ML20207J036

Cycle 9 Reload Rept

ML20207J036
Person / Time
Site:	Arkansas Nuclear
Issue date:	06/30/1988
From:	BABCOCK & WILCOX CO.
To:
Shared Package
ML20151H726	List: 1CAN078803, Application for Amend to License DPR-51,changing Tech Specs Re Cycle-9 Operational Requirements.Cycle 9 Reload Analysis Rept Included for Technical Support.Fee Paid ML20207B179 ML20207J036
References
BAW-2027, NUDOCS 8808300189
Download: ML20207J036 (75)

Text

BAW-2027 June 1988 ARKANSAS NUCLEAR ONE, UNIT 1

- Cycle 9 Reload Report -

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BAK-2027 June 1988 1

ARKANSAS NUCEAR CNE, UNIT 1

- Cycle 9 Reload Report -

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1 BABCDCK & WIICOX Nuclear Power Division P. O. Box 10935 Lynchburg, Virginia 24506-0935 l

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CDUITNIS Page

1. INIPOCUCI' ION AND SLM4ARY . . . . . . . . . . ........... 1-1

2. OPERATING HISICRY .... ................ .... 2-1

3. GDIERAL IESCRIPTION .................. ..... 3-1

4. FUEL SYSTDi DESIGN . ...... ................. 4-1 4.1. Fuel Assembly Hechanical Design . ............ 4-1 4 . ?. . Fuel Rod And Gray APSR Design ............... 4-2 4.2.1. ClaMinJ Collapse ................. 4-2 4.2.2. Clackling Stress . ................. 4-3 4.2.3. Cladding Strain . ................. 4-3 4.3. 'Ihermal Der ign . . . . . . . . . . . . . . . . . . . . . . . 4-4 4.4. Material Design ........... ........... 4-4 l 4.5. Operatirg Experience . . . . . . . ............. 4-4

5. NUC1 EAR DESIGN . . .................. ...... 5-1 5.1. Ihysics Characteristics . ................. 5-1 5.2. Analytical Input . ......... ............ 5-1 5.3. Changes In Nuclear Design. . . . . . . . . . . . . . . . . 5-1

6. 'IHERMAIrHYEFAUIlC DESIGN . . . . ................. 6-1 l I

7. ACCIDENT AND 'IRANSIDir ANAIXSIS ................. 7-1 7.1. General Safety Analysis . ... ............. 7-1  ;

7.2. Accident Evaluation ........ ............ 7-2 J

8. IPOIMED }ODIFICATICtB 'IO TECWICAL SPECIFICATICNS . ....... 8-1 l

9. STARIUP IPCG%M - IHYSICS TESTDC ................ 9-1 )

9.1. Precritical 'Ibsts .......... ........... 9-1 9.1.1. Ocotrol Rod Trip Test ............... 9-1 9.1.2. RC Flow ........... ........... 9-1 9.2. Zero Power Ehysics 'Ibsts . . . . . . . . . . . . . . . . . . 9-1 9.2.1. Critical Boron Cbncentration . . .......... 9-1 1 9.2.2. Temperature Reactivity Coefficient . . . . . . . . . 9-2 9.2.3. Control Rod Group / Boron Reactivity Worth . . . . . . 9-2 9.3. Power Escalation Tests . . . . . . . . . . . . . ...... 9-3 9.3.1. Core Symetry Test . . . . . . . . . . . . . . . . . 9-3 1

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Cbntents (Cont'd)

Page 9.3.2. Core Power DistribItion Verification at Intenwdiate Power Invel (IPL) and 100% FP With Ncninal Control Rod Position. . . . . . . . . . 9-3 9.3.3. Incore vs Excore Detector Imbalanm -

Correlation Verification at the IPL. ........ 9-4 '

9.3.4. 'Ibnperature Reactivity Coefficient at s100% FP . . . 9-4 9.3.5. Power Ibppler Reactivity Coefficient at %100% PP . . 9-5 9.4. Procedure for Use if Acceptance Criteria Not Met . . . .. . . 9-5 ..

10. REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . 10-1 List of Tables Table 1

4-1. Fuel Design Parameters ard Dimensions ............. 4-5 4-2. Fuel 'Ihermal Analysis Parameters . . . . . . . . . . . . . . . . 4-6 [

4-3. Operatirg Experience . . .................... 4-7 5-1. Physics Parameters for N O-1, Cycles 8 and 9 . . . . ...... 5-3 5-2. Shutdcwn Mitgin Calculations for NO-1, Cycle 9 ........ 5-5 6-1. Ihximum Design Coniitions, Cycles 8 ard 9 ........... 6-3 -

7-1. Ccrparison of Cycle 8 and Cycle 9 Accident Doses . . . . . . . . 7-4 7-2. Ctreparisen of Key Parameters for I.ccident Analysis . . . . . . . 7-5 l

7-3. Bouniing Values for Allcwable IOCA Peak Linear Heat Rates ... 7-5 .

List of Flaures Figure ,

3-1. Core Ioading Diagram for N O-1, Cycle 9 ............ 3-3 3-2. Enrichment ard Burnup Distrilm'tlon, NO-1 Cycle 9 Off '

440 EFED Cycle 8 . . . . .................... 3-4 '

3-3. Control Rcd Incations and Group Designations for N O-1, Cycle 9 . . . . . .................... 3-5 3-4. IBP Enrichnent ard Distribution, NO-1, Cycle 9 ........ 3-6 4-1. Renovable Upper End Fitting (Side View) . . . . . . . . . . . . . 4-8 4-2. Holddcun Sprirg Retainer . . . . . . . . . . . . . . . . . . . . 4-9 4-3. BFPA Spider. . . . . . . . . . . . . . . . . . . . . . . . . . . 4-10 "

4-4. BIPA Spider / Upper Erd Fitting / Reactor Intenuls Interaction. . . 4-11 4-5. Gray Axial Power Shapirg Rod . . . . . . . . . . . . . . . . . . 4-12 5-1. NO-1, Cycle 9, B3C (4 EFFD) Two-Dimensional Relative Power Distribution - Rill Pcuer Equilibrium Xenon, Normal Rod Positions . ... . .................... 5-6 8-1. Core Protectinn Safety Limit - NO-1, ('Ibch Spec '

Figure 2.1-1). . . . . . . . . . . . . . . . . . . . ...... 8-6 8-2. Core Protection Safety Limits - NO-1, ('Ibch Spec Figure 2.1-2). . . . . . . . . . . . . . . . . . . . . . . . . . 8-7 iv

List of Fiaures (Cont'd)

Figure 8-3. Core Protection Safety Limits - MD-1, (Tech Spec Figure 2.1-3). . . . . . . . . . . . . . . . . . . . . . . . . . 8-8 8-4. Protective System Maxinum Allowable Setpoints

- MD-1, (7bch Spec Figure 2.3-2) . . . . . . . . . . . . . . . 8-12 r

8-5. Rcd Position Setpoints for Four-PLmp Operation fran 0 to 27 +10/-0 EFPD - MO-1, Cycle 9 (Tech Spec Figure 3.5.2-1A). . . . . . . . . . . . . . . . . . . 8-16 8-6. Rod Position Setpoints for Four-Ptmp Operation frat 27 +10/-0 to 360 +50/-10 EFED - NO-1, Cycle 9 (Tech Spec Figure 3.5.2-1B . .............. 8-17 8-7. Rod Position Setpoints for Four-Ptmp Operation after 360 +50/-10 EFPU - NO-1, Cycle 9 (Tech Spec Figure 3.5.2-1C). . . . . . . . . . . . . . . . . . . 8-18 8-8. Rod Position Setpoints for Three-Ptmp Operation Frcan 0 to 27 +10/-0 EFPD - MD-1, Cycle 9

('Dech Spec Figure 3. 5. 2-2A) . . . . . . . . . . . . . . . . . . . 8-19 8-9. Rod Position Setpoints for Ihree-PLmp Operation Fran 27 +10/-0 to 360 +50/-10 EFPD - MD-1, Cycle 9 (Tech Spec Figure 3.5.2-2B). . . . . . . . . . . . . . . 8-20 8-10. Rod Position Setpoints for Three-PLmp Operation After 360 +50/-10 EFPD - MD-1, Cycle 9 (Tech Spec Figure 3.5.2-2C). . . . . . . . . . . . . . . . . . . 8-21 8-11. Rod Position Setpoints for Iko-Ptmp Operation Fran 0 to 27 +10/-0 EFID - MD-1, Cycle 9 (Tecti Spec Figure 3. 5. 2-3A) . . . . . . . . . . . . . . . . . . . 8-22 8-12. Rod Position Setpoints for 7%o-Ptmp Operation ,

Fran 27 +10/-0 to 360 +50/-10 EFPD - NO-1, Cycle 9 (7bcti Spec Figure 3.5.2-3B . .............. 8-23 8-13. Rod Position Setpoints for Two-Ptmp Operation After 360 +50/-10 EFPD - MD-1, Cycle 9 (Tech Spec Figure 3.5.2-3C). . . . . . . . . . . . . . . . . . . 8-24 8-14. Operational Powr Imbalarce Setpoints for Operation Fran 0 to 27 +10/-0 EFFD - MD-1, Cycle 9 (7bch Spec Figure 3. 5. 2-4A) . . . . . . . . . . . . . . . . . . . 8-25 8-15. Operational Power Inbalance Setpoints for Operation Fran 27 +10/-0 to 360 +50/-10 EFPD - MD-1, Cycle 9 (7bch Spec Figurt 3.5.2-4B) . . . . . . . . . . . . . . . 8-26 8-16. Operational Power Imbalance Setpoints for Operation After 360 +50/-10 EFPD - NO-1, Cycle 9 (7bch Spec Figure 3. 5.2-4C) . . . . . . . . . . . . . . . . . . . 8-27 8-17. IDCA Limited Maxinun Allcuable Linear Heat Rate MD Cycle 9 (Tech Sp'c Figure 3.5.2-5) . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-28 v

1. INIROCUCTION AND SUtt%RY

'Ihis report justifies the operation of the ninth cycle of Arkansas Nuclear One, Unit 1 (ANO-1) at the rated core power of 2568 IWt. Included are the required analyses as outlined in the UStmC h ==nt, "Guidance for Proposed License Amendments Relating to Refuelirg," June 1975.

'Ib support cycle 9 operation of AID-1, this report euploys analytical  ;

techniques and design Mw established in reports that have been sukunitted to and accepted by the USNRC ard its prher, the USAEC (see references) .

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'Ihe cycle 8 and 9 reactor paramete m related to power capability are '

sumarized briefly in section 5 of this report. All of the accidents analyzed in the ISAR1 have been reviewed for cycle 9 operation. In those l cases where cycle 9 characteristics were conservative ocmpared to those aralyzed for previous cycles, new ancident analyses were not performed.

'Ihe Technical Specifications have been reviewed, and the modifications required for cycle 9 operation are justified in this report.

Based on the analyses performed, whictn take into account the postulated effects of fuel densification and the Final Acceptance Criteria for Emergency Core coolirq Systes, it has been concluded that Alo-I can be operated safely for cycle 9 at a rated power level of 2568 IWt.

1-1

2. OPERATDU HISTOP's

'Ihe reference cycle for the nuclear and themal-hydraulic amlyses of Arkansas Nuclear One, Unit 1 is the currently operatirg cycle 8. 'Ihis cycle 9 design is based on a design cycle 8 length of 440 effective full Pwer days (EFPD) .

'Ihe plant was operated at 100% full pwer for the first 2.5 renths of cycle

8. Power was then rcduced in order to avoid a sunmer 1988 refueling. 'Ihe plant was operated at 65% for 2.5 renths, 80% during two sumner raonths and 70% for 2.5 renths.

Follwirg a one renth mintemnce outage the plant was restarted to 80% full pwer. Continued operation at 80% is planned for the reminder of cycle 8.

No ancmlies occurred durire cycle 8 that would advarsely affcct fuel performnce during cycle 9.

2-1

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3. GDIERAL DESCRIPT1Cti 2e MD-1 reactor core is described in detail in section 3 of the Arkansas Nuclear One, Unit 1, Final Safety Analysis Report (PSAR) .1 he cycle 9 core contains 177 fuel assemblies, each of which is a 15 by 15 -

array contr.inirg 208 fuel rods, 16 control rod guide tubes, and one incore instrument guide tube. he fuel is cmprised of dished-end, cylindrical pellets of uranium dioxide clad in cold +crked Zircaloy-4.

We fuel assemblies in all batches have an average ncminal fuel loading of 463.6 kg of uranium. We undensified ncminal active fuel lengths, theore-tical densities, fuel and fuel rod dimensions, and other related fuel pea-meters are given in Tables 4-1 ard 4-2 for all fuel assemblies.

Figure 3-1 is the fuel shuffle diagram for NO-1, cycle 9. he initial

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enrichrnts of batches 6D, 9B,10 and 11 are 3.19, 3.30, 3.35, and 3.45 wt %

U-235, respctively. One batch 7D assembly, all of batch 8B, and 16 of the twice-turned batch 9 assemblies will be discharged at the end of cycle 8.

We center lccation will contain a batch 6 assembly discharged at the end of cycle 5 (designated 6D) . ho remainirg 52 twice-burned batch 9 assemblies (designatcd 9B) will be shuffled to new locations, with 16 on the core periphery. We 64 once-burned batch 10 assemblics will be shuffled to new locations, and the 60 fresh batch 11 assemblies will be loaded in a symmetric checkerboard pattern thrtughout the core. Figure 3-2 is an eighth-core mp showing the assa ly burnup a'd enrichnent distribution at the beginnirg of cycle 9.

Reactivity is controlled by 60 full-lergth Ag-In--Cd control rods, 52 turnable poison rod assemblics (BmAs), ard soluble boron shim. In addition to the full-lergth control rods, eight Irronel axial power shapiry rods (gray APSRs) are provided for additioral control of the axial pcuer distribution. Se cycle 9 locations of the 68 control rcds ard the group desigmtions are indicated in Figure 3-3. We core locations of the total pattern (68 control rods) for cycle 9 are the same as those of the referan cycle but the group l

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designations are different. '1he locations and enridunants of the BPRAs are shown in Figure 3-4.

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u 6naa E-t- 3oaa loeptu6 0Fe6aeut Joa yno-t 34ota 6 i

1 fO 10 68 10 10 y 10( )02 009 )tf 11(

69 !O 10 11 66 11 tO t0 69 9 305 109 NO( J 706 A 41( 110 311 69 11 fO 11 10 II 10 11 10 11 66 3 H0z J wt J net J iiz J niz J ees 6E 11 tc It 6S 11 68  !! 68  !! 10 II 69 l

0 )0( J 3 01 4 Y09 4 409 J Y!O J 311 4 31(

3 10 !O I. 61 11 61 !O 68  !! 68 II fo 10 s06 O!! J' 909 J SOS 009 811 J ' itt i 00$ J04 10 !O 11 eS 11 66 tI 10 11 60 11 6E II 10 10 4 310 312 J JCI i s06 J s09 J e'O! J dtS i 30t 309 tC 11 10 11 68 11 69 10 68  !! 68 11 10  !! 10 0 806 4 006 4 302 J 30t %8 31( 4 319 J 004 s 904 65 63 69 10 10 !O 90 10 10 10 61  !! 69 69 "" A eH H04 001 4'l 1!t dit lit w!! 106 40S J0S HC( J02 4 11$ H06

n,;n 3

!O  !! tO  !! 68  !! 68 10 69 11 68 11 10  !! 10 406 J 406 4 n02 J 00( 308 01( J wtt 4 NC( J d04 10 iO  !! 68  !! 69 11 10 11 69 11 68 11 10 10 1 0!O O!! J 101 i CIS J 310 4 Y04 4 115 4 00t 009

!O 10  !! 66 11 68 10 60 11 68  !! tO 10 w

106 htt J 102 J d0$ 309 dlI 4 410 J NOS 101 h 68  !! 10 11 69  !! 68 11 68 11 !O 11 68 wC( J 105 4 s09 4 ttC J VIC J wtO J WI(

68  !! 10  !! tO II 10 11 fO 11 69 0 406 s 30t J 00t 4 012 J 312 J MIV d 61 10 IC 11 66 11 10 10 68 001 009 COC i s04 J 01( 010 011 IC 10 69 !O 10 y

40t OOi 108 0tt Jt(

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9 9 9 4 8 6 lO ll tE L( tv t$

L 3 C gt234 4c141 J 04u0145 iJ314 464L daepoms ]osa losettow E-E

Figure 3-2. Enrichment and Burnup Distribution.

ANO-1 Cycle 9 off a40 EFPD Cycle 8 l

8 9 10 11 12 13 la 15 1 3.19 3.35 3.35 3.35 3.30 3.45 3.30 3.30 H

20770 18734 18664 18158 28714 0 20852 31040 3.30 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

20863 0 23038 0 13625 17387 3.30 3.45 3.35 3.35 M

28713 0 18133 18684 3.35 3.45 3.30 N

18625 0 31942

=

3.30 0

30940 l

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x.xx Initial Enrichment, wt % U-235 xxxxx BOC Burnup ,

mwd /mtU j i

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

Fuel Transfer l

Canal )

A B 1 6 1 C 3 5 5 3 D 7 8 7 8 7 E 3 5 4 4 5 3 F 1 8 6 2 6 8 1 5 4 2 2 ,4 5 G

H W- 6 7 2 2 7 6 -Y K 5 4 2 2 4 5 L 1 8 6 2 8 1 l6 M , 3 5 4 4 5 3 N l 7 8 7 '8 7 0 l 3 5 5 3 P l l 1 6 1 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 l 1 8 Safaty 2 8 Safety i 3 8 Safety i 4 8 Safety l 5 12 Control i 6 8 Control  !

j 7 8 Control i i 8 8 APSRs

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Figure 3-4. LBP Enrichment and Distribution, ANO-1 Cycle 9 8 9 10 11 12 13 14 15 H 1.355 K 1.355 1.400 0.200 L 1.355 1.400 0.800 M 1.400 1.400 N 1.400 1.400 0 1.355 0.800 P 0.200 R

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x . x x y. LBP Concentration, wt % B4 C in Al.0 3 l

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4. EUEL SYSTD4 DESIGN 4.1 Fuel Asserbly Mechanical Desicrn The types of fuel assemblies and pertinent fuel design parameters for NO-1, cycle 9 are listed in Table 4-1. All fuel wemblles aIV identical in concept and are mechanically interchargeable. Retainer assenblics will be used on the two fuel assemblies containing the regenerative neutron sources (MG) . The justification for the design and use of the retairdts described in references 2 and 3 is applicable to the MG retainers in cycle 9 of NO-1.

The batch 11 fuel uses Zircaloy rather than Inoonel as the material for the intermediate spacer grids as reported in reference 4. The NRC safety evaluations of that report requires that a licensee who is incorporating that design suknit a plant-specific analysis of ccobirrad seismic ard IDCA loads according to Appendix A of the Stardard Review Plan 4.2. The analpis that was presented in reference 4 envelopes the MD-1 plant design requirenents.

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

Batch 11 utilizes the MK-B6 type of ambly. The differences between this assembly and other MK-B types are the method used to retain fixed control ccuponents (BPPAs, orifice red assemblies, and regenerative neutron source ccoponents) during reactor operation, Zircaloy spacer grids, and the fact that it is reconstitutable. The removable upper eM fittirg (Figure 4-1) provides four open slots that align ard allow designed movement of the holddown spring and its rutainer (Figure 4-2) . The fixed control ccrepent spider is shown in Figure 4-3. The holddown spring is pru-loaded through a i stcp pin welded to an ear on each side of the upper erd fitting. Incore, as shown in Figure 4-4, the spider feet are captured between the holddown spring retainer ard the upper grid pads on the reactor internals. This arrargement retains the fixed control cxmtenents at all desi 7 flow conditions. The l

removable upper erd fitting is identical to the Mark B5 upper eM fitting except for the way it is attached to the control red guide t,M. The Mark B5 upper end fitting 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 similated reactor environment, to determine analytical input and to assure goed incore performance.

The removable upper end-fitting of the reconstitutable fuel aM1y is a I direct desce.xlent of the Gadolinia Isad Test Assenbly ;* 9) upper erd fitting. 7he end fittiny design was thoroughly analyzed and tested. These results were subnitted to the NRC in reference 6. The five Cadolinia LTA'a 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 acitieved a burnup of approximately 53000 MM/mtU. By the end of this cycle it will have reached a burnup of 59000 Mwd /nfl\1.

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1he ability to reconstitute the fuel acembly has no detrimental effect on the as=mbly in-core performance. This allows selective replacement of damaged fuel rods within an assembly, stlich has a tr verricus cost-saviny potential.

4.2 Fbel Rcd ard Gray APSR EesigD The rechanical evaluation of the MK-B fuel rods arx1 the gray APSR's is di m W belcw.

4.2.1 Claddina Collarse A. Ebel Red Creep collapse analyses were performcd for the four different fuel batch pcver histories. Because of its longer previtx2s incere exposure time, the batch 9B fuel is nore limiting than the other batches. The batcis 9B ascambly power history was analyzed and the most limiting acembly was determined. (

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The power history for the most limiting assemoly was used to oxpare with a '

conservative generic creep col) apse analysis. The collapse time for the ucst limiting assenbly was conservatively determined to be 'ncre than 35000 EFIH l

(effective full power hours), which is greater than the maximum projected l residence time (Table 4-1). The creep collapse analysis is based on reference 7.

B. Grav APSR l l

The gray ' S Rs used in cycle 9 are designed for improved creep life. i Cladding diickness and rod cuality, %ttich are the primary factors controlling the time until creep collapse, are imprcned to extend the life of the gray 4-2 l

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APSR. 'Ihe minim.n design claddity thickness of the Mark B black APSR is 18 mils, while that of thi MK-D gray APSR is 24 mils. Additionally, the gap width between the end plug and the Iroonel absorber material is reduced.

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

'Ibe creep collapse analysis of the gray APSR shows that it will not creep collapse during the projected lifetime of the rods. 'Ihe gray APSR is shown in Figum 4-5.

4.2.2 Claddina Stmss A. Fimi Bad

'Ihe NO-1 cycle 9 stress parameters are enveloped by a conservative fuel rod stress analysis. 'Ihe same method was used for analysis of cycle 9 that had been used on the previous cycle.

B. Gray APSR

'Ihe gray APSR design demonstrates the ability to meet specified design requirements. 'Ibe APSR cladding stress analysis includes pree.sure, tenperature and ovality effects. 'Ihe gray APSR has sufficient cladding and weld stress margins.

4.2.3 Claddina Strain A. Fuel Rod

'Ihe strain analysis is basal on the tq.per tolerance valtes for the fuel pellet diameter and density and the lower tolerance for the cladding inside diameter. 'Ihe fuel design criteria specify a 1.0% limit on claddirg plastic tensile circumferential strain. 'Ibe pellet is designed to ensure that plastic claddirg strair. is less than 1.0% at design local pellet burnup and heat generatico rat s. 'Ihe design values are higher than the worst-case values the ANO-1 cjcle 9 fuel is expectal to see.

B. Gray APSR

'1he gray APSR strain analysis includes thermal ard irradiation swelling effects. 'Ihe results of this analysis show that no cladding strain is induced due to thernal expansion or irradiation swelling of the Inoonel absorber.

4-3

M Thermal Desian All fuel assernblies in the cycle 9 core am thermally similar. h design of the batch 11 Mark B6 assenblies is such the, the thermal performance of this fuel is equivalent to the fuel design used in the remainder of the core. The analysis for all fuel was performed with the TA002 code as described in reference 8. Nominal undensified input paraneters used in the thenal j analysis are presented in Table 4-2. Densification effects were accounted for in TA002.

'Ihe results of the thermal design evaluation of the cycle 9 core are sumarized in Table 4-2. Cycle 9 core protection limits are based on a linear heat rate (IHR) to c:nterline fuel melt limit of 20.5 kW/ft as deter-mined by the TACO 2 code.

The maximum fuel asserbly burnup at EDC 9 is predicted to be less than 42,800 rfWd/mtU (batch 9B). 'Ihe fuel rod internal pressures have been evaluated with TACD2 for the highest burnup fuel rods and are predicted to be less than the naninal reactor coolant pressure of 2200 psia.

4.4 Material Desian The chemical carpatibility of possible fuel-cladding-coolant-assenbly interactions for batch 11 fuel aMlies is identical to those for previous fuel asserblies because no new materials were introduced in the batch 11 fuel asuxt:blies.

4.5. Oneratina Experience NWk & Wilcox operating experien e with the Mark B 15x15 fuel aMly has verified the adequacy of its design. The accuculated operatin; experience for eight B&W 177 fuel aMly plants with Mark B fuel is shown in Table 4-3.

4-4

Table 4 Fbel Desian Pgrameters and Dimensions Batd 6D Bat & 9B Bat & 10 Bat & 11 Fuel assembly type M:(-B4 MK-B4 MK-B4 MK-B6 Number of assemolies 1 52 64 60 Ibel rod OD raninal, 0.430 0.430 0.430 0.430 in Pal rod ID runinal, 0.377 0.377 0.377 0.377 in Urdensified active 142.25 141.8 141.8 141.8 fuel lergth, in Ibel pellet CD, 0.3695 0.3686 0.3686 0.3686 (mean), in Fuel pellet initial 94 95 95 95 density, (Ncan), % 7D Initial fuel enrichment, 3.19 3.30 3.35 3.45 wt. % U-235 Average burnup, BDC, 20800 26400 17300 0 Wd/mtU Exposure tire, EDC, 28700 31300 20600 10100 EFTH Cla& ling colla,%2 >35000 >35000 >35000 >35000 time, EFTH 4-5 4

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Table 4-2. Fuel % h Bats 6D Batch 9B Bat & 10 Bat & 11 No. of aste rtlies 1 52 64 60 Initial density, % 'ID 94 95 95 95 Irltial pellet OD, in 0.3695 0.3686 0.3686 0.3686 Initial stack height, in 142.25 141.80 141.80 141.80 Enridiment, wt % U-235 3.19 3.30 3.35 3.45 Ncninal linear heat rate 5.73 5.74 5.74 5.74 at 2568 Et, W/ft(a)

TACO 2-based Predictions Average foal tenperature at naninal Um, F (BOL) 1406 1400 1400 1400 Miniruc UR to melt, W/ft 20.5 20.5 20.5 20.5 Core average um = 5.74 W/ft (a) Based on a naminal stack heiaht l

4-6

m -

- . ~.

I Table 4-3. Ooeratirn ExDerie.n Omulative Current Max FA bumun,1641/mt!.I(a) electric Reactor Cvele Incom Disduggg) output.)Mh(D)

Ooonee 1 10 45,908 50,598 66,183,044 Coonee 2 9 40,580 41,592 60,968,626 Ooonee 3 10 33,290 39,701 60,843,663 Th me Mile Island 6 26,090 33,444 29,469,976 Arkansas Nuclear One, 8 51,540 47,560 51,626,035 Unit 1 Rancho Seco 7 26,242 38,268 39,045,954 Crystal River 3 6 35,350 31,420 38,512,798 Davis-Besse 5 36,960 32,790 25,236,663 (a;As of October 31, 1987.

(b)As of M--Aer 31, 1986.

4-7 i

Figure 4-1. Removable Upper End Fitting (Side View)

SLOT STCP

/. L% T /

W

= T

% e f . hf l ) b l l

n - , -

I k <> f Q . . -__

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• There are two stop pin N1es on each side of the upper end fitting. One contains a stop pin and the other is a spare.

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Figure 4-2. Holddown Spring Retainer FOOT k

1. STOP PIN LEDGE

~ uc e -

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5. bu2 EAR DESIGN 5.1. Ihysis Characteristics Table 5-1 lists the oors physics parameters of design cycles 8 ard 9. The values for both cycles were calculated with tM N00 DIE code .9 Figuru 5-1 illustrates a representative relative power distribution for the 1xqinning of cycle 9 at full power with equilibrium xenon ard ncuninal rod positions.

The differences in feed enrichment, BPRA loading, ard shuffle pattern caused little change in the physics parameters between cycles 8 and 9. Calculated ejected red worths ard their adherence to criteria are considered at all times in life ard at all power levels in the developnent of the rod position limits presented in section 8. The maxinum stuck rod worths for cycle 0 are less than for cycle 8 at all times in cycle. All safety criteria a e lated with these worths are met. The adequacy of the shutdown margin with cycle 9 stuck rod worths is denonstrated in Table 5-2. The following conservatisms were applied for the shutdckn calculations:

1. Poison material depletion allowance.

2. 10% uncertainty on ret rod worth.

3. Flux redistribution penalty.

Flux redistribution was accounted for since the shutdchn analysis was calculated usirg a two-dimensional model. The reference fuel cycle shutdown margin is presented in the Ato-1 cycle 8 reload report.10 5.2. Analvtical Intut The cycle 9 incore measurement calculation ocratants to be used for ocrputing core power distributions were prepared in the same manner as those for the reference cycle.

5.3. Charnes in }Melear DesiqD The core design changes for cycle 9 are the use of gray APSRs and the replacement of the Inoonel intermediate spacer grids with Zircaloy spacer grids. Gray APSRs, which are longer ard use a weaker absorber (Inconel),

5-1

replace the silver-iniium-cadmium APSRs ussi in all previous cycles.

Calculations with the standard three-dimensional model verified that those APSRs provide adequate axial power distribution control. The substitution of Zircaloy spacer grids reduces the parasitic absorption of neutrons and has a beneficial effect on fuel cycle cost.

The gray APSRs will be withdrawn fran the core near the end of cycle 9 (360 EFPD) where the stability and control of the core in the feed-and-bleed made with APSRr, ruoved has been analyzed. The calculated stability index at 364 EFID without APSRs is 4.037 h-1 which dcunonstrates the axial stability of the core. The calculational methods used to obtain the important nuclear design parameters for this cycle were the same as those used for the reference cycle. The operating limits (Technical Specifications changes) for the reload cycle are given in section 8.

5-2 l

I Table 5-1. Ittysics Parameters for AND-1. Cveles 8 and 9(a)

Cvele 8(D) Cvele 9(c)

Cycle length, EFIO 420 420 Cycle bu]mup, Mid/mtU 13,147 13,143 Average core burnup - BOC, Mid/mtU 25,522 27,271 Initial core loading, atU 82.0 82.1 Oritical boron - BOC, ppa (no Xe)

HZP(d), grtup 8 inserted 1644 2552 HFP, grtup 8 inserted 1409 1379 critical boren - IDC, ppn (eq Xe)

HZP, group 8 inserted 651 HTP, group 8 inserted 18 539(e)

O Ocritrol rod worths - HFP, IOC, % ak/k Group 6 1.14 1.11 Group 7 1.49 0.98 Group 8 (maxinun) 0.39 0.19 Control ro:t worths - HFP, IDC, % ak/k Group 7 1.52 1.05 Max ejected rod worth - HZP, % ak/k IOC (L-10), groups 5-8 ins 0.55 0.35 360 EFTD (Ie10), groups 5-8 ins 0.60 0.41 IDC (Ie10), groups 5-7 ins 0.59 0.41 Max stuck red worth - HZP, % ak/k IEC (N-12), groups 1-8 ins 1.58 1.49 360 EFTV (N-12), groups 1-8 ins 1.86 1.47 IDC (N-12), groups 1-7 ins 1.63 1.42 Power deficit, HZP to HFP, % ak/k BOC 1,56 1.60 IDC 2.34 2.35 Dt5pler coeff - HFP,10-4 (a@)

ICC (no xe) -0.154 -0.159 IDC (eq Xe) -0.184 -0.186 5-3

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

Cycle 8(D) Cycle 9(C)

Moderator coeff - HFP,10-4 (ak/k/0F)

BOC (no Xe, crit ppn, group 8 ins) -0.51 -0.58 EOC (eq Xe, O ppe, group 8 out) -2.78 -2.82 Boron w rth - HFP, yptV% Ak/k BOC 129 130 EOC 111 111 9

Xenon wrth - HFP, % Ak/k BOC (4 EFID) 2.55 2.56 IDC (equilibrium) 2.72 2.71 Effective delayed neutron fraction - HFP IOC 0.0062 0.0062 EDC 0.0052 0.0052 (a) Cycle 9 data are for the conditions stated in this report. The cycle 8 core conditions are identified in reference 9.

(D) W on 425 EFPD at 2568 )Wt, cycle 7.

I (c) Based on 440 EFID at 2568 PWt, cycle 8.

.I

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

(*)At HFP conditions, O ppm emws at 411 EFIV.

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___.m____._._______.______ __ _ _ _ . _ _ _ . _ _ _ _ _ . _ . _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _

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Table 5-2. Shu+**m Maruin calmlation for ANO-1. Ovele 9 i

BOC, 360 EFPD, 420 EFTO,

% Ak/k . % Ak/k  % Ak/k Available Rod Worth Total rut worth, HZP 8.699 9.265 9.222

, i i Worth reduction due to poison material burmp -0.100 -0.100 -0.100 Mav4== stuck rod, HZP -1.490 -1.471 -1.419 Net Worth 7.109 7.69s 7.703 l Imss 10% uncertainty -0.711 -0.769 -0.770 I

Total available worth 6.398 6.925 6.933 Reauired Rod Worth ,

FCWer deficit, HFP to HZP 1.602 2.291 2.351 i Allowable insertad rod worth 0.276 0.422 0.422 Flux redistribution 0.344 0.616 0.573 r

7btal riglired worth 2.222 3.329 3.346 Shutdown margin (total i available worth aims l total rigtired worth) 4.176 3.596 3.587 1

4 j EZUt The required shutdown margin is 1.00% ak/k.

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5-5

1 Figure 5-1. ANO-1 Cycle 9. BOC (4 EFPD) Two-Dimensional Relative Power Distribution -- Full Power. Equilibrium Xenon, Normal Rod Positions 8 9 10 11  !? 13 la 15 H 0.98 1.12 1.18 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 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

! O 1.29 1.25 1.31 1.07 1.09 0.42 P 0.94 1.18 0.93 0.61 0.32 i

i R 0.41 0.53 0.39 1

4 Inserted Rod group No.

x.xx Relative Power Density 1

f 5-6

6. 'IHERMAIrHYmAUIlC DESIGN 2e thermal-hydraulic design evaluation supporting cycle 9 operation utilized the methods ard models described in references 10,11, and 12 as supplemented by reference 4, which inplements the BWC (reference 13) CHF correlation for analysis of the Zircaloy grid fuel assembly. 'Ihe analyses presented in Section 5 of refererre 4 demonstrate that changes in the flow parameters resulting fran the incorporation of Zircaloy spacer grids do not significantly impact the thermal-hydraulic characteristics of the Zircaloy grid ccre relative to the Inconel grid core values. Implementation of the Zircaloy grid fuel assemblies rto existing reactors, however, is perfonred on a batch basis, with the tr 4nsition cycles having both Zircaloy grid and Inconel grid fuel assemblies.

In a transition core, the Zircaloy grid fuel assemblies, which have a slightly higher pressure drop than the Inconel grid assenblics due to the higher flow resistance of the Zircaloy grids, tend to divert same flow to the Inconel grid fuel. 21s creates the need to consider a "transition core penalty". Se amount of coolant flow reduction in the limitiry Zircaloy grid assembly and consequently, the magnitude of the transiticn penalty, is dependent on the number of Zircaloy grid assemblies (with the smaller nimber of Zircaloy grid assemblies beirg more limiting) .

Another contributire factor in determining the transition core penalty is the  ;

1 core bypass fraction which is depe:x$ent on the number of burnable poison rod l assemblies (BPPAs), since these ocmponents restrict flow throtgh the control rod guide tubes (CRGI's) . For thermal-hydraulic analyses, the most limiting case is that with the higher bypass flow fraction, or smaller number of BPRAs.

We design basis chosen for cycle 9 thermal-hitiraulic analyser was a full core of Zircaloy grid assemblies, containirg 40 BP:%s, for which the core bypass flow is 8.8%. 'Ihis design configuration was used to calculate the 1.77 INBR value shcwn on Table 6-1. 'Ihe actual cycle 9 core configuration consists of 60 fresh Zircaloy grid fuel assemblies, of khich 52 contain BPRAF 6-1

(8.3% core bypass flow) . N DiBR for this configuration, using the same core conditions presented in Table 6-1 is 1.80. h full Zircaloy grid m re configuration is, therefore, conservative for cycle 9 DiBR analyses and a transition core penalty is not mery. h reconstitutable upper end fitting (UEF) and the anti-straddle lower end fitting (LEF) were addressed in the evaluation.

The pressure-tepture safety limits have been recalculated using the BWC CHF correlation in the LYNXIll crossflow analysis. Table 6-1 provides a summary ccmparison of the INB analysis parameters for cycles 8 and 9.

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

6-2

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Table 6-1. Maximum Desian Conditions. Oveles 8 and 9 Cycle 8 Cvole 9 Design power level,IEt 2568 2568 System pressure, psia 2200 2200 Reactor coolant flN, gpn 374880 374880 Core bypass flow, % (a) ** 8.8 D E modeling Cr:,ssflow Crossflcw Reference design radial-local power peaking factor 1.71 1.71 Reference design axial flux rhape 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 0.97 (c)

Active fuel length, in. (b) 141.8 141.8 Avg heat flux at 100% pcuer, 103 Btu /h-ft2 174 174 Max heat flux at 100% power, 103 Btu /h-ft2 492 492 CHF correlation B&W-2 INC 1

CHF correlation ENB lhnit 1.3 1.18 l l

Minimum D E at 112% power 2.08 1.77 (c) j at 102% power (d) 2.37 2.01 l (a)Used in the analysis. l (b) Cold ncminal stack height.

l (c) Calculated for the instrument guide tube subchannel which is limiting for the thrk-B6 assembly.

(d) mis represents initial condition DE for accident analyses. l l

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7. ACCIDDE AND TRANSIDE MG1YSIS I l

7.1. General Safety Analysis Each FSAR accident analysis has been examined with respect to changes in cycle 9 parameters to determine the effect of the cycle 9 reload and to ensure that thermal performance during hypothetical transients is not degraded.

We effect of fuel densification on the ESAR accident results has been evaluated and are reported in referen 15. Since batch 11 reload fuel assemblies contain fuel rods whose theoretical density is higher than those consideral in the reference report, the conclusions in that reference are still valid.

We radiological dose consequences of the accidents presented in Chapter 14 of the updated IEAR were re-evaluated for this reload report except for the waste gas tank rupture. W e waste gas tank rupture was not reeva.luated since Technical Specification 3.25.2.5 controls the maxirum tank inventory on the basis of Xe-133 equivalent curie content such that the analysis of the event is not cycle dependent. We evaluation of the remaining events was made in ortler to incorporate nore current plant data as well as the information in the updated IT#a.

All of the Cycle 9 accident doses are based on radionuclide sources calculated for the actual Cycle 9 core design and irradiation history. In addition, the bases used to analyze same of these accidents were char.]ed to )

l be consistent with the bases in the updated FSAR. Se significant  !

l differences in the bases for the acx:ident analysis between cycle 8 'and cycle j t 9 are: )

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i l o me atmogheric dispersion factors have been increased slightly.

1 o credit has been taken for the penetration roam filter system in l calculatirg the doses associated with the control rod ejection )

accident. (mis makes the control rod ejection accident consistent with the IDCA arxl MHA.)

l 7-1

I o 'Ibe iodine removal rate used to calculate the IDCA ard MHA doses for Cycle 9 was changed to be consistent with the updated FSAR. l l

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

Table 7-1 shows a ocmparison between cycle 8 aM cycle 9 doses for the Chapter 14 acx:idents that result in significant offsite doses. With the exception of the maximum hypothetical accident (MHA), ell doses am either bounded by the values reported for cycle 8 or are a small fraction of the 10CFR100 limits, i.e., below 30 Ram to the thyroid or 2.5 Rem to the whole body. For the MHA, the doses capare to the criteria as follows:

1. 'Ihe 2-hour thyroid dose at the exclusion area boundary (EAB) is 165.1 Rem (55% of the NURE& 0000 limit) .

2. 'Ihe 2-hour whole-body doce at the EAB is 5.0 Rem (20% of the IURM-0800 limit).

3. 'Ihe 30 day thyroid dose at the low population zone (IPZ) is 87.8 Rem (29% of the IURM -0800 limit).

'Ihe radiological dce frm all of the accidents evaluated with the specific nuclide inventory frm cycle 9 aro lower than the IGC acceptance criteria of IURM-0800, and thus are within acceptable limits.

7.2. Accident Evaluation

'Ihe key parameters that have the gmatest effect on deteminiro the outcene of a transient can typically be classified in three major areas: core thermal parameters, thermal-hydraulic parameters, ard kinetics parameters, including the reactivity f=hek coefficients and control rod worths.

'Ihe core thermal properties used in the ESAR accident analysis were design operatirq values based on calculational values plus uncertainties. Thermal parameters for fuel batches 6D, 9B, 10, and 11 are given in Table 4-2. 'Ihe cycle 9 theraal-hydraulic maxinum design conditions are ccmpared with the previous cycle 8 values in Table 6-1. 'Ihese parameters are cxanmon to all the 7-2

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Occidents considered in this report. The key kinetics parameters frcm the ISAR and cycle 9 are capared in Table 7-2.

i A generic IDCA analysis for a B&W 177-FA, lowered-loop ?ES has been performed ,

using the Final Acceptance Criteria ECCS Evaluation Model (reported in BAW-1010316, BAW-1010417, and BAW-1915P18). This analysis is generic sin the limiting values of key parameters for all plants in this category were used.

Furthermore, the ccrnbination of average fuel t@tures as a function of UIR and lifetime pin pressure data used in the B.4l-1915P IDCA limits analysis is conservative cmpared to those calculated for this reload. Thus, the analysis and the IDCA limits reported in BAW-1915P provide conservative results for the operation of the reload cyc13. Table 7-3 shows the bounding values for allowable IDCA peak UIRs for NKF-1 cycle 9 fuel. These IRR limits l include the effects of IURED 0630, TACD2, nl FIECSEI'.

It is concluded frcrn the examistion of cycle 9 core thermal ard kinetics properties, with respect to acceptable previcus cycle values, that this core reload will not adversely affect the NK)-1 plant's ability to operate safely during cycle 9. Considerirq the previous]y accepted design basis us(xl in the FSAR and subsequent cycles, the transient evaluation of cycle 9 is considered to be bounded by previously accepted analyses. The init.ial conditions for j the transients in cycle 9 are bounded by the FSAR, the fuel densificaticn ,

1 report, ard/or subsequent cycle analyses. I I

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Table 7-1. Coroarison of Cycle 8 and Cycle 9 Accident Ibses Cycle 8 doses, Cycle 9 doses, Pm ____ Pm Fuel Handlina Accident S yroid dose at EAB (2 h) 1.15 1.12 Whole body dose at EAB (2 h) 0.21 0.22 sten Line Break hyroid dose at EAB (2 h) 1.71 1.82 Whole body dose at EAB (2 h) 0.008 0.01 Steam Generator Tube Failure Byroid dose at EAB (2 h) 6.14 6.53 Nhole body dose at EAB (2 h) 0.52 0.56 Contiel Rod Eiection Accident S yroid dose at EAB (2 h) 12.2 7.02 hhole body dose.it EAB (2 h) 0.008 0.006 Wyroid dose at LPZ (30 d) 9.(9 5.64 Whole body dose at LPZ (30 d) 0. (. 05 0.005 IDCb Syroid dose at EAB (2 h) 4.02 4.22 Khole body dose at EAB (2 h) 0.026 0.03 Ryroid dose at LPZ (30 d) 2.05 2.47 bhole bcdy dose at IPZ (30 d) 0.018 0.02 Maxim m Hvoothetical Accident

'Ihyroid dose at EAB (2 h) 157.3 165.1 Whole body dose at EAB (2 h) 4.80 5.03 S yruid dose at LPZ (30 d) 73.0 87.8 Nhole body dose at IPZ (30 d) 1.56 1.78 7-4

Table 7-2. Ccrparison of Key Parameters for Accident Analysis -

EAR ard Densification AIK)-1 Parameter Recort Value Cycle 9 Doppler coeff - (BOC) , 10-4 Ak/k/OF -0.117 -0.159 Doppler coeff (II)C),10-4 Ak/k/OF -0.130. -0.186 Moderator coeff- (BOC),10-4' Ak/P/OF 0.0(a) -0.58 Moderator coeff (EOC),10-4 Ak/k/OF -4.0(b) -2.82 All-rod grcup Worth (HZP), % Ak/k 12.90 8.70 Initial boron concentration, ppn 1150 1379 Boron reactivity worth (HFP), 100 130 ppig/% Ak/k Max. ejecttd rod worth (HFP), % Ak/k O.65 0.23 Dropped rod worth (HFP), % Ak/k 0.65 $0.20  ;

(a) +0.5 x 10-4 Ak/k/OF was used for the mcderator dilution analysis.

(b) -3,0 x 10-4 Ak/k/OF was used for the steam line failure analysis.

Table 7-3. Bourdirq Values for Allowable LOCA Peak Linear Heat Rates-1 Allcuable Allowable i Core peak IHR, peak IHR,' i elevation, 0-1000 M4d/mtU, after 1000 M4d/mtU, ft kW/ft kW/ft 1

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 7-5

8. HOPOSED }ODIFICATIONS TO TEQ9TICAL SPECIFICATIONS We Technical Cpecifications have been revised for cycle 9 operation for changes in core reactivity, pwer peaking, and control red worths. The l

Technical Specifications were also revised to describe design features inplemented with cycle 9. The cycle 9 design analysis basis includes the inpact of exterded periods of cycle 8 lw-power operation, with cycle 8 pwer levels ranging between 65% and 100% of rated power. The cycle 9 basis also includes a very lw leakage fuel cycle design, a mixed Mark B4/ Mark B6 fuel assembly wre, gray APSRs, gray APSR withdrawal flexibility, and crossflw analysis. The safety limits in Technical Specification Section 2 (Figures 8-1 through 8-3), have been charged for cycle-sr 3cific cralits in the fuel cycle design, which allowed for additional operating margin beyorx1 the generic limits used for cycle 8. Error adjusted trip setpoints for the reactor protection system are shown in Figure 8-4. The IDCA linear heat rate limits used to develop the 'Dachnical Specification Limiting Conditions for Operation include the impact of NUREG-0630 cladding swell and rupture model, and j inplement the credit frczn FIECSET analyses.18 l A cycle 9 specific analysis was conducted to generate Technical Specification Limiting Conditions for Operation (rod index, axial power imbalance, and l quadrant tilt), based on the methodology described in reference 19. The effects of gray APSR repositioning were included in the analysis, as was an  ;

APSR withdrawal flexibility windw of +50/-10 EFPDs. The burnup-dependent allwable LOCA linear heat rate limits used in the analysis are provided in j Figure 8-17. The analysis also determined that the cycle 9 Technical l Specifications provide protection for the overpower condition that could oxur during an overcooling transient because of nuclear instrumentation errors, and verified removal of the pwer level cutoff hold requirement.

Technical Specification section 3.5.2.4 was revised to ac.c.umcdate a change in the quadrant tilt setpoint. The measurement system-independent rod position arxi axial power imbalance limits determined by the cycle 9 analysis 8-1 I

were error-adjusted to generate alam setpoints for power operation and are reflected in a 'Ibchnical Specification revision to sections 3.5.2.5 and 3.5.2.6. 'Ihe error adjusted alam setpoints are provided in Figures 8-5 through 8-16. Technical Specification section 5.3.1 was' revised to include the reconstitutable fuel assernbly design and gray anal power shaping rods in the design features.

Based on the analyses and Technical Specification revisions described in this report, the Final Acceptance Criteria EOCS limits will not be eW, nor will the themal design criteria be violated. 'Ihe following pages contain the revisions to the Technical Specifications.

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2. SAFEIY LIMITS AND IlMITDC SAFEIY SYSTD4 SETI' DES 2.1 SAFETI LIMITS, REACIOR CORE Acolicability Applies to reactor themal pwer, reactor pwer imbalance, reactor coolant system pressure, coolant tenperature, and coolant flow during power operation of the plant.

Obiective To maintain the integrity of the fuel cladding.

Specification 2.1.1 he ccrbination of the reactor systen pressure and coolant temperature shall not excee( the safety limit as defined by the locus of points established in Figure 2.1-1. If the actual pressure / temperature point is below and to the right of the pressure / temperature line the safety limit is exceeded.

2.1.2 he cambination of reactor thermal power ard reactor power imbalance (power in the top half of the core minus the power in the bottctn half of the core expressed as a percentage of the rated power) shall not exceed the safety limit as defined by the locus of points for the specified flow set forth in Figure 2.1-2. If the actual-reactor-thermal-power / reactor-power-imbalance point is above the line for the specified flow, the safety limit is exceeded.

Bases To maintain the integrity of the fuel cladding and to prevent fission product release, it is necessary to prevent overheating of the cladding under normal operating coMitions. 21s is accct:plished by operating within the nucleate boiling regime of heat transfer, wherein the heat transfer coefficient is large enough so that the clad surface temperature is only slightly greater than the coolant temperature. Be upper boundary of the nucleate boiling regime is termed departure frun nucleate boiling (DG) . At this point there is a sharp ra:1uction of the heat transfer coefficient, which could result in l high cladding temperatures and the possibility of cladiing failure. Although 30 is not an observable parameter during reactor operation, the observable parameters of neutron power, reactor coolant flow, tenperature, and pressure can be related to DB through the use of a critical heat flux (CHF) correlation. Be B&W-2(1) and BWC(2) correlations have been developed to predict De and the location of De for axially unifom aM non-unifom heat flux distributions. B e B&W-2 correlation applies to Mark-B fuel and the BWC correlation applies to Mark-BZ fuel. Se local Da ratio (DBR), defined as the ratio of the heat flux that would cause Da at a particular core location to the actual heat flux, is indicative of the margin to Dm. Se minimum value of the DBR, during steady-state o mration, romal operatiomi transients, and anticipated transients is lim'.ted to 1.30 (B&W-2) and 1.18 (BWC).

8-3

A DE of 1.30 (MW-2) or 1.18 (BWC) corresponds to a 95 percent probability l at a 95 percent confidence level that DB will not occur; this is considered a conse.Ivative margin to Da for all operating conditions. We difference between the actual core outlet pressure and the indicated reactor coolant system pressure for the allowable reactor coolant punp cmbination has been l l considered in determining the core protection safety limits. I he curve presented in Figure 2.1-1 represents the conditions at which the DM is greater than or equal to the minimum allowable Da for the limiting ccubination of thermal power and number of operating reactor coolant pumps.

his curve is based on the following nuclear power peaking factors (3) with potential fuel densification effects:

=2.83;(H = 1.n; = 1.65.

We curves of Figure 2.1-2 are based on the more restrictive of two themal limits and inc1trie the effects of potential fuel densification:

1. he DE limit produced by a nuclear power peaking fact.or of % =

2.83 or the ccobination of the radial peak, axial peak and position of the axial peak that yields no less than the DE limit.

2. We ccrnbination of radial and axial peak that prevents central fuel melting at the hot spot. We limit is 20.5 kW/ft.

Power peaking is not a directly observable quantity and therefore limits have been established on the basis of the reactor power imbalance prv* y w] by the power peaking.

We flow rates for curves 1, 2, and 3 of Figure 2.1-3 correspord to the expected mininum flow rates with four punps, three pumps, and one punp in each loop, respectively.

We curve of Figure 2.1-1 is the most restrictive of all possible rmetcr ecolant pump maximum themal power ccabinations shown in Figure 2.1-3. We curves of Figure 2.1-3 represent the coniitions at which the DE ' limit is l predicted at the maxinum possible themal power for tne number of reactor coolant punps in operation. We local quality at the point of mininum DE is less than 22 percent (MW-2)(1) or 26 percent (BWC)(2) . l Using a local quality limit of 22 percent (MW-2) or 26 percent (BWC) at the point of mininnn DE as a basis for curves 2 ard 3 of Figure 2.1-3 is a conservative criterion even though the quality at the exit is higher than the quality at the point of minimum DE.

We DM as calculated by the NW-2 or the BWC correlation continually l increases frcan point of minimum DE, so that the exit DE is always higher ard is a function of the pressure.

We raximum themal power, as a function of reactor coolant punp operation is limited by the power level trip produced by the flux-flow ratio (percent flow x flux-flow ratio), plus the appropriate calibration ard instrumentation errors.

8-4

For each curve of Figure 2.1-3, a pressure-temperature point above and to the left of the curve would result in a WBR greater than 1.30 (B&W-2) or 1.18 (BWC) or a local quality at the point of minimum NBR less than 22 percent (B&W-2) or 26 percent (EHC) for that particular reactor coolant punp eituation. Curve 1 of Figure 2.1-3 is the most restrictive because any pressure-tenperature point above and to the left of this curve will be above and to the left of the other curves. l RgmJW (1) Correlation of Critical Heat Flux in a Bundle Cooled by Pressurized Water, BAW-10000A, May, 1976.

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

(3) FSAR, Section 3.2.3.1.1.c.

8-5

N Figure 8-1. Core Protection Safety Limit -- ANO-1 (Tech Spec Figure 2.1-1) 2400 2200 cn ACCEPTABLE

OPERATIOfl c.

J

$ 2000 a

et a UNACCEPTABLE j OPERATION a

g 1800 7

/

v 1600 580 600 620 640 660 Reactor Outlet Temperature, OF 8-6

Figure 8-2. Core Protection Safety Limits -- AN0-1 (Tech Spec Figure 2.1-2)

Thermal Power Level

% FP

. 140

- 120

(-33.04,112.0) - (33.04,112.0)

ACCEPTABLE l - - 100 l (45.27,100.55)

OPER N

^

1 I

(-33.04,90.75) l(33.04,90.75)

ACCEPTABLE I

(-62.32,84.45){ l 4 & 3 PUMP -

- 80 (45.27,79.30)

OPERATION l l I I l I I

(-62.32,63.20) l (33.04,64.08)

~ ~ 60 ,

(-33.04,64.08)l ACCEPTABLE l l 4,3&2 PUMP l >(45.27,52.63) l OPERATION l l l

1 -

- 40 l l

(-62.32,3,6.53) g l

i I I I

- 20 l l l I I I I I I I I I li i l i i i i l

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

8-7

Figure 8-3. Core Protection Safety Limits - ANO-1 (Tech Spec Figure 2.1-3) 2400 2200 I f en 7

g 2

2 ~

h 2000 E

c.

U U

8 1800 ft 1500 580 600 620 640 660 Reactor Outlet Temperature, F CURVE GPM POWER PUMPS OPERATING (TYPE OF LIMIT) 1 374,880 (100%)* 112% FOUR PUMPS (DNBR LIMIT) 2 280,035(74.7%) 90.8% THREE PUMPS (QUALITY LIMIT) 184,441 (49.2%) 63.7% ONE PUMP IN EACH LOOP (QUALITY LIMIT) !

3 I

• 106.5% OF DESIGN FLOW l

8-8

2.3 LIMITEC SAFEIY SYSTD4 SEITDES, PRuiu.;nVE INSIRUMENTATION Arolicability Applies to instruments ronitoring reactor power, mactor power imbalance, reactor coolant system pressure, mactor coolant outlet tenperature, flow, number of punps in operation, and high reactor building pressure.

Obiective To provide autcmatic protection action to prevent any ccabination of process variables frca exceeding a safety limit.

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

Bases

'Ihe reactor protection system consists of four instrument channels to monitor each of several selected plant conditions which will cause a reactor trip if any one of these conditions deviates frcxn a preselected operating range to the degree that a safety limit may be reached.

'Ihe trip setting limits for protection system instrumentation are listed in Table 2.3-1. 'Ihe safety analysis has been based upon these protection system instrumentation trip setpoints plus calibration and instrumentation errors.

Nuclear Overpower A reactor trip at high power level (neutron flux) is provided to prevent damage to the fuel cladding frcra reactivity excursions too rapid to be detected by pressure and tenperature measurements.

During normal plant Operation with all react >r coolant punps operating, reactor trip is initiated when the reactor power level reaches 104.9 percent of rated power. Adding to this the possible variation in trip setpoints due to calibration ard instrument errors, the maximum actual power at which a trip would be actuated cculd be 112%, which is the value used in the safety analysis.

A. Overpower Trip Basal on Flow and Imbalance

'Ihe pcuer level trip setpoint produced by the reactor coolant system flcw is based on a power-to-flow ratio which has been established to ach.,m;date the most severe thermal transient considered in the design, the less-of-coolant-flow accident frca high power. Analysis has deronstrated that the specified power-to-flow ratio is adequate to prevent a INBR of less than 1.30 (B&W-2) or 1.18 (BWC) should a low flow condition exist due to any electrical malfunction.

8-9 l

1 1

The power level trip setpoint pMM by the power-to-flow ratio provides both high power level and low flow protection in the event the reactor power level imreases or the reactor coolant flow rate decreases. The power level trip setpoint prrrhrai by the power-to-flow ratio provides overpower DiB protection for all modes of pump operation. For every flow rate there is a maximum permissible power level, and for every power level there is a mininum permissible low flow rate. Typical power level and low flow rate ocznbinations for the pu::p situations of Table 2.3-1 are as follows:

1. Trip would occur when four rector coolant punps are operating if power is 107 per nt and reactor flow rate is 100 percent or flow rate is 93.5 percent and power level is 100 percent.

2. Trip would m,m when three reactor coolant punps are l

I operating if power is 80 percent and reactor flow rate is 74.7 l percent or flow rate is 70 percent and power level is 75 percent.

3. Trip would occur when one reactor coolant punp is operating in each loop (total of two punps cperating) if the power is 52 percent and reactor flow is 49.2 percent or flow rate is 45.8 percent and the power level is 49 percent.

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

No penalty in reactor coolant flow through the core was taken for an open core vent valve because of the core vent valve surveillance program during I each mfueling outage. For safety analysis calculations the maximum calibtation a d instrumentation errors for the power level were used.

The power-inbalance boundaries are established in order to prevent reactor j thennal limits frtxn beirg exceeded. These thermal limits are either power l peaking kW/ft limits or NBR limits. The reactor power inbalance (power in l top half of core minus power in the bottczn half of core) reduces the power

! level trip produ d by the power-to-flow ratio so that the boundaries of I

Figure 2.3-2 are prrrhM. The power-to-flow ratio reduces the power level I trip associated reactor power-to-reactor power itnbalance bourdaries by 1.07 percent for a 1 percent flow reduction.

l B. Punp Monitors I In conjunction with the power 12 balance / flow trip, the punp

! nonitors prevent the mininum core WBR frcxn decreasing below 1.30 f (B&W-2) or 1.18 (BWC) by tripping the reactor due to the loss of l

! reactor coolant pmp(s) . The punp monitors also restrict the power level for the nunber of punps in cperation.

8-10

I C. RCS Pressitre During a startup accident frcn low power or a slow rod withdrawal frcm high power, the system high pressure trip is reated 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 th:2 system pressure below the safety limit (2750 psig) for any design transient. (2)

The icw pressure (1800 psig) and variable low pressure (11.75 Tout

-5103) trip setpoint shown in Figure 2.3-1 have been establishal to maintain the DB ratio greater than or equal to the minimum allowable DG ratio for those design accidents that result in a pressure reduction.(2,3)

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

D. Coolant Outlet Temperature The high reactor coolant outlet temperature trip setting limit (618F) shcun in Figure 2.3-1 has been established to prevent excessive core coolant tmperatures in the operatirq range. Due to calibration ard instrumentation errors, the safety analysis used a trip setpoint of 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 failure in the reactor building or a loss-of-coolant accident, even in the absence of a low reactor coolant system pressure trip.

F. Shittdown 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 inposed when the bypass is used:

1. A nuclear overpchur trip setpoint of 55.0 percent of rated power is autcnatically imposed during reactor shutdcun.

2. A high reactor ecolant system pressure trip setpoint of 1720 psig is autcnatically inposed.

8-11

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

Thermal Power Level, %FP

- 140

-- 120

(-18.0,107) '

- (18.0,107) l - - 100 l l ACCEPTABLE g i 4 PUMP '

l (34.7,90.1)

OPERATION l I

(-18.0,79.9) --8 (18.0,79.9)

(-51.0,74.0)

ACCEPTABLE 3 & 4 PUMF l

l l OPERATION (34.7,63.0) l - - 60 l

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

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

I l l l

l (-51.0,19.6) , . 20 l l l

l l l l l l l l l

t I i il li li  :

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

1 8-12

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 percent of the thennal pwer allowable for the reactor coolant pump ccanbination my continue provided the rods in the group are positioned such that the red that was declared inoperable is contained within allowable grcup average position limits of Specification 4.7.1.2 and the withdrawal limits of Specification 3.5.2.5.3.

3.5.2.3 Tne worth of sirgle inserted control rods during criticality are limited by the restrictions of Specification 3.1.3.5 ard 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 exe 4.12%,

reduce power so as not to exceed the allowable power level for the existing reactor coolant pmp ccrnbination less at least 24 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 follcuire adjustments in setpoints and limits shall be made:

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

b. The control rod group ard APSR withdrawal limits shall be reduced 2% in pcuer for each 1% tilt in excess of 4.12%.

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

3. If quadrant tilt is in excess of 25%, except for physics tests or diagnostic testirg, the reactor will be pla&d in the hot shutdown condition. Diagnostic testirg during power operation with a quadrant power tilt is pemitted provided the themal power allcuable for the reactor coolant pmp cmbination is restricted as stated in 3.5.2.4.1 above.

4. Quadrant tilt shall be monitored on a mininum frequency of once every two hcurs durire power operation above 15% of rated pwer.

8-13

3.5.2.5 Control red positions:

1. 'Ibchnical Specification 3.1.3.5 (safety red withdrawal) does l not prohibit the exercising of individual safety rods as required by Table 4.1-2 or apply to inoperaole safety rod limits in Technical Specification 3.5.2.2.

2. Operating rod group overlao shall be 20% 5 between two sequential groups, except for physics tests.

3. Except for physics tests or exercising control rods, the l control rod withdrawal limits are specified on Figures 3.5.2-l 1(A-C), 3. 5. 2-2 (A-C) , and 3.5.2-3(A-C) for 4, 3 and 2 punp operation respectively. If the aplicable control rod position limits are exW, corrective measures shall be taken imediately to achieva an accepts.ble ocotrol red 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. Except for physics tests or exercising axial power shaping l rods (APSR's), the fcilowing limits apply to APSR position:

1

! Up to 410 EFPD, the APSR's may be positioned as m ary for transient imbalance control, however, the APSRs shall be fully withdrawn by 410 EFPD. After 410 EFPD, the APSR's shall not be reinserted.

With the APSR's inserted after 410 EFPD, corrective measures shall be taken imediately to achieve the fully withdrawn position. Acceptable APSR 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 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 #1ysics tests, imbalance shall be maintained within the envelope defined by Figure 3.5.2-4(A-C) . 7.f 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 imbalance is not achieved within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, reactor power shall be reduced until imbalance limits are met.

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

8-14 j

1 Sises

'Ihe power-irbalance envelope defined in Figure 3.5.2-4(A-C) is based on IOCA analyses wttich have defined the maxinum linear heat rate (see Figure 3.5.2-5), such that the maximum cladding taperature will not exceed the Final Acceptance Criteria. Corrective measures will be taken imediately should the irdicated quadrant tilt, red position, or irbalance be outside their specified boundaries. Operation in a situation that would cause the Final Acceptance Criteria to be approached should a IOCA mir is highly irprobable because all of the pomr distribution parameters (quadrant tilt, rod position, and imbalance) nust be at their limits while l 8-15

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

+10/-0 EFPD -- AN0-1 Cycle 9 (Tech Spec Figure 3.5.2-1A) 110 (84.1,102) (270.8,102) ; -

(300,102) igg _

OPERATION IN THIS REGION IS NOT

( 66.0,90) ,

90 -

ALLOWED SHUTDOWN MARGIN 80 -

(248.0,78)

M E

OPERATION l 70 - RESTRICTED N -

% 60 o

. 50 -

(41.5,48) (212.0.48)

E O 40 30 -

PERMISSIBLE OPERATING 20 (5.5,13) 10

' (0,6.3) 0 0 20 40 60 80 100 120 14 0 160 180 200 220 240 260 280 300 0 20 40. 60 80 100 l l l 1 l l Group 7 0 20 40 60 80 l

100 i

1 1 I I Group 6 0 20 40 60 80 100 t I f f I l Group 5 Rod Index, t, WD 8- 16

Figure 8-6. Rod Position Se+ points for Four-Pump Operation from 27 +10/-0 to 360 +50/-10 EFPD -- ANO-1 Cycle 9 (Tech Spec Figure 3.5.2-1B) 110 100 -

(159.5,102) (268.8,102) - (300,102)

SHUTDOWN MARGIN 90 -

LIMIT (264.0,90)

(244.0,78) 80 u OPERATION E 70 RESTRICTED e

N 60 -

OPERATION IN THIS REGION IS NOT o ALI.0WED (79.5,48) (1 8.0,48) g 40 30 '

20 - PERMISSIBLE OPERATING (31.5,13) REGION 1g ,

' (0,6.3) 0 i e i i I e i e i i e i i 1 0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 ,

0, 20 40 6,0 8,0 100 Group 7 0 20 40 60 80 100 l l e i I I Group 6 0

20 40 60 80 1p0 Rod Index, % WO Group 5 8-17

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

+50/-10 EFPD -- AN0-1 Cycle 9 (Tech Spec Figure 3.5.2-1C) 110 (1 9. ,102) (268.8,102)- n (300,102) 100 - SHUTDOWN MARGIN LIMIT (264.0,90) 90 l

OPERATION 80 -

(244.0,78)

RESTRICTED h 70 - OPERATION IN THIS REGION IS NOT w

g

~

60 -

ALLOWED

", 50 -

(79.5,48) (198.0,48)

E2 j 40 -

30 -

PERMISSIBLE OPERATING l 20 REGION (33.5,13) 10 i

(0.6.3) 0 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 p 2,0 4,0 6,0 80 10,0 Group 6 0 20 40 60 80 100 i l I i i I Group 5 Rod Index, % WD 8-18

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) 110 100 -

90 -

80 (84.6,77) (271.0,77)- (300,77)

, OPERATION IN THIS g 70 -

REGION IS NOT e ALLOWED (266.0,67) e SHUTDOWN g 60 -

MARGIN 248.0,58) g LIMIT g 50 _

, OPERATION b RESTRICTED 2 40 -

E (41.5,36) (212.0,35.5) 30 -

20 -

PERMISSIBLE (5.5,9.75) OPERATING 10 REGION 0 (0,j.75), , , , , , , , , , , , ,

0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 0, 2g 4p 6,0 8,0 igg Group 7 q 2p 4Q p0 8p 190 Group 6 1

0 20 j0 6p 8,0 1p0 Rod Index, % WD Group 5 i

l 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) i 110 100 -

l 90 -

SHUTDOWN N

(150.4,77) (269.0,77) - ':(300,77) 70 -

OPERATION IN THIS (264.0,67) e REGION IS NOT ALLOWED 60 -

OPERATION 244'0,58) g RESTRICTED e 50 -

C I 40 c -

(79.5,36) (198.0.35.5) 30 _

PERMISSIBLE 20 -

OPERATING REGION 10 -

(31.5,9.75)

< 0,4.75) 0 t i e i e 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

,0 2g 49 6p 8,0 190 Group 7 l

' 0 20 40 60 80 100 l I I l t i Group 6 0 20 40 60 80 100 Rod Index, % WD i l i I I i Group 5 I

i l

C- 20 I

Figura 8-10. Rod Position Setpoint: for Three-Pump Operation After 360

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

SHUTDOWN N

(160.4,77) (269.0,77) ^  : (300,77) 70 (264.0,67)

@ OPERATION IN THIS W 60 -

REGION IS NOT (244.0,58) g ALLOWED a 50 -

OPERATION g RESTRICTED

[  !

' 40

~

(79.5,36) (198.0,35.5) 30 20 -

PERMISSIBLE OPERATING REGION 10 - s33.5,9.75) 0.4.75) 0 i t I i i i i i i i i i I '

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

, ,0 30 j0 60 8,0 Ig0 Group 7 p 30 4Q 6p, 8p ig3 Group 6 '

f i 0 4,0 6p 8,0 igg Rod Inc x, % 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) 110 100 -

90 -

80

.a E 70 - l 8 l N 60 -

(85.7,52) (303,52) 50 _ RG IS f (271.3,52) 7 J ALLOWED y (266.0,44)

~

Syy[,jN pShf (248.0.38) 30 - LIMIT (41.E.24) (212.0.23)

PERMISSIBLE 10 (5.5,6.51 OPERATING REGION i s' ,3.1 ' ) ' ' ' ' ' ' ' ' ' ' ' '

0 0 20 40 60 80 100 120 140 160 180 200 220 240 260 280 300 0 20 40 60 80 100 e i e i i i Grouo 7 0 20 40 60 80 100 i i e i e i Group 6 0 2,0 4,0 60 ,80 igg Rod Index, % WD Group 5 8- 22

Figure 8-12. Operation Froni 27 +10/-0 to Rod PositionEFPD 360 +50/-10 Setpoints

-- ANO-1 for Two-Pump Cycle 9 (Tech Spec Figure 3.5.2-38) 110

100 -

l l 90 80 h 70 -

8 g 60 -

SHUTDOWN

• MARGIN (269.3,52) -

) 50

- LIMIT (162.4,52) -

(300,52)

(

2 40 -

OPERATION IN THIS REGION IS NOT (264.0.44 E OPERATION (244.0,38)

ALLOWED RESTRICTED 30 -

' (198.0,23) 20 -

PERMISSIBLE OPERATING .

10 -( 0,3 .1 REGION l' (31.5,6.5)

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

0 20 40 60 80 100 l i i i i i l l Group 7 g 2,0 4p 80 1p0 90 Group 6 0 20 40 60 80 100 Rod Index, % WD 1 i i e 1 l .

Group 5 i

l 8 23 l l

Figure 8-13. Rod Position Setpoints for Two-Pump Operation Af ter 360

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

90 80 -

5 x 70 -

60 -

SHUT 00WN

" (162.4,52) y (300,52)

• 50 _

(269.3.52) -

(264.0,441 2 40 -

OPERATION IN THIS OPERATION (244.0,38)

REGION IS NOT RESTRICTED 30 - ALLOWED (79.5,24)

~

(198.0.23) 20 PERMISSIBLE 10 OPERATING ud.5,6.5)

REGION 0

0 20 40 60 80 100 120 140 160 180 20C 420 240 260 280 300 0 20 40 60 80 100 I I I I i l Group 7

]

? 2,0 9 p 8p 19 Group 6 0 20 40 60 80 100 Rod Index, % WD  ;

I I I I I i .

Group 5 I

8-24

- . . - . . - . . - - - _ .m-l l M l

c" C

1 O

CO e

w 4

l l NO NV i n O

+ -5 e

U1 - u e

Pt2 WW O O m O rt

. M Un ^ N*

w m -4 e s O e CD DN N ^ O3 I

A - - -- W s ^ @

O m OO

• w t m*

O 2 --4 N to w m v m 03 - A "U u O - (D

• OO 6 CO U1 AD K O O , m w -

e - . 5 O N **

- O- ww

=a O

> e PERMISSIBLE OPERATIt;G REGI0fl -

N cr N O. =

• x. O w-N $17 Os U1 "

e ao U w - xo O O O@

r om o,

Power, % of 2568 MWt m i . . . . . . . . . . my

• -* O 8 s s a s a a 3 s a  : V u O

^

tr +-* N w a un Ch M 00 4.0

• w a-

@ O O O O O O O O O C w ~4 3

" ~ C O O r+

Dr O =

.7 n w 3

o  %

G .MO

- ^ Vm 5 f3 ._

^ w kt O e* Q O O m . V

• N Mm C ^ w u) *5 w ^ M N A - to Oe O ~

N Mm W - w C r+

. Ch Mm - LO O 7 4

• O --4 W N N Oo O -% W v v s b b O *-* - M O - - 2l: O Co .

m a rt u O un 5 O m w

• O v O NB9 b

U1 O ~ >Q

• c+

O

-w- ->-- n a_. _,

Figure 8-15. Operational Power Imbalance Setpoints for Operation From 27 +10/-0.to 360 +50/-10 EFPD -- ANO-l' Cycle 9 (Tech Spec Figure 3.5.2-48) l l

. 110. , j

(-20.21.102) 7 (21.58,102)

-100

(-20.83,92) , _gg 'y(21.74,92)

(-28.95,80) - 80 f(23.41,80) a RESTRICTED g--70 RESTRICTED REG 10t1 m REGION g $--60 i

G t

(-31.06,50) O E a. -50 6-(26.29,50)

E J C l' --40 E e t

o --30

- -20 E

g -

-10 t I  ! I l_ _ t g g g 3 40 20 -10 0 10- 20 30 .40 50 Axial Power Imbalance, %

f 8- 26

3

-Figure.8-16. Operational Power Imbalance Setpoints for Operation After a 360 +50/-10 EFPD -- ANO-1 Cycle 9 (Tech Spec Figure 3.S.2-4C)-

t 1

l i

-- 110

(-20.21,102) '

- - tuu (21.58,102)

(-25,62,92) -

- 90 (25.03,92)

(-28.95,80) -- 80 (27.54,80)

" 70 RESTRICTED RESTRICTED REGION $ REGION i

= 2- 60 ,

5

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

E J P 5' -- 40

$ E W

o .

- 30 "i

S y - 20 a

g - - 10 ,

I e I l I i i i i 1 40 20 -10 0 10 20 30 40 50 Axial Power Imbalance, %

i i 1 i 8-27 i

Figure 9-17. LOCA Lirnited Maximum Allowable Linear Heat Rate -

Ad0-1 -- Cycle 9 (Tech Spec Figure 3.5.2-5)

F 1

20 , , , , , , , , , , ,

, 18 .

t 5 _

v /

E 16 -

,/

[s/ -

2 /

1 o 14 - / -

Q g - '

5 ^

4 x 12 -

- - 1000 mwd /mtU AFTER 1000 mwd /mtV _

10 ' I i i i i i 1 e e i 0 2 4 6 3 10 12 Axial Location From Slottom of Core, f t.

8-28

-i

5.3 REACIOR Soecification 5.3.1 Reactor Core 5.3.1.1 'Ihe reactor core contains approximately 93 metric tons of slightly enriched uranium dioxide pellets. 'Ihe pellets are encapsulated in Zircaloy-4 tubing to form fuel rods. 'Ibe reactor oore is made up assemblies. Each fuel assembly is fabricated with 208 of 177 fue:(1,2) fuel rods. Starting with Batch 11, a reconstitutable fuel assembly design is inplemented. 'Ihis design allowc the replacement of up to 208 fuel rods in the a=ambly, 5.3.1.2 'Ibe reactor core approximates a right circular cylinder with an equivalent diameter of 128.9 inches ard a height of 144 inches.

'Ibe active fuel length is approximately 142 inches. (2) 5.3.1.3 'Ibe average enrichment of the initial core is a ntninal L62 we.ight per nt of 23N. 'Ihree fuel enrichments are used in the initial core. 'Ihe highest enrichment is less than 3.5 weight percent 23%, ,

5.3.1.4 'Ihere are 60 full-length control rod assemblies (GA) and 8 axial pwer shapiry rod amamblies (APSRA) distributed in the reactor cure as shown in FSAR Figure 3-60. Each full-length GA contains a 134-inch length of silver-indium-cadmium alloy clad with stainless steel. Each APSRA contains a 63-inch length of Inoonel-600 alloy clad with stainless steel.(3) J l

5.3.1.5 'Ibe initial core has 68 burnable poison spider a=amblies with similar dimensions as the full-length control rods. 'Ihe clackling is Zircaloy-4 filled with alunina-boron and placed in the core as shown in FSAR Figure 3-2.

5.3.1.6 Reload fuel assemblies and rods shall confonn to the design ard  !

evaluation described in FSAR and shall not exceed an enrichment of '

3.5 percent of 23N, l 5.3.2 Reactor Coolant Systen 1 1 5.3.2.1 'Ibe reactor coolant system is designed ard w wtructed in l accordance with code requirements. (4) 5.3.2.2 'Ibe reactor coolant system and any wisected auxiliary systens exposed to the reactor coolant conditions of tenperature and pressure, are designed for a pressure of 2500 psig and a  ;

tenperature of 660 F. 'Ibe pressurizer and pressurizer surge line l are designed for a tenperature of 670 F.(5) 5.3.2.3 'Ibe reactor coolant system volume is less than 12,200 cubic feet. )

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9. STARIUP PRDGRAM - HIYSICS TESTDG me planned startup test prcgram associated with core perforrarye is outlined belcw. mese tests verify that oore perfornance is within the assumptions of the safety analysis and provide informtion for continued safe operation of the unit.

9.1. Precritical Tests 9.1.1. Control Rcd Trio 'Ibst Procritical control red drop times are recorded for all control rods at hot full-flcu corditions before zero pcuer physics testing bcgins. Acceptance criteria state that the rod drop time frca fully withdrawn to 75% inserted shall be less than 1.66 secords at the corditions above.

It shculd be noted that safety analysis calculations are bascd on a rod drop frcra fully withirawn to two-thirds inserted. Since the mcst accurate position indication is obtained frcra the zone reference switch at the 75%-inserted position, this position is used instead of the two-thirds inserted position for data gathering.

9.1.2. PC Flcv Reactor coolant flcw with four RC pumps running will be measured at hot shutdcun conditions. Acceptance criteria require that the reasured flcu be within allcMable limits.

9.2. Zero Power M1vsics Tests 9.2.1. Critical Boron Concentration once initial criticality is achieved, equilibrium boron is obtained ard the critical boron conocatration dete2 rained. We critical boren concentration is calculatcd by correctire for any rcd withdrawal rtquired to achieve equilibrium boron, he acceptance criterion placed on critical boron corantration is that the actual boren conmntration rust be within 100 prn baron of the predictcd value.

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9.2.2. 74!m erature Reactivity Coefficient

'Ibe isothermal HZP tenperature coefficient is measured at approxirately the all-rods-cut configuration. During charges in tenparature, reactivity faarTMck may be va p s.ated by control rod movement. The charge in reactivity is then calculated by the sumation of reactivity (obtained frun a reactivity calculator strip chart recorder) associated with the

temperature change. Acceptance criteria state that the measured value shall not differ fran the predicted value by more than 0.4x10-4 A)y)y0F The moderator coefficient of reactivity is calculated in conjunction with the temperature coefficient measurement. After the temperature coefficient has been measured, a predicted value of the fuel Doppler coefficient of reactivity is added to obta n the moderator coefficient. This value must not be in excess of tho acceptance criteria limit of +0.5x10-4 a)y)y0F l 9.2.3. Control Rod Grouo/ Boron Reactivity Worth centrol rod group reactivity worths (groups 5, 6, arri 7) are measured at hot zero power conditions using the bororVred swap method. This technique i

consi:sts of establishirg a deboration rate in the reactor coolant systen and ocmpensating for the reactivity changes fran this deboration by inserting control rod groups 7, 6, and 5 in ir umental steps. The reactivity charges 1 that occur during these reasurements are calculated Maai on reactimeter data, and differential rod worths are obtained fran the measured reactivity worth versus the charge in rod group position. The differential rod worths of each of the controlling groups are then sumed to obtain integral rod group worths. The acceptance criteria for the control bank group Worths are as follows:

1. Individual bank 5, 6, 7 worth:

1 credicted value - measured value x 100 $ 15

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

Dredicted value - reasured value x 100 $ 10 The boron reactivity worth (differential boron worth) is measured by dividivy the total inserted red worth by the boren change made for the red worth test.

l 'Ibe acceptance criterion for measured differential boron worth is as follows:

9-2 -

-,--- -- - - , , , n.n.- , - - - . - - - - - ,- - ------,-----v.,n---, - ,~.-e . - - - - - , . -

1. credicted value - measured value yg x 100 $ 15 The predicted red worths and differential boren worth are taken frun the PIM.

9.3. Power Escalation 7bsts 9_. 3 .1. Core Svmetry Test The purpose of this test is to evaluate the synretry of the core at low power during the initial power escalation followirq a refueling. Symetry evaluation is based on incere quadrant power tilts during escalation to the intermediate power level. The core symetry is acceptable if the absolute values of the quadrant power tilts are less than the error adjusted alarm limit.

9.3.2. Core Power Distribution Verification at Intermediate Power Invel (IPL) ard 100% FP With Noninal Control Rod Position Core power distribution tests are performed at the IPL and 100% full power (FP). Equilibrium xenon is established prior to both the IPL ard 100% FP tests. The test at the IPL is essentially a check on p wer distriktion in the core to identify any abnormalities before escalatirg to the 100% FP plateau. Peaking factor criteria are applied to the IPL core power distriMtion results to determine if additional tests or analyses are required prior to 100% FP operation.

The following acceptance criteria are placed on the IPL and 100% FP tests:

1. The worst-case maxinzm um nust be less than the IICA limit.

2. The mininra DE nust be greater than the initial condition DER 11mit.

3. The value obtained fran extrapolation of the mininnn D E to the next power plateau overpower trip setpoint uust be greater than the initial condition DE limit or the extrapolated value of imbalance must fall outside the RPS power / imbalance / flow trip envelope. l

4. The value obtained from extrapolation of the worst-case maximum Um to the next power plateau nyerpcwer trip setpoint nust be less than the fuel melt limit, or the extrapolatal value of imbalance must fall l outside the RPS power / imbalance / flow trip envelope.

5. 'Iha quadrant power tilt shall not exceed tha limits specified in the Technical Specifications.

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6. The highest measured ard predicted radial peaks shall be within the folloaing limits:

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

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

pig 31cted value - measured value x 100 more positive than -7.5 measured value I't. ems 1, 2 and 5 ensure that the safety limits are maintained at the IPL ard 100% FP.

Items 3 and 4 establish the criteria whemby escalation to full power may be acomplished without the potential for e>'rwrHng the safety limits at the overpower trip setpoint with regard to DiBR and linear heat rate.

Items 6 and 7 are established to determine if measumd and predicted core power distributions are consistent.

9.3.3. Incore Vs. Excore Detector Imbalance Correlation Verification at the IPL Inbalances, set up in the core by control rod positioning, are read simultaneously on the incere detectors and exoore power range detectors. The excore detector offset versus incore detector offset slope mst be greater than 0.96. If this criterion is not met, gain arplifiers on the excore detector signal processing equipnent are adjusted to provide the required 9&$D 9.3.4. 'Qlroerature Reactivity Coefficient at m100% FP 1

The average reactor coolant temperature is decreased ard then increased by about SoF at constant reactor power. The reactivity associated with each towture change is obtained frun the change in the controllirq rod group position. Ctritrolling rod group worth is measured by the fast insert /witMraw methcd. The tenperature reactivity coefficient is calculated frun the measured changes in reactivity and temperature.

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

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9.3.5. Power Doroler Reactivity Coefficient at N100% FP

'Ihe power Dcppler reactivity coefficient is calculated frcan data recorded during control rod worth measurenents at power using the fast insert / withdraw method.

'Ihe fuel Doppler reactivity ocefficient is calculated in conjunction with the power Doppler coefficient measurement. 'Ihe power Doppler crefficient as measured above is nultiplied by a precalculated conversion factor to obtain the fuel Doppler coefficient. 'Ihis measured fuel Doppler coefficient nust be more negative than the acceptance criteria limit of -0.90 x 10-5 gg, 9.4. Procedure for Use if Acceptance Criteria Not Met I If acceptance criteria for any test are not met, an evaluation is performed before the test program is continued. Further specific actions depend on evaluation results. 'Ibese actions can include repeating the tests with more detailed attention to test prerequisites, added tests to search for ancrnalies, or design personnel performing detailed analyses of potential safety problems because of parameter deviation. Power is not escalated until evaluation shows that plant safety will not be c.4+v- M by such escalation.

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10. REFERENCES

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

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2. BPRA Retainer Design Report, BAW-1496, MWk& Wilcox, Lynchburg,  ;

Virginia, May 1978.

3. J. H. Taylor (B&W) to S. A. Varga (NRC), letter, "BRA Retainer Reinsertion," January 14, 1980.

4. Rancho Seco Cycle 7 Reload Report - Volume 1 - Mark BZ Fuel Assably Design Report, B W 1781P, MWk & Wilcox, Lynchburg, Virginia, April r 1983. i

5. Rancho Seco Nuclear Generating Station - Evaluation of Mark BZ Fuel Assertbly Design, U.S. Nuclear Regulatory Ctumission, Washington, D.C. ,

Novenbar 16, 1984.

6. Gadolinia-Bearing Imad '1%st Assettblies Design Report, BAW-1772-P, NWk & Wilcox, Lynchburg, Virginia, June 1983.

7. Frugaru to Detarmine In-Reactor Perfortnance of B&W Fuels - ClaMing Creep Collapse, BAW-10084P-A. Rev. 2, hWk& Wilcox, Lynchburg, Virginia, October 1978.

8. Y. H. Hsii, et al. , TACD2-Fuel Pin Perfortnance Analysis, BAW-10141P-A, Rev.1, hWk & Wilcox, Dfnchburg, Virginia, June 1983.

9. N00 DIE - A Multi-Dimensional 'NcK3roup Reactor Silm11ator, BMf-101525, hW4 & Wilcox, Lynchburg, Virgini3., June 1985.

10. Arkansas Nuclear One Unit 1, Cycle 8 Reload Report, BAW-1913, h W k &

Wilcox, Dfnchburg, Virginia, Noverrber 1986.

11. J. H. Jones, et al., LYNXT - Core Transient '1hermal-Hydraulic Program, BAW-10156A, hWX & Wilcox, Dfnchburg, Virginia, February 1986.

12. R. L. Harne and J. H. Jones, 'Iherinal-Hydraulic Crossflow Applications, BAW-1829, hWk & Wilcox, Lynchburg, Virginia, April 1984.

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13. BWC CLrrelation of Critical Heat Flux, BAW-10143FA, mWk & Wilcox, Dfnchburg, Virginia, April 1985.

14. Fuel Rod Bowing in MWk & Wilcox Fuel Designs, BAW-10147P-A, Rev.1, mWk & Wilcox, Lynchburg, Virginia, May 1983.

15. Arkansas Nuclear One, Unit 1 - Fuel Densification Report, BAW-1391, Bal.cxk & Wilcox, Dfnchburg, Virginia, June 1973.

16. ECCS Analysis of B&W's 177-FA Iowered-Irep NSS, BAW-10103A, Rev. 3, m W k & Wilcox, Lynchburg, Virginia, July 1977.

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

Wilcox, Lynchburg, Virginia, February 1985.

18. Bounding Analytical Assessment of NUREG-0630 Models on IDCA kW/ft Limits With Use of FIECSE'r, BAW-1915P, mWk & Wilcox, Lynchburg, Virginia, May 1986.

19. Normal Operating Controls, BAW-10122A, Rev. 1, mWk & Wilcox, Dfnchburg, Virginia, May 1984.

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