l ation 'Ibsts . . . . . . . . . . . . . . . . . . . 9-3 9.3.1. Core Sy: metry Test . ................ 9-3 iii r - - - - - - - - - - ,

Contents (Cont'd)

Pago 9.3.2. Core PcAur Distributicn Verification at Intemediato Ibwer Invol (IPL) ard 100% FP With Nonimi Ocotml Red Ibsition. . . . . . . . . . 9-3 9.3.3. Incore vs Excare Detector Irbalanco Correlation Verificatico at the IPL. . ....... 9-4 9.3.4. Torperature Reactivity Coefficient at s100% FP . .. 9-4 9.3.5. Pcuor Deppler Reactivity Cbofficient at $100% FP . . 9-5 9.4. Pmcodure for Use if Acocptance Criteria Not Met . . .... 9-5

10. Fe.ru<u1CES . . .. . .. . . . . . . . . . . . . . . . . . . . . 10-1 List of Tables Table 4-1. Fuel Design Parameters ard Dinensions ............. 4-5 4-2. Ibel Thomal Aralysis Paramotaru . . . . . . . . . . . . . . . . 4-6 4-3. Operatirn Expericrca . . . . . . . . . . . . . . . . . . . . . . 4-7 5-1. Ihysics Paramr:ters for NO-1, Cycles 8 ard 9 . . ........ 5-3 5-2. Shutdcun thrgin Calculations for NO-1, Cycle 9 ........ 5-5 6-1. Fhxirum Design Cbrditions, Cycles 8 ard 9 ........... 6-3 7-1. Ocrparison of Cycle 8 ard Cycle 9 Acx:ident Dre . . . . . . . . 7-4 7-2. Ocrparison of Fey Parameters for Accident Analysis . . . . . . . 7-5 7-3. Dourdirg Values for Allowable IOCA Peak Linor.r Hoat Rates ... 7-5 List of Ficures Figure 1

3-1. Core Imadirn Diagram for NO-1, Cycle 9 ............ 3-3 l 3-2. Enrichnent ard Eurnup Distribution, NO-1 Cycle 9 Off 440 EFTD Cycle 8 . . . . .................... 3-4 3-3. Control Rod locations ard Grtup Desigmtions for N O-1, Cycle 9 . . . .. .................... 3-5

34. IEP Enrichnent ard Distributico, NO-1, Cycle 9 ........ 3-6 4-1. Re ovable Upper Erd Fittirn (Sido View) . . . .......... 4-8 4-2. Hold $cWn Sprirq Retainer . . . . . . . . . . . . . . . . . . . . 4-9 4-3. EmA Spider. . . . . . . . . . . . . . . . . . . . . . . . . . . 4-10 4-4. BWA Spider / Upper Erd 71ttirg/Paactor Intermis Interaction. . . 4-11 4-5. Gray Axial Power Shaplig Red . ................. 4-12 5-1. N+1, Cycle 9, LOC (4 EFFD) Two-Dirensiomi Relative PcAur Distributico - Full Ptser Equilibrium Xenon, Norml Rcd Positions . .. .. .................... 5-6 8-1. Core Protection Safety Limit - NO-1, (Tech Spcc Figure 2.1-1). . . . . . . . . . . . . . . . . . . . . . . . . . 8-6 8-2. Core Protection Safety Linits - NO-1, (Toch Spec Figure 2.1-2). . . . . . .................... 8-7 iv

t List of Ficnims (Cbnt'd)

Figure 8-3. Cbre Protection Safety Limits - NO-1, (7bch Spoc Figure 2.1-3). . . . . . . . . . . . . . . . . . . . . . . . . . 8-8 8-4. Protective System Mixinum Allowable Setpoints f - N O-1, (7bch Spec Figure 2.3-2) . . . . . . . . . . . . . . . 8-12 8-5. Rod Position Setpoints for Four-nmp Opexation fran 0 to 27 +10/-O EFPD - NO-1, Cycle 9 (7bch Spoc Figure 3. 5. 2-1A) . . . . . . . . . . . . . . . . . . . 8-16 8-6. Rcd Position Setpoints for Four-hrp Operation fras 27 +10/-0 to 360 +50/-10 EFTD - N O-1, Cycle 9 (Tech Spec Figure 3.5.2-1B . . . . . . . . . . . . . . . 8-17 8-7. Rod Tbsition Setpoints for Four-amp Operatico after 360 +50/-10 DTD - NO-1, Cycle 9 (7bch Spec Figure 3.5.2-1C). . . . . . . . . . . . . . . . . . . 8-18 8-8. Rod Position Setp31nts for Throo-Rep Cperation Prem 0 to 27 +10/-0 DTD - NO-1, Cycle 9 (7bch Spoc Figure 3. 5. 2-2A) . . . . . . . . . . . . . . . . . . . 8-19 8-9. Ro.1 Position Setpoints for Three-Rep Cperation Fras 27 +10/-0 to 360 +50/-10 U7D - NO-1,

( Cycle 9 (7bch Spoc Figure 3.5.2-2B) . . . . . . . . . . . . . . . 8-20 8-10. Rcd Position Setpoints for Three-Rep Operation After 360 +50/-10 DTD - NO-1, Cycle 9 (7bch Spec Figure 3.5.2-2C) . . . . . .............. 8-21 8-11. Rod Position Setpoints for Two-hmp Operatico Fras 0 to 27 +10/-0 DTD - NO-1, Cycle 9 (Toch Spoc Figure 3.5.2-3A). . . . . . . . . . . . . . . . . . . 8-22 8-12. Rcd Position Setpoints for ho-Rmp Operatico Frun 27 +10/-0 to 360 +50/-10 DTD - NO-1, Cycle 9 (7bch Spec Figure 3.5.2-3B . . . . . . . . . . . . . .. 8-23 8-13. Ecd Pcsition Sctroints for ho-nep Cperation After 160 +50/-10 DTD - NO-1, Qtle 9 (7bch Spec Figure 3. 5. 2-3C) . . . . . . . . . . . . . . . . . . . 8-24 8-14. Operaticml Ftuer Irialame Setpoints for Operatico Frus 0 to 27 +10/-0 DTD - NO-1, Cycle 9 (7bch Spoc Figure 3. 5. 2-4A) . . . . . . . . . . . . . . . . . . . 8-25 8-15. Operatiorni Power Irtalame Setpoints for Operation

( Frus 27 +10/-0 to 360 +50/-10 DTD - NO-1, Cycle 9 (Tech Spoc Figure 3.5.2-4B) . . . . . . . . . . . . . . . 8-26 S-16. Operatierul Twer Irbalame Setpoints for Cperation After 360 d 50/-10 DTV - NO-1, Cycle 9 (7bch Spec Figure 3.5.2-4C) . . ................. 8-27 8-17. LOCA Limited Mwinn Allcuable Linear Ikut Rata NO Cycle 9 (To:h Spoc Figute 3.5.2-5) . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-28

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1. DTITODUCTICli NO SGMARY mis report justifics the cporatico of the ninth cycle of Arkansas !?oclear Ono, Unit 1 (No-1) at the rated core pcAur of 2568 mt. Incitdod are the rcquirtd amlyscs as out'inod in the US!mc docunent, "Guicianco for Prepcocd License Amer &cnts Relating to Refueling," .Tuno 1975.

% styport cycle 9 operation of NO-1, this report crploys amlytical techniques ard design bases establithod in trports that have teen suinitted to ard accepted by the US!ac ard its predemr, the USADC (eco referunces) .

We cycle 8 ard 9 reactor parametcrs relato:1 to pcAur capability are sumarized briefly in section 5 of this report. All of the accidents amlyzcd in the IEAR1 have bocn reviewed for cyclo 9 cporation. In thoco cases shcro cycle 9 duracteristics sero conservativo ccrTared to theco amlyzcd for previous cycles, new accident aml)Tes worn rot perfomcd.

20 Technical Specifications have been reviewcd, ard the rodifications requircd for cycle 9 cperation arn justificd in this rep 3rt.

Inscd on the amlysca perfornd, which take into account the postulated

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offects of fuel densificatico ard the Fimi Acceptarce Criteria for Encrtyctry Core Cooliin Systens, it has toen oorcitdod that NO-1 can be cperatcd safely 1

' for cyclo 9 at a ratcd pcher level of 2568 St.

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2. CIDATDG HISIORY 1ho reference cycle fer the ruclear ard themsi-hydraulic amlyces of Arkarcas Ibclear Ono, Unit 1 is the currently operating cyclo 8. This cycle 9 design is based cri a dcoign cycle 8 len7th of 440 offectivo full

}xuer (byn (DTO) .

The plant was cperated at 100% full gxscr for the first 2.5 ranths of gelo

8. Incr sus then rtduced in ortler to avoid a sterer 1988 refueling. 'Iho 1

plant sus cperated at 65% for 2.5 months, 80% durirn tso sternr rmths ard 70% for 2.5 renths.

Follcuiry a one ranth mintemroe outago the plant sus restarttd to 80% full pcher. Contintni cperation at 80% is plannod for the reairdor of cycle 8.

16 arrra)ics occurrtd durity cycle 8 that sulld adversely affect fuel perfomyce during gele 9.

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3. GDEDAL IHKRIPTICli

'Iho NO-1 Imactor core is dcocribed in detail in section 3 of the Arkarnas

!Mclear Ono, Unit 1, Fimi Safety Amlysis Reprt (TSAR) .1

'Iho cycle 9 core contains 177 fuel assmblics, each of which is a 15 by 15 array catainirn 208 fuol ruis, 16 contrul rod guido tubes, ard ono incore instrum.nt guido tube. 'Iho fuel is crrprised of dic?xd-eni, cylindrical pollots of uranita dioxide clad in cold-workcd Zircalcy-4.

l 'Iho fuel anscublics in all batches Mvo an average nxtim1 fuel leadin] of 463.6 kg of uranitn. 'Iho urdensified ncnimi active fuel lergths, thocro-tical dernitics, fuel and fuel rod dironsions, ard other related fuel pra-roters are givan in Tables 4-1 and 4-2 fcr all fuel asscrblics.

Figure 3-1 is the fuel chuffle dia7 ram for NO-1, cycle 9. 'Iho initial enrichrents of batchns 6D, 9B,10 ard 11 are 3.19, 3.30, 3.35, ard 3.45 wt %

U-235, respectively. Ono tatch 70 assebly, all of batch 88, ard 16 of the twice-burncd batch 9 assmblics will bo dischutJod at the crd of cycle 8.

'Iho conter locaticn will contain a ba.tch 6 assmbly dischurycd at the crd of cycle 5 (desigmted 6D) . 'Iho rtminin] 52 twico-burrrd batch 9 assmblics (dosigmted 9B) will bo chufficd to now lccaticos, with 16 on the core perit tery. 'Ibe 64 orm-turTxd tatch 10 assatlics will be shufficd to new locations, ard the 60 fresh tatch 11 anscrblics will be loadod in a syrretric e

[ choiertoard pttern thrughout the core. Figure 3-2 is an eighth-core rap shcuirg the anscrbly barTup ard enrictrent distriluticn at the bcginnin] of

[ cycle 9.

Fectivity is controllcd by 60 full-1crgth Ag-In-ci ocotrol ruis, 52 turmble l poison Ini assenblics (BWAs), ard soluble term shin. In altitico to the full-1cruth ocotrol ruis, eight Ircenal axial pcher shapin] ruds (gray AP3Rs) are pzwided for additicm1 control of the axial to.ur distribution. 'Iho cycle 9 locations of the 68 oontrol Inis and the group desigmticns are s indicated in Figum 3-3. 'Ihe core 1ccations of the total pattcIn (68 control s Inis) for cycle 9 are the sa o as thoco of the reference cycle tut the gtwp 1

3-1

designations are difforent. The locations ard enrictrents of the BmAs are rh:sn in Figure 3-4.

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Fi9ure 3-2. Enrichment and Burnup Distribution.

ANO-1 Cycle 9 off 440 EFPD Cycle 8 8 9 10 11 12 13 la 15 3,19 3.35 3.35 3.35 3.30 3.45 3,30 3.30 20770 18734 18664 18158 28714 0 20852. 31040

3. T 3.45 3.30 3.45 3.35 3.45 3.35 K

19390 0 26049 0 18549 0 15225 3,30 3.45 3.30 3.45 3.35 3.35 (

L 20853 0 23038 0 13625 17387 3.30 3.45 3.35 3.35 M

28713 0 18133 18684 3.35 3.45 3,30 18625 0 31942 3.30 30940 1

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x.xx Initial Enrichment, wt % U 235 ]

xxxxx BOC Burnup Wd/mtU

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Figure 3-3. Control Rod Locations and Group Designations for ANO-1 Cycle 9 X

Fuel Transfer l

Canal )

B 1 6 1 C 3 5 5 3 D 7 8 7 8 7 E 3 5 4 4 5 3

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F 1 0 6 2 6 8 1 G 5 4 2 2 4 5

-Y H W- 6 7 2 2 7 6 g 5 4 2 2 4 5 L 1 8 6 2 6 8 1

( H , 3 5 4 4 5 3 7 8 7 8 7 N l 0 l 3 5 5 3 P l l l 1 6 1 R I I Z

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

Group Number

( X Grouo flo of Ross Function Safety

( 2 1 8 8 Safety 3 8 Safety r 4 8 Safety 5 12 Control 6 8 Control 7 8 Control 8 8 APSRs 3-5 I

Figure 3-4. LBP Enrichment and Distribution, i ANO-1 Cycle 9 8 9 10 11 12 13 la 15 [

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H 1.355 ,

K 1.355 1.400 0.200 L 1.355 1.400 0.800 M 1.400 1.400 l

N 1.400 1.400 0 1.355 0.800 p 0.200 R

x.xxx LBP Concentration, wt % B4 C in A10 23 3-6 -

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4. FUEL SYSTDi DESIC21 4.1 Fuel Assembly Mechanical Desian The types of fuel assablies and pertinent fuel design parameters for No-1, cycle 9 are listed in Table 4-1. All fuel asserrblies are identical in w-@t. and are . 11y interchsgeable. Retainer assablies will be used on the tw fm_d assenblies containing the regenerative neutron sources (RNS). The justification for the design and use of the retainers described in referen s 2 and 3 is applicable to the RNS retainers in cycle 9 of N O-1.

The batch 11 fuel uses Zircaloy rather than Inconel as the material for the intermediate spacer grids as reported in reference 4. The NRC safety evaluation 5 of that report requires that a licensee who is incorporating that design subnit a plant-specific analysis of cxrnbined seismic and IDCA loads accordirq to Appendix A of the Standard Review plan 4.2. The analysis that was presented in reference 4 envelopes the ND-1 plant design requirements.

Therefore, the margin of safety reported for the Mark BZ fuel assembly is applicable to N O-1.

Batch 11 utilizes the MK-B6 type of assembly. The differences between this j

assembly and other MK-B types are the method used to retain fixed control L cxxponents (BIPAs, orifice rod assemblies, and regenerative neutron source corponents) during reactor operation, Zircaloy spacer grids, and the fact

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that it is reconstitutable. The rernovable upper end fitting (Figure 4-1)

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provides four open slots that align and allow designed movement of the r holcklown spring and its retainer (Figure 4-2). The fixed control ccmponent spider is shown in Figure 4-3. The holddawn e.prirg is pre-loaded through a g stcp pin welded to an ear en each side of the upper end fitting. Incore, as k shown ja Figure 4-4, the spider feet are captured between the hol& lown spring retainer and the upper grid pads cr1 the reactor internals. This arrargement

( retains the fixed control www.nts at all design flow conditions. The reencuable upper end fitting is identical to the }hrk B5 upper end fittirg except for the way it is attached to the control Itd guide tubes. The thrk B5 upper end fittiry has been tested extensively, both in air and in over 4-1

1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> of simulated reactor environment, to determine analytical inptt and to assure good incere performance.

The removable upper erd-fitting of the reconstitutable fuel asscably is a direct ckscendent of the Gadolinia Lead mast Assenbly (IT9 upper end fitting. The eM fitting design was thoroughly analyzed and tested. These results were subnitted to the NRC in reference 6. The five Gadolinia LTA's with removable upper end fittings have performed as expected. The last of the LTA's that remains in-core is in its fourth cycle and has achieved a burnup of approximately 53000 mwd /mtu. By the end of this cycle it will have reached a burnup of 59000 Mwd /nml.

The ability to reconstitute the fuel assembly has no detrimental effect on the assembly in-core perfonnance. This allows selective replacement of damaged fuel rods within an assembly, which has a tremendous ca ".-saving potential.

4.2 Fbel Rod and Gray APSR Desian l

The mechanical evaluation of the MK-B fuel rods and the gray APSR's is l

dis m W below.

4.2.1 Claddina Cbilapse A. Fuel Rod Creep collapse analyses were performcd for the four different fuel batd power histories. Because of its longer previous incore exposure tim, the batch 9B fael is nore limiting than the other batches. The batch 9B assembly power history was analyzed aM the mect limiting assembly was detemined.

The power history for the mcst limiting asserrbly was used to acrpare with a conservative generic creep collapse analysis. The collapse time for the most limiting assembly was conservatively determined to be nore than 35000 EFHi (effective full power hours), which is greater than the maxinum projected residence tire (Table 4-1). The creep collapse analysis is bascd on reference 7.

B. Gray AFSR The gray APSRs used in cycle 9 are designed for improved croop life.

Clack 11rg thickness ard rod ovality, which are the prirary factors controlling the timo until creep collapse, are improved to exterd the life of the gray 4-2

! APSR. The minirum design clackling thickness of the Mark B black APSR is 18 mils, while that of the hk-B gray APGR is 24 mils. Additionally, the gap width between the end plug and the Inconel absorber material is mduced.

Finally, the gap area ovality is controlled to tighter tolerances.

The creep collapse analysis of the gray APSR shows that it will not creep ,

collapse during the projected lifetime of the rods. The gray APSR is shown in Figure 4-5.

4.2.2 Claddirn Stress A. Fuel Rod The N O-1 cycle 9 stress parameters are enveloped by a conselJative fuel rod stress analysis. The same method was used for amlysis of cycle 9 that had been used on the previous cycle.

B. Gray APSR The gray APSR design demonstrates the ability to moet wified design requirements. The APSR cladding stress analysis includes pressure, tenperature ard ovality effects. The gray APSR has sufficient cladding and weld stress margins.

4.2.3 Claddirn Strain A. Puel Rod The strain analysis is based on the upper tolerance values for the fuel

( pellet diameter and density and the lower tolerance for the cladding inside diameter. The fuel design criteria specify a 1.0% limit on cladding plastic tensile circumferential strain. The pellet is designed to ensure that

{ plastic cladding strain is less than 1.0% at design local pellet burnup ard heat generation rate. The design values are higher than the worst-caso values the NO-1 cycle 9 fuel is expected to see.

B. Gray MSB The gray APSR strain analysis includes thcmnal and irradiation sellirg effects. The results of this analysis show that no clackling strain is irduced due to thermal expansion or irradiation selling of the In:onel absorter.

4-3

.. J

4.3 'Ibermal Desian All fuel assenblies in the cycle 9 core are thermally similar. The design of the batch 11 Mark B6 assenblies is such that the thermal performance of this fuel is equivalent to the fuel design used in the remainder of the core. The f analysis for all fuel was performed with the TACD2 code as hibed in reference 8. Nominal undensified input parameters used in the thermal analysis are pres, v.ted in Table 4-2. Densification effects were accounted for in TACO 2.

'Ihe results et the thermal design evaluation of the, cycle 9 core are sumarized in Table 4-2. Cycle 9 core protection limits are haami on a linear heat tate (I.HR) to centerlira fuel raalt limit of 20.5 kW/ft as deter-mined by th e TACD2 code.

The maxirun fuel assemoly bumup at DDC 9 is predicted to be less tlan 42,800 Wd/mtU (batch 9B) . The fuel red internal pressures have been evaluated with ,

TA002 for the highest burnup fuel rods and are predicted to be less than the ncminal reactor coolant pressure of 2200 psia. -

4.4 Material Desian The chemical capatibility of p:ssible fuel-cladding-coolant-assenbly interactions for batch 11 fuel assablies is identical to those for previous fuel assecblies because no new materials were introduced in the batch 11 fuel assablies.

4.5. Operatim ExDerience nahk & Wilcox operatirs experience with the Mark B 15x15 fuel assmbly has verified the adequacy of its design. 7he accumulated cre. rating experience for eight B&W 177 fuel assably plants with Mark B fuel is shown in Table 4-j 3. ]

l i

l 4-4

i

[

Table 4 Fuel Desian Parameters and Dimensions

[

Batch 6D Bat & 9B Batch 10 Batch 11 Fuel assesnbly type MK-B4 MK-B4 MK-B4 MK-B6 Nunter of assemblies 1 52 64 60 s

Ibel rod OD ncaninal, 0.430 0.430 0.430 0.430 in Fuel red ID naninal, 0.377 0.377 0.377 0.377 Undensified active 142.25 141.8 141.8 141.8 foal length, in Fuel pellet OD, 0.3695 0.3686 0.3686 0.3686 (mean), in Fuel-pellet initial 94 95 95 95 density, (Ncan), % 'ID Initial fuel enrichment, 3.19 3.30 3.35 3.45 wt. % U-235

/

Average burnup, BOC, 20800 26400 17300 0 mwd /mtU i Exposure time, EDC, 28700 31300 20600 10100 EFIH Cladding collapse >35000 >35000 >35000 >35000 L time, EFEH I

e ,

4-5 f - _ _ - - - - - - - - _ - - - - - - - -_

Table 4-2. Fuel 'Ihermal Analysis Parameters i

Batch 6D Batch 9B Batd1 10 Batch 11 No. of assenblies 52 64 60 1

f Initial density, % 'ID 94 95 95 95 t

Initial pellet 00, in 0.3695 0.3686 0.3686 0.3686 Initial stack height, in 142.25 141.80 141.80 141.80 l 3.35 3.45 Enrichment, wt % U-235 3.19 3.30 Naninal linear heat rate 5.73 5.74 5.74 5.74 at 2568 Mt, W/ft(a)

TACO 2-based, Predictions Average fuel tenperature at ncani))al IRR, F (BOL) 1406 1400 1400 1400 ,

Mininun IRR to melt, W/ft 20.5 20.5 20.5 20.5 Core average DiR = 5.74 W/ft

(

(a) Based on a naninal stack height

[

~

4-6

Table 4-3. Goeratim Experience I

Cumulative Current Max FA burmo.Mki/mtU(a) electric Reactor Cvele Incore Discharm d outpfuHi}(b) f Ocones 1 10 45,908 50,598 66,183,044 Oconee 2 9 40,580 41,592 60,968,626 Ooonee 3 10 33,290 39,701 60,843,663

'Ihree Mile Islani 6 26,090 33,444 2:s,4ew, vie Arhnsas Nuclear or,e, 8 51,540 47,560 51,626,035 Unit 1 Rancho Seco 7 26,242 38,268 39,045,954 C::ystal River 3 6 35,370 31,420 38,512,798 Davis-Ihsse 5 36,960 32,790 25,236,663 (a)As of October 31, 1987.

(b)As of December 31, 1986.

(

r L

( '

4-7

[ . - - - - - - - - - - - - - -

Figure 4-1. Removable Upoer End Fitting (Side Vie.,)

SLOT STOP E* f N - '[

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l 3 F '

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• There are two stop pin holes on each side of the upper end fitting. One contains a stop pin and the other is a spare.

1 4-8

Figure 4-2. Holddown Spring Retainer FOOT k

j ~ ;d 7__

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4-10 l

Figure 4-4. BPRA Spider / Upper End Fitting /

Reactor Internals Interaction v-I t

REACTOR INTERNALS , f UPPER GRID PAD STOP ASSEME.Y PIN

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5. NUCIEAR DESIGN 5.1. mvsics Characteristics Table 5-1 lists the core physics parameters of design cycles 8 and 9. 'Ihe values for both cycles were calculated with the N00 DIE code9 . Figure 5-1 illustrates a representative relative power distribution for the beginning of cycle 9 at full power with equilibrium xenon and nminal red positions.

'Ibe differences in feed enrichment, BPRA loading, and shuffle pattern caused little change in the physics parameters between cycles 8 and 9. Calculated ejected rod worths and their adherence to criteria are considered at all times in life and at all power levels in the developnent of the rod position limits presented in secticn 8. 'Ihe maximum stuck red worths for cycle 9 are less than for cycle 8 at all times in cycle. All safety criteria associated with these worths are met. 'Ibe adequacy of the shutdown margin with cycle 9 stuck rod worths is demonstrated in Table 5-2. 'Ihe following conservaties were applied for the shutdown calculations:

1. Poison material depletion allch'ance.

2. 10% urcertainty on net rod worth.

3. Flux redistribution penalty.

Flux redistribution was accounted for since the shutdcAn analysis was .

calculated using a thional nodel. 'Ihe reference fuel evele shutdcr.n

( margin is presented in the No-1 cycle 8 reload report.10 1

5.2. Analvtical Irrut

'Ihe cycle 9 incore measurunent calculation constants to be used for ccrputin; core power distributions were prepared in the same manner as those for the reference cycle. l 5.3. Chances in Ihrlear Desian l

'Ihe core design charges for cycle 9 are the use of gray AP5Rs and the replacement of the Inconel interrediate spacer grids with Zircaloy spacer grids. Gray Alus, which are lorger ard use a wa9mr absorter (Inconel),

5-1 f _ - - - - -

replace the silver-iniium-cadmium APSRs used in all previous cycles. i calculations with the standard three-dimensional 1xx$el verifiM that these APSRs provide adequate axial power distribution control. 'Ibe substitution of Zircalcy spacer grids redu s the parasitic absorption of trutzuns and has a beneficial effect on fuel cycle cost.

f l

Mm gray APSRs will be withdrawn frtn the core near the end of cycle 9 (360 EFFD) where the stability and control of the core in the feed-and-bleed rrode with APSRs ruoved has been analyzed. 'Ibe calculated stability index at 364 DTD without APSRs is -0.037 h-1 whidt desnonstrates the axial stability of the core. 'Ihe calculational methods used to obtain the irportant nuclear design parameters for this cycle are the same as those used for the j reference cycle. 'Ibe operating limits (Technical Specifications changes) for (

the reload cycle are given in section 8.

l l

I 1

5-2

__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - - _ l

Table 5-1. Ihysics Parameters for Ato-1. Cveles 8 and 9(a)

Cycle 8(D) Cvele 9(C)

Cycle length, EFIV 420 420 Cycle burt1up,IWd/mtU 13,147 13,143 Average core burT1up - EOC, }Nd/mtU 25,522 27,271 Initial core loading, atU 82.0 82.1 f

Critical boren - BOC, ppu (no Xe)

}EP(d), group 8 inserted 1644 1552 1 p , group 8 inserted 1409 1379 Critical boren - EOC, ppu (eq Xe) l IEP, group 8 inserted 651 HFP, group 8 inserted 18 539(e)

O

(

Control red worths - HFP, BOC, % aP/k Group 6 1.14 1.11 Group 7 1.49 0.98 Group 8 (maximum) 0.39 0.19 Control red worths - HFP, DDC, % AP/k G'.oup 7 1.52 1.05 Max ejected rod worth - 1EP, % AP/k

[ BOC (Ie10), groups 5-8 ins 0.55 0.35 360 EtTO (Ie10), grtJps 5-8 ins 0.60 0.41 DOC (Ie10), groups 5-7 ins 0.59 0.41

( Max stuck rod worth - IEP, % AP/k BOC (N-12), groups 1-8 ins 1.58 1.49

( 360 EFID (N-12), groups 1-8 ins 1.86 1.47 DDC (N-12), groups 1-7 ins 1.63 1.42 Power deficit, IEP to IW, % AP/k BDC 1.56 1.60 D3C 2.34 2.35

{

Doppler coeff - HFP,10~4 (6P/P/ 0F)

BDC (no %) -0.154 -0.159 IDC (eq Xe) -0.184 -0.186 5-3 I. .____ _ - - - - _ - - - - -

Table 5-1. (Carit'd) (a)

Cvele 8(b) Cvele 9(C)

Moderator coeff - HFP,10-4 (AM)

IOC (no Xe, crit ppa, group 8 ins) -0.51 -0.58 '

I B)C (eg Xe, O ppa, group 8 out) -2.78 -2.82 Boron worth - HFP, pprV% ok/k IOC 129 130 BOC 111 111 Xenon worth - HFP, % Ak/k IOC (4 EFFD) 2.55 2.56 EOC (equilibrium) 2.72 2.71 Effective delayed neutron fraction - HFP ICC 0.0062 0.0062 l EOC 0.0052 0.0052 l

! (a) Cycle 9 data are for the conditions stated in this report. 'Ibe cycle 8 l core conditions are identified in reference 9.

i (b)nn,md on 425 EFFD at 2568 m t, cycle 7.

(c) Based on 440 EFPD at 2568 Wt, cycle 8.

(d)HZP denotes hot zero power (532F Tavg); HFP denotes hot full power (581F Tavg)-

(e)At HFP conditions, O gxn occurs at 411 ETED.

]

l 5-4

l i

, Table 5-2. Shttih 41 Maruin Calculation for ANO-1. Cvele 9 I BOC, 360 EFPD, 420 EFPD,

% Ak/k  % Ak/k  % Ak/k Available Rod Worth

'Ibtal rod worth, HZP 8.699 9.265 9.222 f Worth reduction due to poison material burnup -0.100 -0.100 -0.100 Maxt=wa stuck rod, HZP -1.490 -1.471 -1.419

{

Net Worth 7.109 7.694 7.703 Iess 10% uncertainty -0.711 -0.769 -0.770

'Ibtal available worth 6.398 6.925 6.933 Reautred Red Worth

( Power deficit, HFP to HZP 1.602 2.291 2.351 Allcuable inserted rod worth 0.276 0.422 0.422 f

Flux redistribution 0.344 0.616 0.573

'Ibtal required Wrth 2.222 3.329 3.346 r shutdown margin (total L available worth minus

• total required worth) 4.176 3.596 3.587 LQIE: 'Ibe required shutdown margin is 1.00% Ak/k.

I

[

[

[ 5-5 '

( .

Figure 5-1. ANO-1 Cycle 9, BOC (4 EFPD) Two-Dimensional Relative Power Distribution -- Full Power, Equilibrium Xenon, Normal Rod Position 1 8 9 10 11 12 13 11 15 H 0.98 1.12 1.10 1.13 1.02 1.29 0.94 0.41 K 1.13 1.12 1.29 1.05 1.29 1.24 1.17 0.52 L 1.19 1.31 1.16 1.28 1.12 1.31 0.93 0.39 1

l M 1.14 1.06 1.28 1.04 1.28 1.07 0.61 N 1.02 1.29 1.12 1.28 1.16 1.09 0.32 l

l 0 1.29 1.25 1.31 1.07 1.09 0.42

! P 0.94 1.18 0.93 0.61 0.32 i

R 0.41 0.53 0.39 Inserted Rod group No.

x.xx Relative Power Dansity 5-6

- - _ - - _ - _ . _ _ - - - - - _ _ _ _ - - _ - - _)

1

6. THERMAL-HYERAULIC DESIGN The thermal-hydraulic design evaluation supporting cycle 9 operation utilized the methods and nodels described in references 10,11, and 12 as supplemented by reference 4, which inplements the BWC (reference 13) QiF correlation for

{

analysis of the Zircaloy grid fuel assembly. The analyses presented in Section 5 of reference 4 di:awatrate that changes in the flow parameters resulting frun the incorporation of Zircaloy spacer grids do not significantly inpact the themal-hydraulic characteristics of the Zircaloy grid core relative to the Inconel grid core values. Inplementation of the Zircaloy grid fuel assenblies into existing reactors, however, is perfomed on a batch basis, with the transition cycles having both Zircaloy grid and Inconel g..d fuel assemblies.

In a transition core, the Zircaloy grid fuel assemblies, which have a slightly higher pressure drop than the Inconel grid assemblies due to the higher flow resistance of the Zircaloy grids, terd to divert scoe flow to the f

Inconel grid fuel. This creates the need to consider a "transition core penalty". The amount of coolant flow reduction in the limiting Zircaloy grid assecbly ani consequently, the magnitude of the transition pemity, is e

dependent on the number of Zircaloy grid assecblies (with the smaller nu:nbar I of Zirdoy grid assc211es being more limitirg).

Another contributing factor in deteminiry the transition core penalty is the core bypass fraction which is dependent on the nucber of turmble poison rod assemblies (BmAs), since these ccrponents restrict flow through the control

[ rcd guide tubes (CETs) . For themal-hydraulic analyses, the most limiting case is that with the higher bypass flow fraction, or smaller nu&cr of BmAs.

The design basis chosen for cycle 9 themal-hitiraulic analyses was a full core of Zircaloy grid asserblics, containirg 40 BEAs, for khich the core bypass flow is 8.8%. This design configuration was used to calculate the l.77 EtHR value shcun on Table 6-1. The actual cycle 9 core configuration

{ consists of 60 fresh Zirtaloy grid fuel assablies, of khich 52 contain BmAs 6-1

(8.3% core Ir/ pass flow) . 'Ihe DE for this configuration, usirg the same core conditions presented in Table 6-1, is 1.80. 'Ibe full Zircaloy grid core configuration is, therefore, conservative for cycle 9 DM analyses and a transition core penalty is not evry. 'Ibe rmstitutable upper end fittirq (UEF) ard the anti-straddle lower end fitting (IEF) were addressed in the evaluation. f

'Ihe pressure-ter:perature safety limits have been recalculated usinJ the BWC OF correlation in the LY!TXT11 crossflow analysis. Table 6-1 prtnides a sumary ccmparison of the DS analysis parameters for cycles 8 and 9.

No rod bow penalty has been considered in the cycle 9 analysis based on the justification prtwided by reference 14. Reference 14 was verified as applicable for Zircaloy grid fuel assemblies in reference 4.

l l

6-2 l

Table 6-1. Maximm Desian conditions. Cveles 8 and 9 Cycle 8 Cycle 9 Design power level,)Wt 2568 2568 System pressure, psia 2200 2200 Deactor coolant flw, gpn 374880 374880 Core bypass fl w , % (a) 8.4 8.8 De waialing crossflw crossf1w

[

Reference design radial-local power peaking factor 1.71 1.71 Reference design axial flux shape 1.65 cosine 1.65 cosino Hot channel factors

{ Enthalpy rise 1.011 1.011 Heat a ux 1.014 1.014 F1w area 0.98 0.97 (c)

Active fuel length, in. (b) 141.8 141.8 l Avg heat flux at 100% power, 103 Btu /h-ft2 174 174 Max heat flux at 100% power,

( 103 M -ft2 492 492 r QF correlation B&W-2 BWC l

QF correlation DiB limit 1.3 1.18

( Mininn Da at 112% power 2.08 1.77 (c) at 102% power (d) 2.37 2.01 (a)Used in the analysis.

(b) Cold ncair .d stack height.

(C)Calmlated for the instnment guide tube subchannel Wich is limiting for the Mark-B6 assa bly.

(d)'Ihis represents initial coniition DiBR for accident analyses.

( 6-3 f - _ - - - - - - - - -

7. ACCIDE2C AND TPRISIDE NRLYSIS 7.1. General Safety Analysis Each FSAR accident analysis has been examined with respect to chames in cycle 9 parameters to detemine the effect of the cycle 9 reload and to ensure that themal performance during hypothetical transients is not I degraded.

The effect of fuel densification on the FSAR accident asults has been evaluated and are riported in reference 15. Since batch 11 reload fuel ,

asserblies contain fuel rods whose theoretical density is higher than those f considered in the reference report, the conclusions in that reference are still valid. l

'% radiological dose consequen s of the accidents presented in Chapter 14 of the updated FSAR were re-evaluated for this reload report except for the waste gas tank rupture. The waste gas tank rupture was not reevaluated since

(

Technical Specification 3.25.2.5 controls the maximum tank inventory on the basis of Xe-133 equivalent curie content such that the analysis of the event is not cycle dependent. The evaluation of the remainig events was made in order to incorporate more current plant data as well as the infomation in the up: lated FSAR.

All of the Cycle 9 accident doses are based on radionuclide sources calculated for the actual Cycle 9 coro design and irradiation history. In addition, the bases used to analyze scos of these accidents were chamed to r

L te consistent with the bases in the updated ISAR. The significant differences in the bases for the accident an11ysis betwocn cycle 8 and cycle

( 9 aret o The atrocpheric dispersion factors have been in:reased slightly.

g o credit has been taken for the penetration room filter systan in L calculating the h associated with the control rod ejectico accident. (This makes the control rod ejection accident censistent

( with the 1DCA and MM.)

7-1 I - - - - - - - - - - - - -

i o 'Ihe iodine removal rate used to calculate the IJDCA ard MHA doses for Cycle 9 was changed to be consistent with the updated FSAR.

All of the calculated cycle 9 accident doses are below the dose acceptance criteria that are specified in the NRC's Standard Review Plan (NUREG-0800).

Table 7-1 shows a ocuparison between cycle 8 and cycle 9 dnaca for the Chapter 14 accidents that result in significant offsite anaam. With the exception of the maxinna hypothetical accident (MR), all anaaa are either bounded by the values reported for cycle 8 or are a small fraction of the 10CFR100 limits, i.e. , below 30 Rem to the thyroid or 2.5 Ren to the whole J body. For the MHA, the doses ocmpare to the criteria as follows:

1 1. 'Ihe 2-hour thyroid dose at the exclusion area boundary (EAB) is 165.1 Rem (55% of the IURID-0800 limit) .

l 2. 'Ibe 2-hour whole-body dose at the EAB is 5.0 Rem (20% of the IURID-0800 limit).

l l 3. 7he 30 day thyroid dose at the low population zone (LPZ) is 87.8 Rem (29% of the ! URIN-0800 limit) .

i

'Iho radiological doses fran all of the accidents evaluated with the specific nuclide inventory frcn cycle 9 are lower than the NRC acceptance criteria of

! URIN-0800, and thus are within acceptable limits.

L2 Accident Evaluatien

'Ihe key paranoters that hava the greatest effect on detemining the outocre of a transient can typically be classified in three major areast core thermal parameters, themal-hydraulic paranoters, and kinetics paramoters, )

including the reactivity feedback coefficients and control red worths.

'Ihe core thermi prtperties used in the I'SAR accident analysis were design cperatirg values based cm calollational values plus uncertainties. 'Ibermi paranaters for fuel batches 60, 9B,10, ard 11 are giwa in Table 4-2. 'Ihe cycle 9 tMrml-hydraulic nuinra design conditions are cxrptred with the pzwvious cycle 8 values in Tabin 6-1. 'Ibese }nramotors are ccrron to all the 7-2

{

accidents considered in this report. Se key kinetics parameters frcan the ISAR ard cycle 9 are ccupared 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-1010316, BAW-1010417, ard BAW-1915P18). Eis analysis is generic since the f limiting values of key parameters for all plants in this category were used.

Furthermore, the ocanbination of average fuel testperatures as a function of IIR and lifetime pin pressure data used in the BAW-1915P IOCA limits analysis

[

is conservative ccupared to those calculated for this reload. Bus, the analysis ard the IDCA limits reported in BAW-1915P provide conservative results for the operation of the reload cycle. Table 7-3 shows the bounding values for allowable IDCA peak Ilms for NO-1 cycle 9 fuel. % ese IJR limits include the effects of NURB3 0630, TACD2, ard FIICSET.

It is concluded frun the examination of cycle 9 core thermal ard kinetics properties, with respect to acceptable previous cycle values, that this core reload will not adversely affect the MD-1 plant's ability to operate safely durirq cycle 9. CcnsiderinJ the previously accepted design basis used in the FSAR ard subsequent cycles, the transient evaluation of cycle 9 is considered to be bourded by previously accepted analyses. We initial conditions for

{

the transients in cycle 9 are tourded by the FSAR, the fuel densification report, ard/or subacquent cycle analyses.

(

(

(

(

(

(

7-3

( _ --

Table 7-1. %H_a_m of Cvele 8 and Cvele 9 Accident h Cycle 8 h , Cycle 9 doses, Ren Ren Fuel Handlim Accident

%yroid does at FAB (2 h) 3.15 1.12 1 hole body dose at EAB (2 h) 0.21 0.22 1

Steam Line Break hyroid does at EAB (2 h) 1.71 1.82

1 hole bo.1y dose at EAB (2 h) 0.008 0.01 l

l S+mam Generator %t Failuru

% yroid does at EAB (2 h) 6.14 6.53 1 hole body dose at EAB (2 h) 0.52 0.56 Control Rod EMection Accident hyroid dose at EAB (2 h) 12.2 7.02 Nhole body dose at EAB (2 h) 0.008 0.006

%yroid dose at IPZ (30 d) 9.09 5.64 Whole body dose at IPZ (30 d) 0.005 0.005

]

10 3

%yroid dose at EAB (2 h) 4.02 4.22 )

Whole body dose at EAB (2 h) 0.026 0.03 W yroid dose at IPZ (30 d) 2.05 2.47 Whole body dose at LPZ (30 d) 0.018 0.02 Maxinun itmothetical Accident Wyroid dose at EAB (2 h) 157.3 165.1 b le body dose at EAB (2 h) 4.80 5.03 Wyroid does at IJ'Z (30 d) 73.0 87.8 1 hole body dose at IJ'Z (30 d) 1.56 1.78

)

7-4

Table 7-2. Ccrpariccn of Key Parameters for Accirient Analysis MIAR and Densification NO-1 Paramter Prcort Value Cycle 9 Doppler coeff (IOC),10-4 AP/P/OF -0.117 -0.159 Mppler coeff (EDC),10-4 AP/P/OF -0.130 -0.186 Ftdorator coeff (IDC),10-4 AP/P/OF 0.0(a) -0.58 Ftdcrator coeff (EDC),10-4 ok/P/OF -4.G(b) -2.82 All-rul group korth (!!2P), % ok/k 12.90 8.70 Initial borcn cen entration, ppra 1159 1379 Doren reactivity korth (IIFP), 100 130 ppy% AP/k thx. ejectcd red korth (ID), % AP/k 0.65 0.23 Dropped rod korth (IR), % ak/k 0.65 50.20 (a) +0.5 x 10-4 Ak/P/OF kus uscd for the acdcrator dilution analysis.

(b) _3,o x 1o- 4 3pjyjoF was uced for the steam lino failuro analysis.

Tablo 7-3. Dcurding Values for AllcAuble ICCA 1 Yak Linear Heat 3119s AllcAnble Allcwablo Coro peak Um, peak UM, elevation, 0-1000 M11/ctU, after 1000 FRl/mtU, ft .)M/ft 19/ft

( 2 14.0 15.5 4 16.1 16.6

( 6 16.5 18.0 8 17.0 17.0

(

10 16.0 16.0 i 7-5

_ _ _ _ _ _ _ _ _ _ _ _ J

f I

8. HOPOSED )CDITICATIQG 70 TEQUICAL SPECIFICATIGE

{

f the 'Iwinical Specifications have been revised for cycle 9 operation for charges in core reactivity, power peaking, and control red worths. 7he Technical Specifications were also revised to deu: ribe. design features

(

implananted with cycle 9. 'Ihe cycle 9 design analysis basis includes the impact of extended periods of cycle 8 low-power operation, with cycle 8 power levels rarging between 65% and 100% of rated power. The cycle 9 basis also includes a very low leakage fuel cycle design, a mixed Mark N/ Mark B6 fuel asset / core, gray APSRs, gray APSR withdrawal flexibility, and crossflow analysLs. The safety limits in Twinical Specification Section 2 (Figures 8-1

( through 6-3), have been changed for cycle-specific credits in the fuel cycle design, which allowed for additional operating margin beyce,i the generic j limits used for cycle 8. Error adjusted trip setpoints for the reactor protection system .=re shown in Figura C-4. '1he IOCA linear heat rate limits used to develcp the Technical Specification Limiting conditions for Cperation include the impact of NURm-0630 cladding swel. and rupture model, and duplement the credit frun FIICSET analyses.18 A cycle 9 specific analysis was conducted to generate 7tchnical Specification

( Limiting Conditicos for Operation (rod irdex, axial power imbalance, ard quadrant tilt), Mwi on the methodology described in refereno's 19. 'Ihe etfacts of gray AFSR repositionirg were included in the analysis, as was en

{

APSR withcrawal flexibility wirdow of +50/-10 LTFOs. The burmp-deperdent allowable loCA linear heat rate limits used in the analvsis are prwided in Figure 8-17. '1he analysis also dctamincd that the cycle 9 'Dechnical Specifications prwide protection for the overpower condition that could

( cocur durire an overcoolirg trarsient because of ruclear instrunentation errors, and verified rurval cf the power level cutoff hold requirenant.

Technical Specification section 3 5.2.4 was revised to acecrtcdate a change in the quadrant tilt setpoint. 'Ihe measurtrent system-irdeperdent rod

{

positicn ard axial power irnbalance limits detemined by the cycle 9 analysis 8-1 r _- _

l were error-adjusted to generate alam *mtpoints for power optration and are reflected in a hinical Specification revision to sections 3.5.2.5 and l 3.5.2.6. 'Em error adjusted alam setpoints are prwided in Figures 8-5 through 8-16. hinical Specification section 5.3.1 was revised to include l the tw.stitutable fuel assembly design ani gray axial power shaping rods in the design featurns.

)

Bamm1 on the analyses ard hinical Specification revisions describad in thia ruport, the Final Acceptance Critaria ECCS limits will not be exceeded, nor will the thermal design critaria be violated. 'Ihe following pages contain the revisions to the M inical Specifications. }

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8-2

- --- --- -- - - - - - J

2. SAFTIY LIMITS AND LIMITDC SAFETI SYSTEM SETITdg6 2.1 SAFEIY LIMITS, REACIOR CORE Aeolicability

{

Applies to reactor themal power, reaccor power inbalance, reactor coolant

system pressure, coolant terperature, ard coolant flow during power operation i of the plant.

Obiective

'Ib maintain the integrity of the fuel cladding.

j Suecification 2.1.1 'Ihe ocrbination of the reactor syste pressure and coolant teperature shall not exceed the safety limit as defined by the

( locus of points established in Figure 2.1-1. If the actual pressure /tmperature point is below ard to the right of the pressure /terperature line the safety limit is exmaahi.

2.1.2 'Ibe ocebination of reactor thermal power ard reactor pcw.r imbalance (power in the tcp half of the core minus the power in the

(

bottm half of the core expr-ai as a percentage of the rated I power) shall not exceed the safety limit as defined by the locus of points for the specified flow set forth in Figure 2.1-2. If the actual-reactor-thermal-pomr/ reactor-powes,-inbalance point is above the Ifne for the specified flow, the safety limit is exnaa%4

(

Pases To maintain the integrity of the fuel cladding and to prevent fission product release, it is r-my to prevent overheating of the cladding un$er nomal cperating conittf ons. This is acecrplished by operating within the rucleate

( boiling regime of heat transfer, wherein the heat transfer coefficient is large enough so that the clad surface taperature is only slightly greater than the coolant tarperature. 'Ihe upper boandary of the rucleate boilirg regime is termed departuru frm nucleata boilirs (DE). At this point there

( is a sharp reduction of the heat transfer coefficient, which could result in l high cladding taperatures and the possibility of cladding failure. Although r

DiB is not an nhaarvable parameter during reactor operation, the nhnervable l parameters or neutrta power, reactor coolant flow, tapemture, ard prer.sure can be related to Da through the use of a critical heat flux (OIF) correlation. 'Ihe B&W-2(1) and BWC(2) correlations have been develcped to

( predict Da ard the location of Da for axially unifom and non-unifom heat flux distributions. 'Ibe NW-2 correlation applies to Mark-B fuel ard the EWC correlation applies to Mark-BZ fuel. 'Ibe lccal Da ratio (DER), defined as the ratio of the heat flux that would cause Da at a :particular coce location

( to the actud heat flux, is inlicative of the marg;.n to Dm. 'Ibe minirum value of the DER, durity steady-state coeration, nomal cperatienti transists, ard anticipated transients is lin:,ted to 1.30 (MW-2) ard 1.18 (IWC) .

[ 8-3 f - -

A DE of 1.30 (NW-2) or 1.18 (BWC) correspords to a 95 percent probability l at a 95 percent confidence level that DB will not occur; this is considered l a conservative margin to Da for all operating coalitions. The difference between the actual core outlet pressure ard the iniicated reactor coc,lant system pressure for the allcwable reactor coolant punp cambination has been l 1 considered in determinirg the core protection safety limits. I J The curve presented in Figure 2.1-1 represents the corriitions at which the 3 DE is greater than or equal to the minimum allowable DE t'or the limiting j ccznbination of therral power anl nurber of operatirg reactor coolant punps.

This curve is base 1 on the follcuing mclear pwer peakirg factors (3) with potential fuel densification effects: ]

q.2.83,go.1.mg.1.6s.

The Mrves of Figure 2.1-2 are based on the rcre restrictive of two therral limits and include the effects of potential fuel densification:

1. The DE limit prcduced by a nuclear power peakirg factor of =

2.83 or the ocrnbination of the radial peali axial peak and posit on of the axial peak that yields no less than the DE limit.

]

2. The cx2nbimtion of radial and axial peak that prevents central fuel reltirg at the hot spot. The limit is 20.5 kW/ft.

Ruer peaking is not a directly observable quantity ard thereforn limits have l been established on the basis of the reactor power imbalance produccd by the l

power peakirg. )

The flow rates for curves 1, 2, ard 3 of Figure 2.1-3 cottu ic-d to the expected minimum flow rates with four purps, three punps, and one purp in 1 each locp, respectively. J The ouve of Figure 2.1-1 is the most restrictive of all possible reactor u:aolant pu p mxinum tMul power cxxbinations shown in Figure 2.1-3. The )

curves of Figure 2.1-3 represent the conditions at which the DE limit is l praticted at the maximum possible thernal power for the nunber of reactor coolant prps in cperation. The local quality at the point of mu.irum DE i

]

is less than 22 perrent (B&W-2)(1) or 26 percent (EWC)(2) . I Using a local quality limit of 22 percent (B&W-2) or 26 percent (BWC) at the 1 point of minirum DE as a basis for curves 2 crd 3 of Figure 2.1-3 is a J conservative criterion even though the quality at the exit is higher than the quality at the point of minimum DR.

The DE as calculated bf the NW-2 or the ENC correlation continually l increases fren point of minirun DE, so that the exit DE is always higher ard is a function of the pressure. .

The maximum therm 1 power, as a function of reactor coolant pep cperation is limitai by the power level trip prcduced by the flux-flew ratio (percent flow x flux-flow ratio), plus the appropriate calibration ard instrunentation )

errurs.

8-4

l For each curve of Figure 2.1-3, a pressure-terperature point above and to the left of the curve Wuld rest 9.t in a DM greater than 1.30 (B&W-2) or 1.18 (BWC) or a local quality at the point of miniman Da less than 22 percent (B&W-2) or 26 percent (IHC) for that particular reactor coolant punp situation. Curve 1 of Figure 2.1-3 is the nest restrictive because any pressure-tartparature point above and to the left of this curve will be above arx1 to the left of the other curves. l PDTRDCES (1) Correlation of Critical Heat Flux in a Bandle Cboled bf Pressurized Water, BAW-10000A, May, 1976.

(2) BWC Cbrrelation of Critical Heat Flux, BAW-10143P-A, April,1985.

(3) ISAR, Section 3.2.3.1.1.c.

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8-5 r --- - - - - - - - - - - - - -

Figure 8-1. Core Protection Safety Limit -- ANO-1 (Tech $pec Figure 2.1-1) 2400 2200 en ACCEPTABLE OPERATION

. j 4 >

$ 2000 O

t I UNACCEPTABLE t OPERATION C

8

$ 1800 v

1600

• l 580 600 620 640 660 Reactor Outlet Temperature. OF ]

]

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)

8-6

r Figure 8-2. Core Protection Safety Limits -- ANO-1

{ (Tech Spec Figure 2.1-2) l Thermal Power Level l  % FP

- 140

- 120

(-33.04.112.0) -

(33.04.112.0) j l OPER N

- - 100 l (45.27.100.55)

^

l I

(-33.04.90.75) l(33.04.90.75)

(-62.32.84.45) gfT OPERATION

-- 80 l

l p(45.27.79.30) l I I l I I #

(-62.32.63.20) ~~ l (33.04.64.08)

(-33.04.64.08;1 ACCEPTABLE l

l 4.3&2 PUMP >(45.27.52.63)

[ l OPERATION l

l l -

- 40 l l

[ (-62.32.36.53) l l l I I I

[ 1

- 20 l  ;

l I I I I l i I I l li i l i i i i i i

-60 -40 -20 0 20 40 60

( Reactor Power Imbalance, t

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8-7

Figure 8-3. Core Protection Safety Limits - ANO-1 (Tech Spec Figure 2.1-3) b 2400 l l

~

2200 1 g

~

CT - )

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c.

J 7

~ l

$ 2000 b

3 8

uo 1800 tr 1600 580 600 ,

620 640 660 }

Reactor Outlet Temperature. F l

GPM POWER PUMPS ODERATING (TYPE OF LIMIT)

CURVE 374,880 1005)* 112%

FOUR PUMFA (C,NBR LIMIT) ]

1 i 2 280,035 74.7% 90.8% THREE PUMPS (OVALITY LIMIT) 3 184,441 49.2% 63.7% ONE PUMP IN EACH LOOP (QUALITY LIMIT)

• 106.5% OF DESIGN FLOW

)

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J 8-8 1

1.3 LD4ITDG SAFETY SYSTDi SI?ITDCS, IhrunVE DIS'IFUMDTTATICit Arolicability l A; plies to instruments renitoring reactor power, reactor power irbalance, l reactor coolant systen pressure, reactor coolait cutlet teperature, flew, number of prps in operation, ard high reactor building pressure.

f obicctive 7b provide autcxmtic protection action to prevent any ccrnbimtion of prme l variables fren excoodirn a safety limit.

(

Srveification ,

f 2.3.1 'Iho reactor protection systen trip settirg 1 kits ard the pomissible bypasses for the instrument channels shall be as statcd in Table 2.3-1 ard Figure 2.3-2.

Ihses

[ 'Iho reactor protection system consists of four instru-ent channels to ronitor I cach of several selected plant conditions which will cause a reactor trip if any one of these conditions deviates fren a presolect d operatirg rargo to the degree that a safety limit my be reached.

'Iho trip settirg limits for protection system instrunentation are listed in Table 2.3-1. 'Iho safety amlysis ht.s bcen based upon these protection system instrunentation trip setpoints plus calibration ard instrunentation errors.

Nuclear Cverrower A reactor trip at high pcur level (ncutron flux) is providcd to prevent damgo to the fuel claddiny fren reactivity excursions too rapid to bo detected by pressure ard taperature noasurments.

Durire norm 1 plant operation with all reactor coolant prps cperatirg, reactor trip is initiated khen the tcactor power level reaches 104.9 percent of rated pcur. Addirg to this the possible variation in trip setroints due to calibration ard instrwent errors, the raxirum actual pcar at which a trip would be actuated could be 112%, which is the value uscd in the safety amlysis.

A. Cverpcur Trip tuscd on Flcv ard Irbalance

'Iho pcuer Icvol trip setpoint prtduced by the reactor coolant

( system ficw is insed on a power-to-ficu ratio shich has been establishcd to a m ..eiste the rest severe therm 1 transient

[

considered in the design, the loss-of-coolant-ficw accident frtn l high power. Amlysis has dcronstratcd that the specificd pcuer-to-ficw ratio is adequate to prevent a DIIR of less than 1.30 (IEW-2) or 1.18 (BC) should a Icv ficw cordition exist due to any elcctrical m1furction.

B-9 I _

The power level trip setpaint produccd by the pcuer-to-ficw ratio provides both high power 1cvol ard Icw flos protection in the event the reactor power level increases or the reactor coolsnt ficw rate decreases. The power level trip setpoint prcduced by the pcher-to-ficw ratio pzwides overpcNer DIB prettction for all redes of pep operation. For every ficw rate there is a nuirum pemissible p.wer level, ard for every pcNer level there is a minirum pemissible Icw ficw rate. Typical power icvel. ard Icw ficw rate acrbimtions for the prp situations of Table 2.3-1 are as follcus: )

1. Trip would emm when four rector ocolant prps are cperating if power is 107 percent ard reactar ficw rate is 100 percent )

or ficw rate is 93.5 percent and power level is 100 perant.

2. Trip would occur when three reactor coolant prps are )

cparatirg if FAur is 80 pcIcent ard reactor ficw rate is 74.7 I pcIrent or flcw rate is 70 perrent ani pser level is 75 pertent.

3. Trip sculd emm when one reactor coolant prp is cperatirg in cach locp (total of tua prps cperatirg) if the power is 52 percent ard reactor ficw is 49.2 perocnt or ficw rate is 45.8 perrent ard the Fher level is 49 perewit.

The flux /flcw ratios acrount for the nuirum calibration ard instrunentation errors ard the mxirum variation frcn the average value of the RC ficw sigml in such a mnner that the reactor protective system receives a conservative indication of the PC ficw.

No pemity in reactor ecolant flcw thrcugh the core was taken for an cpen core vent valve tocause of tha core vent valve surveillarre program durirq cach refuelirn cntage. For safety amlysis calculations the mxirum calibratico ard instnrentation errors for the pcuer level were uscd.

1

% pcher-irbalarce boundaries are established in order to pruvent reactor thermi limits from teing exceo$od. These therm 1 limits are either power }

l pr'Airn kw/ft limits or DitR limits. The reactor pcuer irtalarce (pcker in tcp half of core minus p3wer in the bottcn half of cere) Itduces the power icvel trip prtduced by the guer-to-ficw ratio so that the bcurdaries of ]

Fisjare 2.3-2 are prcduced. The pcuer-to-ficw ratio reduces the p3wer 1cvel J trip wmiated reactor pcAur-to-Imctor power irbalan e bcuniaries by 1.07 percent for a 1 pertent flow reduction. )

)

1 B. hrp Manitors In oordunction with the pcuer irbalarre/ficw trip, the prp ]

nonitors prevent the minirum core DER frm docreasing telow 1.30 (EEW-2) or 1.18 (SC) by trippirn the reactor due to the loss of l reactor coolant prp(s) . The prp ronitors also restrict the FAer ]

level for the number of prps in cperation. J 8-10

_ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ]

1 C. RCS Pressure Durirg a startup accident frun low power or a slow red witMrawal l

fran high power, the syste high pressure trip is reached before the nuclear overpower trip setpoint. The trip setting limit shchn in Figure 2.3-1 for high reactor coolant syste pressure (2355 psig) has been established to maintain the syste pressure below the safety limit (2750 psig) for any design transient.(2) l The low pressure (1800 psig) and variable low pressure (11.75 Tout

-5103) trip setpoint shown in Figure 2.3-1 have been established to maintain the DiB ratio greater than or equal to the minimum

( allowable ENB ratio for those design accidents that result in a pressure reduction.(2,3) f Due to the calibration ard instnmentation errors, the safety analysis used a variable low reactor coolant system pressure trip value of (1.1. 75 Tout -5143).

D. Cbolant Cutlet Terperature The high reactor coolant outlet te:Terature trip setting limit (618F) shcAn in Figura 2.3-1 has been established to prevent excessive core coolant t:rperatures in the cperatirq rarge. tue to calibration ard instnrentation errors, the safety analysis used a trip setpoint of 620F.

E. Reactor B.lildirg Pressure The high reactor buildirn pressure trip settirg limit (4 psig) prtnides p sitive assurance that a reactor trip will occur in the l

unlikely event of a steam line failure in the reactor building or a i loss-of-coolant accident, even in the absence of a Irv reactor coolant system pressure trip.

( F. Shutdchn Bjpass In ortler to provide for control red drive tests, zero power physics testiry, ard startup procedures, there is prcuision for bypassirg

( certain segments of the reactor protection systm. The reactor protection syste sognents sMch can be bypassed are shown in Table 2.3-1. 7%o ocniitions are irposed when the bypass is used:

1. A nuclear overpcwer trip a tpoint of 55.0 percent of rated power is autcratically irposed durity reactor shutdcAn.

2. A high reactor coolant systm pressure trip setpoint of 1720 psig is autcratically irposed.

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8-11 f - _ - - - ------ - --

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Figure 8-4. Protective System Maximum Allowable Setpoints -- ANO-1 (Tech Spec Figure 2.3-2)

T,'.ermal Power Level, %FP l

. 140

- 120

(-18.0.107) ^ (18.0.107)

- - 100 l )

ACCEPTABLE g (34.7,90.1)

OPER N l

(-18.0,79.9) -

-8 (18.0,79.9)

(-51.0,74.0) ACCEPTABLE l

l 3 & 4 PUMP l OPERATION (34.7,63.0) l - - 60 l

(-18.0,52.6) l I(18.0,52.61) l

(-51.0,46.9) ACCEPTABLE l l l l 2,3&4 PUMP -- 40 DPERATION l >(34.7,35.7) l

. l l l

(-51.0.19.6) - 20 l l

l l l l l l l 1 l i l i is li li i )

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

)

)

J i

8-12 1

__ _ _ _ _ _ _ _ _ _ . _ _ - _ _ _ _ _ _ _ . - . - - i

6. If a control rod in the regulatirq or axial power shapirq

( groups is declared incperable per Specification 4.7.1.2 operation above 60 percent of the thermal power allavable for the reactor coolant punp oortination may contime provided the rods in the group are positioned such that the rod that was l declared inoperable is contained within allowable group average position limits of Specification 4.7.1.2 ard the withdrawal limits of Specification 3.5.2.5.3.

3.5.2.3 %e worth of sirgle inserted control rods during criticality are limited by the restrictions of Specification 3.1.3.5 and the Centrul Rod Position Limits defined in Specification 3.5.2.5.

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3.5.2.4 Quadrant Tilt:

1. Dccept for physics tests, if quadrant tilt exnaain 4.12%,

reduce power so as not to exceed the allowable power level for the existirg reactor coolant punp ccatimtion less at least 2%

f for each 1% tilt in excess of 4.12%.

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 reduced to less than 4.12% except for physics tests, or the

( followirg adjust:nents in setpoints ard limits shall be mde:

l

a. Se protection system maxirum allowable setpoints (Figure l

2.3-2) shall be reduced 2% in power for each it tilt.

b. We control red group and APSR withdrawal limits shall be reduced 2% in power for each 1% tilt in excess of 4.12%.

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c. Se operational irnbalance limits shall be reduced 2% in power for each 1% tilt in excess of 4.12%.

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3. If quadrant tilt is in excess of 25%, except for @ysics tosts or diagnostic testirn, the reactor will be placed in the hot f shutdown otniition. Diagnostic testing durirg power cperation with a quadrant power tilt is permitted provided the themal power allowable for the reactor coolant purp otrbimt'on is restricted as stated in 3.5.2.4.1 abcNe.

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4. Quadrant tilt shall be monitored on a minirum frequency of cnce everf two hours durirg power cperation above 15% of rated

( power.

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( 8-13 I--. _ - - - - - - -

t 3.5.2.5 Centrol rod positions:

1. 'Nchnical Specification 3.1.3.5 (safety red withdrawal) does not prtMit the exercisirg 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. operatirg red group overlap shall be 20% d5 between two l sequential groups, except for physics tests.

3. Except for physics tests or exercisirg control rods, the '

control rod withdrawal limits are specified on Figures 3.5.2-1(A-C), 3.5.2-2(A-C), and 3.5.2-3 (A-C) for 4, 3 and 2 pump cperation r + tively. If the applicable control red position limits are exemaiad, correct:.ve measures shall be taken imediately to achieve an acceptable control rod position. Acceptable control 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 />.

4. L,coept for physics tests or exercising axial power shapirg rods (APSR's), the followirq limits arply to APSR positiont Up to 410 EFFD, the APSR's may be positioned as m aa 7 for transient imbalance control, however, the APSRs shall be fully i withdrawn by 410 EFED. After 410 EFTD, the APSR's shall not l be reinsertad.

With the APSR's inserted after 419 EFTD, corrective measures l shall be taken imediately to achieve the fully withdrawn positico. Acceptable APSR positions anall 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 Inbalance shall be monitored on a frequency not to exceed 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 3.5.2-4(A-C) . If the imbalance is not within the envelope defined by Figure 3.5.2-4(A-C), corrective measures shall be taken to achieve an acceptable imbalance. If an acceptable inbalance 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 reized until imbalance limits are met.

I 3.5.2.7 'Ibe control rod drive patch panels shall be locked at all times )

with limitad access to be authorized by the Superinterdent.

. )

)

8-14

BRASE

) 'Ihe power-inbalance erwelope defined in Figure 3.5.2-4(A-C) is based on IOcA analyses which have defined the =vl== linear heat rata (see Figure 3.5.2-5), such that the myi== cladding terrporature will not exceed the Final Acceptance critaria, corrwetive naasures will be taken imediately should the indicated quadrant tilt, red position, or imbalance be outside their specified bourdaries. Operation in a situation that would cause the Final Acceptance critaria to be approachei should a IDCA occur is highly improbable because all of the power distribution parameters ( W tilt, rod position, and inbalance) must be at their limits while

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._ . _ [

Figur( 8-5. Rod Position Setpoints for Four-Pump Operation from 0 to 27

+10/-0 EFPO -- ANO-1 Cycle 9 (Tech Spec Figure 3.5.2-1A) 110

~

0PERATION IN THIS KEG 10N IS NOT '

ALLOWED (266*0*90) 90 SHUTDOWN MARGIN 80 -

(248.0.78)

OPERATION E 70 - RESTRICTED

$ l

[ 60 o

50 -

(41.5.48) (212.0.48) ]

c W

E 40

~

l 30 - PERMISSIBLE OPERATING l REGION I 20 (5.5.13) 10 0

O 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 0 20 40 00 8,0 100 Group 7

,0 2p 4,0 60 8,0 Igo Group 6 0 20 40 60 80 100 t i l I i I )

Group 5 Rod Index, *. WD J

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)

8 16

- _ - - - - - - - - - - - - - - - - - - - - - - - - - - - . - -. - --]

f Figure 8-6. Rod Position Setpoints for Four-Pump Operation from 27 +10/-0 to 360 +50/-10 CFPD -- ANO-1 Cycle 9 (Tech Spec Figure 3.5.2-1B)

(

110 (159.5.102) (268.8.102) F (300.102) 100 SHUTDOWN MARGIN LIMIT f

90 - (264.0,90) ,

80 (244.0,78) u OPERATION E 70 RESTRICTED N 60 -

OPERATION IN THIS REGION 15 NOT e ALLOWED j

• 50 -

(198.0.48)

(79.5.48) f40 -

30 PERMISSIBLE 20 -

OPERATING N

10 -

(31.5.13) i (0.6.3) 0 i i i i e i i i i

( i e i i i 0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 O O n I?  ? i O0 n I?O Group 7 0 20 40 60 80 100 l t i i i i i Group 6 q 20 40' 60 80 1p0

{ ' ' ' Rod Index, t WD Group 5 f

I 8-17

( - - - - - - - - - - - - - - - - - - - - - - - _ - - - - - - - - - - - - - - _ - - - _ _

(

Figure 8-7. Rod Position Setpoints for Four-Pump Operation After 360

+50/-10 EFPD -- ANO-1 Cycle 9 (Tech Spec Figure 3.5.2-1C) 110 (159.5.102) (268.8,102);  :(300,102) l 100 - SHUTDOWN MARGIN

'IMII (264.0,90) 90 -

OPERATION 80 -

RESTRICTED (244.0,78)

N 70 - OPERATION IN THIS '

REGION IS NOT ALLOWED 60 -

t 0 (79.5,48)

. (198.0.48) i  ! 40 l

30 .

PERMISSIBLE l OPERATING 20 REGION l

10 - (33.5.13) i (0.6.3) 0 O 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 0 20 40 60 80 100 1 I I i t i Group 7 2,0 4,0 6,0 80 10,0 9

Group 6 0 20 40 60 80 100 t t t i e __j Group 5 Rod Index, ; WD J

8-18

- - - - - - - - - - - - - - - - - - - - . - - - - - - - _ )

I Figure 8-8. Rod Position Setpoints for Three-Pump Operation From 0 to 27

+10/-0 EFPD -- ANO-1 Cycle 9 (Tech Spec Figure 3.5.2-2A) 110 I

100 -

90 -

80 (84.6.77) (271.0. 77) - (300.77))

, OPERATION IN THIS '

E 70 REGION IS NOT

, ALLOWED (266.0.67) e SHUTDOWN '

g 60 -

MARGIN (248.0.58) g LIMIT

~

", OPERATl0N 5- RESTRICTED I 40 -

2 (41.5.36) (212.0.35.5) 30 -

20 - PERMISSIBLE OPERATING 5.5.9.75) REGION 10 0

(0i.75), , , , , , , , , , , , ,

0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 0, 29 49 6,0 8,0 igg Group 7 q 2p 4Q p0 Sp 1Q0 Group 6

,0 20 40 60 8,0 100 Rod Index, t WD Group 5 8- 19

[ - - - - - - - - - - -

\

Figure 8-9. Rod Position Setpoints for Three-Pump Operation From 27 +10/-0 to 360 +50/-10 EFPD -- ANO-1 Cycle 9 (Tech Spec Figure 3.5.2-28) 110 l

100 90 -

SHUTDOWN N

(160.4,77) (269.0,77) F

':(300,77) g 70 -

OPERATION IN THis (264.0,67) e REGION IS NOT ALLOWED 60 - OPERATION 244.0,58) g RESTRICTED

" 50 -

u

! 40 (79.5.36) (199.0.35.5) 30 _

PEPJilSSIBLE 20 -

OPERATING REGION 10 -

(31.5.9.75)

< (0.4.75) i 0 i i e i e i i i i i i i 1 0 20 40 60 80 100 120 140 160 180 200 220 240 260 290 300 0 6p 1

2g 49 80 g 190 Greup 7

,0 20 40 60 8,0 100 Group 6 0 20 40 60 80 100 Rod Index, % WD i i i i i

)

Group 5 i

8- 20

)

_ _____--__-- ---- _ _ _ )

Figure 8-10. Rod Position Setpoints for Three-Pump Operation After 360

+50/-10 EFPD -- ANO-1 Cycle 9 (Tech Spec F19ure 3.5.2-2C) 3 110 100 90 -

f H TDOWN 80 -

N (160.4.77) (269.0.77) ^ t (300,77)

E 70 (264.0.67)

$ OPERATION IN THIS N 60 -

REGION IS NOT (244.0.58) g ALLOWED o 50 - OPERATION

,. RESTRICTED k

40 (79.5.36) (198.0.35.5) 30 20 - PERMISSIBLE s OPERATING 10N 10 - i43.5.9.75)

< (0.4.75) ' ' ' ' '

0 i i ' ' ' ' ' ' '

{

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

,0 30 40 6p 8,0 igg Group ?

p {0 4Q 6p 8p 1g0

( Group 6 p 2,0 j0 6p S,0 igg Rod Index. t WD

{

Group 5

(

8- 21

[ -. - - - - -

Figure 8-11. Rod Position Setpoints for Two-Pump Operation From 0 to 27

+10/-0 EFPD -- ANO-1 Cycle 9 (Tech Spec Figure 3.5.2-3A) <

l l

l l 110 i 100 -

90 -

80 70 -

2 N 60 -

50 15 NO (35.7.52) (271.3.52) - (300.52) c ALLOWED y (266.0.44)

I I R5 CD (248.0.38) fjN 30 - LIMIT (41.5.24) (212.0.23) 20 -

PERM 155!BLE 10 (5.5.6.5) OPERATING REGION s' . 3.1') ' ' ' ' ' ' ' ' ' ' ' '

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

t I i i i i Group 7 f 2f 4,0 6p 8,0 igg )

Group 6 j0 80 f 2,0 4,0 190 Rod Index *. WD Group 5 8- 22

__ - - - - - - - - - - - - - - . - - - -- - .]

Figure 8-12. Rod Position Setpoints for Two-Pump Operation From 27 +10/-0 to 360 +50/-10 EFPO -- ANO-1 Cycle 9 (Tech Spec Figure 3.5.2-38) 110 100 -

90 -

80 k 70 -

5 g 60 -

SHUTDOWN o " (269.3,52) - -

(300.52)

(162.4.52)

• • 50 5- OPERATION IN THIS (264.0.44

! 40 - REG ti NOT OPERATION (244.0.38)

RESTRICTED 30 -

(79 5 24) (198.0.23) 3, _

PERMISSIBLE OPERATING 10 IO,3.11 REGION (31.5,6.5)

O d

i i i i i i i i i i i i i 1 0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 O c, 2,0 4,0 6,0 80 10,0 Group 7 0 2,0 4,0 j0 80 1,00 Group 6 0 20 40 60 80 100 Rod Index t WD -

i i i i i 1

( Group 5

(

(

{

8 23 I .. . - - - - - - - - _ - - - - -

I Figure 8-13. Rod Position Setpoints for Two-Pump Operation Af ter 360 l +50/-10 EFPD -- ANO-1 Cycle 9 (Tech Spec Figure 3.5.2-3C) ,

110 100 -

90 80 -

N 70 -

~

SHUTDOWN

" (269.3.52) - y (300.52) i 50 .

(162.4.52) u g (264.0.441 5- OPERATION IN THIS OPERATION (244.0.38)

REGION IS NOT RESTRICTED 30 - ALLOWED

( 9.5.24) (198.0.23)

PERMISS!BLE

- OPERATING 10

.u.5.6.5) REGION

'3'2 ' ' ' ' ' ' ' ' ' ' ' '

O O 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 0 2,0 4,0 60 80 10,0 Group 7 2,0 4,0 j0 80 10,0 Group 6 0 20 40 60 80 100 Rod Index, t WO t t t t i I .

Group 5 )

)

9 8-24

_ - )

i Figure 8-14. Operational Power Imbalance Setpoints for Operation From 0 to 27 +10/-0 EFPD -- ANO-1 Cycle 9 (Tech,SpecFigure3.5.2-4A)

-- 110

(-14.90.102) . . wa y(13.79.102)

(-19.95.92) -

- 90 (15.34.92)

(-23.78.80) - - 80 (25.19.80)

RESTRICTED y- 70 RESTRICTED REGION  ; REGION

'" 60 5

G  %

(-30.16.50) < g e-- 50 4(26.04.50)

II u

=

E g.

o.

. 40 -

W o -

- 30 M

- 20 E

a.

-- 10 t i e 1 i i i i 1 40 20 -10 0 10 20 30 40 50 Axial Power Imbalance, t 8- 25

l l

Fiqure 8-15. Operational Power Imbalance Setpoints for Operation From

27 +10/-0 to 360 +50/-10 EFPD -- ANO-l' Cycle 9 l (Tech Spec Figure 3.5.2-4B) l l

1

- 110

(-20.21.102) '? -

100 3(21.58.102)

(-20.83,92) j(21.74.92)

- -90

(-28,95,80) - -80 f(23.41.80) l RESTRICTED h- 70 RESTRICTED REGION co REGION g k--60 G t

(-31.06.50) h e . . - -50 4 (26.29.50) 8 C I h-40 a m t

o -

-30 b

O - -20 IS g -

-10 l l 1 I I I I I I i 40 20 -10 0 10 20 30 40 50 Axial Power 1r. balance, %

8- 26 l,

Figure 8-16. Operational Power Imbalance Setpoints for Operation Af ter 360 +50/-10 EFPD -- ANO-1 Cycle 9 (Tech Spec Figure 3.5.2-4C) 110

( 20.21.102) ,, ,vg (21.58.102)

(-25,62,92) ..

93 (25.03,92) l l

(-28,95,80) --

80 (27.54,80) l

-- 0 RESTRICTED RESTRICTED l

REGION $ REGION l = g.. 60

. b  %

i (-31.06,50)b " "-- 50 (31.20,50)

E t E3 5' 40

& e-kf -

o .-

30 b

m 5.

20 E  !

l gl -- 10 i i 1 I i 1 i t i 1 40 20 -10 0 10 20 30 40 50 Axial Power Irbalance.

t i

I l

8-27 l l

l i

Figure 8-17. LOCA Limited Maximum Allowable Linear Heat Rate - ,!

ANO Cycle 9 (Tech Spec Figure 3.5.2-5)  !

P

?

(

L 20 , , ,~

i i i r- i i  !

- f f

, 18 _

. l t  !

3 i c' .1 e

• .==""*,,,==/ I t

~

/ k E

• ~

/ i l

o / '

?

1 I o 14 - / -

E l

• 3

\

E 12 -

f 0 - 1000 mwd /mU  !

t

. AFTER 1000 mwd /mtU '

10 ' 1 e i 1 1 i i i i I

(

0 2 4 6 3 10 12 I Axial Location From Bottom of Core. f t. -

i

)

l

[

l t

a-28  !

_ - _ -I

5.3 REACfIOR Soccification 5.3.1 Bry: tor Core 5.3.1.1 %e reactor core contains approximately 93 metric tons of slightly enriched uranium dioxide pellets. We pellets are encapsulated in Zircaloy-4 tubing to form fuel rods. The reactor core is made up of 177 fuel assemblies. Each fuel assernbly is fabricated with 208 fuel rods. (1,2) Startirg with Batch 11. a reconstitutable fuel assernbly design is inplemented. Thi:s design allows the replacement of up to 208 fuel rods in the asserrbly.

5.3.1.2 The reactor core approximates a right circular cylinder with an equivalent diamp.ter of 128.9 inches ard a height of 144 inches.

The active fuel lergth is approximately 142 inches. (2) 5.3.1.3 The average enrichment of the initial core is a ntninal 2.62 weight percent of 235U Wree fuel enridirents are used in the initial corn. The highest enrichment is less than 3.5 weight percent 235U 5.3.1.4 There are 60 full-length control rod assemblies (CRA) ard 8 axial power shapirs rod assemblies (APSRA) distributed in the reactor core as shown in FSAR Figure 3-60. Each full-length CRA contains a 134-inch length of silver-indium-cadmium alloy c'ad with stainless steel. E Ach APSRA contains a 63-irch length of Inconel-600 alley clad with stainless steel.(3) 5.3.1.5 %e initial core has 68 burnable poison spider asscrblics with

< similar dimensions as the full-length control rods. The cladding is Zircalcy-4 filled with alumina-boron and placed in the core as shown in FSAR Figure 3-2.

5.3.1.6 Reload fuel asserrblies ard rcds shall conform to the design and evaluation described in FSAR and shall not exceed an enrichment of 1.5 percent of 235g, 5.3.2 Reactor Coolant Systs f 5.3.2.2 The reactor coolant system is designed and constructed in accordance with code requirements. (4) 5.3.2.2 The reactor coolant system and any connected auxiliary systens exposed to the reactor coolant conditions of tertperature ard pressure, are designed for a pressure of 2500 psig and a terrperature of 650 F. The pressurizer and pressurizer surge line are designed for a terrporature of 670 F.(5) 5.3.2.3 The reactor coolant systm volume is less than 12,200 cubic feet.

8-29

{ - _ _ - - - -

(

k

(

(

9. STARIUP FROGRAM - HIYSICS TEIDG The planned startup test program associated with core perfomance is cutlined belcw. These tests verify that core performance is within the asstmptions of the safety analysis and provide information for continued safe cperation of the unit.

9.1. Precritical Tests 9.1.1. Control Red Trio Test Procritical control rod drop times are recorded for all control rods at hot full-ficu conditions before zero por physics testinJ begins. Acceptance criteria state that the rod drop time from 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 red drop from fully withirawn to two-thinds inserted. Since the rest accurate position indication .s obtained from the zone reference switch at the 7'5%-inserted position, this position is used instead of the two-thirds inserted position for data gathering.

9.1.2. RC Flw Reactor coolant flcu with four RC punps runnity will be reasured at hot shutdcun cerviitions. Acceptance criteria require that the measured f1cu be within allcuable limits.

9.2. Zero Power Ihysics Tests 9.2.1. Critical Boron ConcentratigD Once initial criticality is achieved, equilibrium boron is obtaincd ard tN critical boron concentration detemined. The critical boron concentration in calculated by correctiry for any red withdrawal required to achicvo equilibrium boron. The acceptarm criterjon placed on critical boron concentration is that the actual boron concentration mist be within 100 pyr. boron of the predictcd value.

(

I 9-1

9.2.2. Temerature Reactivity Coefficient

%e isothermal HZP tenperature coefficient is measured at approximately the all-rods-cut configuration. Durirg changes in terperature, reactivity fmTMck may be ccupensated by control rod movement. The charge in reactivity is then calculatiad by the summation of reactivity (obtained frun a reactivity calculator strip chart recorder) associated with the ter:perature change. Acceptance criteria state that the measured value shall  ;

not differ frtn the predicted value by nere than 0.4x10-4 aM.

The nederator coefficient of reactivity is calculatcd in conjunction with the l tenperature coefficient neasurement. After the tenperature coefficient has been measured, a predicted value of the fuel Dcipler coefficient of reactivity is ad: led to obtain the rederator coefficient. This value nust not be in excess of the acceptarce criteria limit of +0.5x10-4 AM.

9.2.3. Control Rcd Groun/ Boron Reactivity Worth Control rod group reactivity worths (groupo 5, 6, and 7) are mez.sured at hot zero power conditions usirg the boron / red swap method. This technique consists of establishirq a deboration rate in the reactor coolant system and cxxpensatirg for the reactivity charges frem this deboration by inserting control red groups 7, 6, ard 5 in arremental steps. The reactivity changes that occur durtry these measurenents are calculated based on reactimeter data, and differential rod worths are obtained frcn the measured reactivity worth versus the cM nge in Itd group position. The differential Itd worthe of each of the controllirg grcupa are then sumed to obtain integral rod group worths. The acceptance criteria for the control bank (,o:up worths are .

as follchs:

1.Irdividual bank 5, 6, 7 worth:

Dredictcd value - reasurcd value x 100 < 15 measured value -

2. Sums of ea ups 5, 6, ard 7:

uredictcd value - reasurcd value x 100 $ 10 gg The boron reactivity worth (difforential boron worth) is measured by dividing the total inserted rod worth by the boren charge mde for the red korth test.

The acceptance criterion for reasurcd differential boren worth is as follows:

9-2

1. credicted value - measured value x 100 $ 15 M ue to predicted rul worths and differential boron worth are taken frun the PIM.

9.3. Power Escalation Tests 9.3.1. Cbre Syrmtry Test me purpose of this test is to evaluate the symetry of the core at lw power durim the initial power escalation following a refuelity. Symetry evaluation is based on incore quadrant power tilts during escalation to the intermcrilate pcuer level. We core symetry is acceptable if the absolute l values of the quadrant power tilts are less than the error adjusted " sru limit.

l 9.3.2. Core Power Distribution Verification at Interry:xiiate Power IcVel I (IPL) aM 100% FP With Nomimi Control Rcd Position

! Core power distribution tests are performod at the IPL and 100% full power (FP). Equilibrium xenon is establishcd prior to both the IPL ard 100% FP l tests. The test at the IPL is essentially a check on power distribution in the core to identify any abnorralities before escalatirg to the 100% FP 1

l plateau. Peakirq factor criteria are applied to the IPL mre power distribution results to determine if additior.al tests or amlyses are required prior to 100% FP cperation.

We followirg acceptance criteria are placed on the IPL and 100% FP tests:

1. The worst-caso maxinum um nust bo less than the IOCA limit.

2. The minirum DE nust be greater than the initial cxxviition DE limit.

3. The value obtained from extrapolation of the mininum D E to the next pcuer plateau overpower trip setpoint nust be greater than the initial cordition DE limit or the extrapolated value of irbalanco nust fall outside the RPS power /imtalance/ flow trip envelopo.

4. The value cbtainori frun extrapolation of the worst-caso maxinum Um to the next pcuer plateau overpower trip setpoint nust be less than the fuel melt limit, or the extrapolated value of irbalance nust fall L outsido the RPS pcAur/irbalance/ flow trip envelopo.

5. The quadrant power tilt shall not exceed the limits specificd in the f 7bchnical Specifications.

{ 9-3

6. 'Ihe highest Incar' red ard predictcd radial peaks shall bu within the followirg limits:

credicted value - measurtd value x 100 rnore positive than -5 Ineasured value

7. 'Ihe highest incasured ard pralictai total peaks shall be within the followiry linits:

ortdicted valc., - reasured valu x 100 more positive than -7.5 Incasured value Items 1, 2 and 5 ensure that the safety limits are maintaincr'. at the IPL and 100% FP.

Items 3 aM 4 establish the criteria whereby escalation to full power may be acomplishcd without the potential for exceeding the safety limits at the overpower trip setpoint with regard to I2ER ard linear heat rate.

Items 6 airl 7 are establishcd to detemine if reasured ard predicted core power distributions are consistent.

9.3.3. Incore Vs. Dccore Detector Irbalance Correlation Verification at the IPL Irrbalances, set up in the core by control rui positionity, are read simultanecusly on the incore detectors and excore pcser rarge detectors. 'Ihe excore de' actor offset versus incore detector offset slope nust be greater thar 0.96. If this criterion is not met, gain arplifiers on the excore detector signal procccsirq equignent are adjusted to provide the required gain.

9.3.4. '1Wrerature Reactivity Coefficient at s100% FP

'Ihe average reactor ccolant tcrporature is decreascd ard then increascd by ahcut SO F at constant reactor power. 'Ihe reactivity associatcd with each

torporature charge is obtained frun the charge in the controllirq rod group l position. Controlling rod g

cup worth is reasurtd by the fast insert / withdraw rnothed. 'Ibe terporature reactivity coefficient is calculatcd frun the Incasured changes in reactivity arri teqcrature.

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

( 9-4

9.3.5. Pcuer Droler Reactivity Cbefficient at N100% FP

[ 2e pcuer Doppler reactivity coefficient is calculated frcra data recorded r

durirg control rod worth measurements at ocwer usirq the fast insert /withiraw method.

W e fuel Deppler reactivity coefficient is calculated in conju d. ion with the pcwor Doppler coefficient measurement. Se pcuer Doppler coefficient as measured above is multipliM by a precalculated conversion factor to obtain f

the fuel Dcppler coefficient. 'Ihis measured fuel Ibppler coefficient rust be tore negative than the acaptance criteria limit of -0.90 x 10-5 ag, 9.4. Procedure for Use if Acceptance Criteria Not Met If acceptance criteria for any rest are not met, an evaluation is performM before the test program is continucd. Further specific actions depend on evaluation results. Resa actions can ircitxle repeating the tests with rore detailed attention to test prerequisites, aMcd tests to search for ancrnlies, or design personnel perfomirg detailed azalyses of potential safety pr@ lems because of parameter deviation. Power is nce escalated until evaluation shows that plant safety will not be ccupmnised by such escalation.

I I

E

10. N1.t L.KuiCES

1. Arkansas INelear One. Unit 1 - Final Safety Analysis ReDort, Docket 50-313, Arkansas Power & Light.

2. BEFA Rotainer Design Report, BAW-1496, B,Wk & Wilcox, Lyndturg, Virginia, May 1978.

3. J. H. Taylor (B&W) to S. A. Varga (IRC), Intter, "BPRA Retainer Reinsertion," January 14, 1980.

4. Rancho Seco Cycle 7 Reload Report - Volume 1 - }hrk BZ Fuel Assmbly Design Reprt, BAW-1781P, Babcock & Wilcox, Lynchburg, Virginia, April 1983.

5. Rancho Seco Nuclear Generating Station - Evaluation of !brk BZ Fuel Asserbly Design, U.S. Nuclear Rcquiatory Camission, Washington, D.C. ,

November 16, 1984.

6. Gadolinia-Bearing Lead Test Assemblics Design Report, BAW-1772-P, f

i B'hk & Wilcox, Lynchburg, Vhginia, June 1983.

7. Program to Detemino In-Reactor Performance of B&W Fuels - Cladding l Creep Collapco, PAW-10084P-A. Rev. 2, Babccck & Wilcox, Lyndturg, l

Virginia, October 1978.

8. Y. H. Hsii, et al. , TACO 2-Fuel Pin Performance Analysis, ImW-10141P-A, l

Pcv.1, Babcock & Wilcox, Lynchturg, Virginia, June 1983.

\

9. !DODIE - A Malti-Dimensional 'Iko-Group Reactor Sirulator, IWW-101.52A, Babcock & Wilcox, Lynds%.1Ig, Virginia, June 1985.

10. Arkansas IMclear One Unit 1, Cycle 8 Reload Report, PAW-1918, Ihbcock &

Wilcox, IJynChburg, Virginia, NOVCKber 1986.

11. J. H. Jcnes, et al. , LYNXP - Coro Transient 'Iherml-lfydraulic Program, IWW-10156A, TbWk & Wilcox, Lynchburg, Virginia, February 1986.

( 12. R. L. Ihrno and J. H. Jones, 'Ihcrml-lfydraulic Crocaficu Applications, IWW-1829, Babcock & Wilcox, Lynchburg, Virginia, April 1984.

10-1

[ _ --- - - - - _ - - -- -- --- _ - --_--- _

13. BC Correlation of Critical Heat Flux, BAW-10143P-A, R'Wk & Wilcox, Lynchburg, Virginia, April 1985.

14. Fuel Rod Bcuing in B'Wk & Wilcox Riel Designs, BAW-10147P-A Rev.1, Babcock & Wilcox, Lynchburg, Virginia, May 1983.

15. Arkansas Nuclear One, Unit 1 - Fuel Densification Report, BAW-1391, B'Wk & Wilcox, Lynchb.irg, Virginia, June 1973.

16. ECCS Analysis of B&W's 177-FA lowertd-Icop NSS, BAW-10103A. Rev. 3, Babcock & Wilcox, Lynctiburg, Virginia, July 1977.

17. B&W ECCS Evaluation Model Revision 5, BAW-10104. Rev. 5, Babcock &

Wilcox, Lynchburg, Virginia, February 1985.

18. Boundirg Analytical Assessment of NUREG-0630 Mcdels on I.OCA kW/ft Limits With Use of FIECSET, PAW-1915P, Mbcock & Wilcox, Lynchburg, Virginia, May 1986,

19. Norral Operating Controls, BAW-10122A. Rev. 1, B'W_k & Wilmx, Lynchburg, Virginia, my 1984.

l 10-2

_ _ _ _ _ _ _ _ _ _ _