Text

Reload Rept Catawba Unit 1,Cycle 6
	ML20066E070
Person / Time
Site:	Catawba
Issue date:	10/31/1990
From:	; BABCOCK & WILCOX CO.
To:	;
Shared Package
ML20066E065	List: ML20066E063; ML20066E067; ML20066E070;
References
	BAW-2119, NUDOCS 9101170268
	Download: ML20066E070 (136)
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'N j . Contents Page

1. INIBODUCTION AND SUPEARY . . . . . . . . . . . . . . . . . . . . . 1-1

2. OPEPATDC HISIORY ........................ 2-1

3. GENDAL DESCRI1 TION ....................... 3-1

4. FUEL SYSTEM DESI*.N . . ...................... 4-1 4.1. Fuel Assembly Mechanical Design . . . . . . . . . . . . . . 4-1 4.2. Fuel Ibd Design . . . . . . . . . . . . . . . . ...... 4-2 4.2.1. Ebel Id ClaMiry Collapse ............ 4-2 4.2.2. Ebel Rod ClaMirg Stress ............. 4-2 4.2.3. Fuel Rod ClaMirg Strain ............. 4-3 4.2.4. Ebel Rod Cladding Fatiguo . . . . . . . . . . . . . 4-3 4.3. h W Design ...................... 4-4 4.4. Material Ccupatibility .................. .

4-4 4.5. Operating Experience ...,,.............. 4-4

5. NUCIEAR DESIGN ......................... 5-1

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k / 5.1. Ehysics Chararfieristics . . . . . . . . .......... 5-1

'd 5.2. Charges in Nuclear Design . . . . . . . . . . . . . . . . . 5-1

6. 'IEERMAIrHYIEAULIC DESICM . . . . . . ............... 6-1

7. ACCIDENT AND TPANSIENF ANALYSIS ................. 71 7.1 General Safety Analysis . . . . . . . . . ........ 7-1 7.1.1. Steam Line Break Analysis . . . . . . . ...... 7-1 7.1.2. Med Ejection ................... 7-2 7.1.3. RCCA Miscocration . . . . . . . . . . . ...... 7-2 7.1.4. Locked Rotor ................... 7-2 7.2 ECCS Analysis . . . . . . . . . . . . . . . . . . . . . . . 7-3 7.3 Radiolcgical Consequences . . . . . . . . . . . . . . . . . 7-4 7.3.1 Loss of Coolant Accidents . . . . . . . . . . . . . 7-4 7.3.2 Ircked Ratcr Accident . . ............. 7-6 7.3.3 Single RCCA Withdrawal at Power . . . . . . . . . . 7-6

8. PROFOSED lODIFICATIONS 'IO TEQiNICAL SPECIFICATIONS AND CDIR ... 8-1 Chargas to Technical Spectfications ............... 8-5 Charge.s to Core Operating Limits Report ............. 8-73 Charges to Final Safety Analysis Report ............. 8-75 A 9. STARIUP PHYSICS TESTDC ..................... 9-1

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List of TablCD Table Page 4-1. Ibel Design Parameters ard Dirensions . . . . . ... . ..... 4-5 5-1. niysics Paranaters, Catawba 1 Cycles 5 aryd 6 . . .. . . ..... 5-3 5-2. Shutdown Ihrgin Calculation for Catakte Cyclo 6 . .. . ..... 5-5 6-1.11caninal 'Ihernal-Ilydraulic Design Conditions, Cycle 6 . .. .... 6-2 6-2. Statistical Core Design Application Sunmary Measuremnt thcertainties ................. .. 6-3 6-3. Statistical core Design Application Summary . . ... . ..... 6-4 7-1. Safety Aralysis Chocklist for M1ysics Data . . . . . . . . . . . . 7-7 7-8 7-2. Rod doction Paramters ..................... 7-9 7-3. BL11oltgical Consecperces Doce Results Summary . . . . . . . . . .

8-1. Technical Specifications Chargos . . . . . . . . . . . . . . . . . 8-3 8-2. Core Operating Limits Report Changes . . . . . . ... . ..... 8-4 List of Fiaures Figure 3-1. Core Loadirg Diagram for Catawba Unit 1 Cycle 6 ... . ..... 3-2 3-2. Enrichment and DOC Durnup Distribution for Catawba Unit 1 Cycle 6 ......................... 3-3 3-3. Rod Cluster Control Assembly Isations ard Bank Designations for Catawba Unit 1 Cycle 6 . . . . . . . . ..... 3-4 3-4. Durnablo Poison Pin Distribution for Catawba Unit 1 Cycle 6 ... 3-5 4-1. Mark-DW 17 Fuel Assembly . . . . . . . . . . . . . . . . . . . . . 4-6 5-1. DOC (4 EFFD), Cyclo 6 ho-Dimensional Relative Power Distribution - HFP, Equilibrium Xenon .. . . .... . ..... 5-6 7-1. Doppler Power Coefficient . . . .... .. . . .. . . ..... 7-11 7-2. Scram Curve . . ......................... 7-12 7-3 Catawba 1 Cycle 6 Dropped Rod Peaking . . . . ... . . ..... 7-13 0

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1. INITrOUK. TION AND Sutt%RY tis report justifies the operation of the sixth cycle of catasta Nuclear Station, Unit 1 at the rated com pcwer level of 3411 !&t. Included are the required analyses as outlined in the USNRC ^~wnt "Guidarce for Proposed License Amendments Relating to Refueling," July 1975..

Cycle 6 for Catauba Unit 1 will be the first cycle for which the reload fuel is supplied by B&W Fuel Ctopany (IMFC) and will therefore be the reference cycle for IMFC fuel in the Catawba units. Se inocaning batch 8 fuel assemblies are designated as Mark-IM. To support implementation of Mark-IM fuel in the Catawba and M:Guire nuclear plants, EMFC has developed new methods ard mdels to analyze the plants during normal and off-normal operation. Wese methods and models are documnted in topical reports ard have been reviewed by the NRC. Most of the topical reports have already been approved. Approval of the final four tcpical

/ reports is scheduled for ccarpletion by n=har 14, 1990.

('v ]) - Section 2 of this report is the operating history for fuel in Catakta Unit 1.

Section 3' is a general description of the reactor core, and the fuel system design is provided in Secticri 4. - Reactor and system parameters and conditions are sumarized in Sections 5, 6, ard 7. Changes to the Technical Specifications, Care Operatirg Limits Report (CDIR) , and Final Safety Analyis Report are provided in Section 8. We secpe of Riysics Startup Testing for Catakta Unit 1, Cycle 6 is provided in Section 9.

All of the accidents analyzed in the FSARI have been revicued for Cycle 6 operation. In those cases where Cycle 6 characteristics were conservative compared to those analyzed for previous cycles, new analyses were not performed.

Several bourding transients were analyzed in detail to demonstrate the capability of IMFC calculational techniques. W e results of these analyses were reported in BAW-10173P.*

On May 17, 1990 the NRC issued Amerdment Number 74 and Amenimnt Number 68 to the Catawba Nuclear Station Facility Operating License. %ese amerdments allow

,m the removal of cycle-specific core parameter limits frcan %chnical Specifications I1 I 1-1 v/

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ard require that these limits be ircitrkd in a Cbre Operating Limits Report (COIR) . 'Ibe Core Operating Limits Report is subnittal to the IGC upon issuance ard does not require approval prior to implementation. 01arges to the operating limits are nade via the Core Operatirg Limits Report.

The 'Itchnical Spocifications have been reviewed, ard the nodifications for Cycle 6 are justified in this report. Based on the analyses performed, Wich talm into-account the postulated effects of fuel densification ard the Firal Icoeptarce Criteria for emergency core coolirg (DI), it has been corcluded that Catawb3 Unit 1 Cycle 6 can be safely operated at a core power level of 34111&t.

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2. OPERATDIG RIFIORY

'Ihe current operatirg cycle for Catawba Unit 1 is Cycle 5 which achieved criticality on April 22, 1990 ard reachcd 100% full power on 7pril 29, 1990.

Cycle S is scheduled to shut dcun in Mur:h 1991 after 300 EFPD. 'Ihis-cycle ard all previous cycles have operated with fuel assablies of the Westinghouse design.

Cycle 6 will be the new reference cycle - and will be the first fuel cycle containing NFC Mark-w fuel assemblies (FAs) . It is scheduled to. start up in June _1991 at-a rated power level of 3411 MRt and has a design cycle length of 350 EFPD. No operating anmalies have occurred durirg previous cycle operations that would adversely affect fuel performance in Cycle 6.

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3. GDERAb DESCRIPITON .

f We Catawba Unit i reactor core is described in detail in chapter 4 of the ITJd.

The core consists of 193 fuel assemblies, each of which is a 17-by-17 array containirg 264 fuel Irds, 24 guide tubes, ard one incore instnennt guide tube.

%e 121 burncd FAs are of the Westirghouse Optimized Fuel Assembly (OFA) design, and the 72 fresh FAs are of the Mark-N design *; all the fuel rods have Zircaloy-

.4 cladding. We fuel rod outsido diameters are 0.360 ard 0.374 inch, and the

-wall thicknesses - are 0.0225 ard 0.024 inch for the OFA ard Mark-N designs, respectively. We Mark-N fuel consists of dished-end, cylirdrical pellets of uranium dioxide (see Tabic 4 for data) . The average naminal fuel loadings are 423.119, 424.898, 423.119, ard 456.300 kg of uranium per fuel assembly in batches 3D, 68, 7, ard 8 respectively. Figure 3-1 is the core loading diagram for Cycle 6 of Catawba Unit 1. - The initial enrichments of batches 3D, _6B, and ? were 3.10,

.b . 3.279', ard 3.40 wt % 23 5 ) , respectively. The design enrichment of fresh batch 8

! is 3.55 wt % *hJ.

-ihe fifty-two batch 6B ard sixty-eight batch 7 assemblies will be shuffled to new locations. One batch 3D FA discharged at the end of Cycle 2 will be re-inserted as the center assembly. The seventy-two fresh batch 8 assemblies will -

-be loaded into the core in a symetric checkerboard pattem. Figure _3-2 is an

-eighth-core map showing the burnup and initial enrichment of each assembly at the beginnirg of Cycle.6.

Cycle 6 will be operated in a feed-ard-bleed mode. Core reactivity is controlled by 53 rod cluster control assemblies (RCCAs), 52 BmAs, ard soluble boron shim ~

1he Cycle 6 locations of the 53 rod cluster control' assemblies with their respective designations are irdicated-in Figure 3-3. The Cycle 6 locations and number _of pins -of 3.0 wt % BgC per BWA cluster are shown in Figure 3-4.

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4 Figure 3-2. Enrichment and BDC hirnup Distribution for Catawba

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-H G F- E D C B A 3.10 '3.40 3.40 3.55 3.40 3.40 3.40 3.55 lt 8

27,098- 16,567 16,660 0 16,3EB 10,434 16,700 0 i i

3.40 3.55 3.279 3.55 3.279 3.55 3.40 i 9 i 16,362 0 27,965 0- 24,997 0 13,453

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3.40 3.55 3.279 3.55 3.40-11 16,639 'O 23,089 0 14,895-

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l. X.XX Enrichment, Initial L XX,XXX Burmp, - (mwd /mtU) , . BOC 1 .

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Figure 3-3. rcd Clir@.r Control Anscobly I.ccations ard hut.

Designations for Catawba Unit 1 Cycle 6 1... . ........

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Figure 3-4. Durreble Poison Pin Distrih2 tion for Catawba Unit 1 Cycle 6

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4. IUEL SYSTEM DESIGN 4.1. Fuel Aerenbly Mechanical Desian he Catawba 1 Cycle 6 batch 8 feed ccrprises 72 Mark-N fuel assemblies with ari enrichment of 3.55 wt% 2hJ. A total of 52 Mark-N BPRA's are used with either 4, 8 or 12 DP pins each. Se remainder of the fuel assenblies in Cycle 6 are Westirghouse Optimized Fuel Assemblies. We Mark-N fuel asceably is a 17x17, u W lattice, Zirraloy spacer grid fuel assembly designed for use in Westirghouse designed reactors. We fuel assembly incorporates many standard SE design features while maintainirg orpatibility with the Westirghouse reactor internals and resident fuel assemblies. Se nozzles, nozzle attachment and end spacer grids are based on proven Nuclear Fuel Irdustries (NFI) designs currently in operation in Westinghouse-designed reactors in Japan. Se guide ex thimble tcp section, dashpot diameters, instrument sheath diameter, and the fuel -

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( ) rod outside diameter are the same as the stardard 17x17 Westirghouse design.

me fuel red design has been developed based on standard SE methods applied to the Westinghouse standard outside claddire diameter. S e unique features of the Mrk-N fuel assembly design include the Zircaloy intermediate spacer grid, the spacer grid restraint system, and the use of Zircaloy spacer grids with the standard lattice design.

Se fuel assembly, shwn in Figure 4-1, consists of 264 fuel rods, 24 guide thimbles, and one instrumnt sheath in a 17x17 square array. We guide thimb]es provide guidance for RCCA insertion and are attached to nozzles at the top and bottcm of the fuel assembly and to the bottcn end spacer grid $> form the structural skeleton. A reduced diameter section at the bottcm c: the guide thimbles decelerates the RCCA during trips. W e instrument sheath occupies the center lattice position and provides guidance ard protection for the incere instrumentation assemblies. We top nozzle assembly cx>ntains the fuel assembly holddown sprirgs and is attached to the guide thimbles by remwable nuts ard lockiry cups We bottom nozzle is attached to the guide thimbles by bolts which

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are mchanically capturtd by tack solding. me bottcn nozzle utilizes a unique flow hole pattern consisting of sml1 rourd holes and a cloverleaf shapcd hole to prevent passaga of debris which could lead to fuel rod failure due to fretting. We fuel rcd and guide thimble spacirg are maintained along the length of the assembly by six Zircaloy intemodiate spacer grids.

4.2 Fbel Rod Daian Analyses were performd on the Mark-N fuel rod design to assure that its mechanical perform 1rce in-reactor would be adequate; the methods are described in Reference 3. %e areas that were analyzed are:

A. Creep Collapse B. Claddirg Stress C. Cladding Strain D. Cladding Fatigue 4.2.1 Fuel Rod Claddina CollaDse te fuel rods were analyzed for creep collapse using nothods outlined in Reference 3 and the creep collapse code 0 c#. Usirg nuclear design inputs, a power history was determined wttich enveloped all past fuel red operating ccoditions for the Catawba Plant. mis power history with appropriate uncertainty factors was input into the camputer code TACO 52 ktlich detemined the temperature, the pressure, and the fast neutron flux level history of the Mark-N fuel reds. mis history was input to CRCV usirg conservative claddirq dimensions. Frcn the output of C30V the creep collapse point of the Mark-N fuel rods was determined to be greater than 60,000 mwd /mtU. m is burnup exceeds the naximum burnup and exposure the Mirk-N fuel rois are expected to experien:e in Catawba 1, Cycle 6.

4.2.2 Fuel Rod Clpddina Stress te fuel red cladding was analyzed for the stresses irduced durirg Condition I ard II operation. We ASME pressure vessel stress intensity limits were used as guidelines. Conservative valucs were used for cladding thickness, oxide layer buildup, external pressure, internal fuel rod pressure, differential tempnture and unirradiated cladding yield strength. We analysis results show that the 4-2 B&W Fuel Company

jm mximum claddirg stress intensities are within limits under all Icedirg

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Stress Analysis Summary Condition Limit S Margin,%

Pm Sm (30,000 psi) 22,926 psi 30.9 PWFb 1.5Sm (45,000 psi) 26,272 psi 71.3 PMIt+Pl 1.SSm (45,000 psi) 27,888 psi 61.4 PW Pb+P14Q 3.0Sm (90,000 psi) 65,157 psi 38.2 Where:

Pm = general prinny membrane stress intensity, Pb = primary bendirg stress intensity, Pl = local primary membrane stress intensity, Q = secondary stress, Sm = allowable mambrane stress = 2/3 Sy or 1/3 Su (whichever is less),

g Sy = yield stress, ard Su = ultimate stress.

4.2.3 Fuel Red C1addirn Strain 1he fuel rod was analyzed to determine the maximum allcvable local power change that the fuel rod could experience before the very conservative transient strain limit of 1% 'is exceeded. The transler.t strain limit is based on diametral claddirq strain resulting from a local pcuar transient. The maximum. allowable local pcwor charge determined from the analysis was compared to the maximum calculated local power charrje induced by a worst case core maneuvering scenario.'

This comparison determined that margin exists to the 1% strain limit.

L 2.4 - Fuel Rod C1addirn FaticTue the fuel rod was analyzed for the total fatigue usage factor tisirg the ASME pressure vessel code as a guideline. A maximum fatigue usage factor of 0.9 is allowed. A fuel rod life of 8 years was assumed. All possible Condition I &

II events expected and one Condition III event were analyzed to determine the total fatigue usage factor experienced by the fuel rod cladding. Conservative inputs in terrs of claddirq thickness, oxide layer buildup, external pressure, 4-3 B&W FuelCompany

l intomal fuel rod pressure and differential teqwrature across the cladding were asstmod. The results of the fatigue analysis shw a raximum fatigue usage factor of 0.35.

L3 7hermal Derdgp 7he therml performnce of the fresh batch 8 Park-BY asserblies was evaluated with the TACD3 code as described in DAW-10162P-A. Nomiml undensified input 6

inrameters used in the analysis are presentcd in Tablo 4-1. Densification offects were accounted for in TACD3.

The results of the therml design evaluation of the Mark-Bf fuel are summarized in Table 4-1. Cycle 6 core protection limits for the Park-EW fuel are based on a linear heat rate (1HR) to centerline fuel molt limit of 21.86 MV/ft as determined by the TACO 3 code.

The mxinn fuel assembly bumup at EOC 6 is predicted to be less than 18000 M4d/mtU for the Mark-BY fuel (batch 8) . 7he fuel rod internal pressures have been evaluated with TACD3 for the highest burnup rods and are predicted to be less than the nomin:tl reactor coolant pressure of 2280 psia.

4.4 Material Comatibility The compatibility of all possible fuel-cladding-coolant-assembly interactions for batch 8 feed fuel assemblics is identical to that of present fuel assemblies. -

4.5 Coeratim B"perienge Catawba Unit 1 Cycle 6 is the first complete reload batch of Mark-BY 17x17 fuel asserblies. BiFC cxperience with fuel assenbly irrad.iation extends to 58.3 Gid/mtU for the Park-B 15x15 product line with over 5,000 assemblies irradiated.

D:perience with the Mark-BY 17x17 fuel assemblies for Westinghouse designed reactors consists of two sets of lead asserblics in two reactors. This experience started with the irradiation of four lead assorblies in McGuire Unit L 1 rg.ie 5. 'ihe i4ct.nire lead assemblies cartuiilly ili uie third cycle of irradiation with an anticipated assembly bumup of 42.8 GWd/mtU by the end of Cycle 7. The M::Guire lead assemblics were examined poolside after Cycle 5 and Cycle 6. The last post irradiation examimtion on the McGuire lead assemblies was done after Cycle 6 at a burnup of 27.6 CMd/mtU. Fuel assorbly bw, twist,

-grwth ard holddown spring set were within nominal bounds. Four other lead asserblies are undergoing their first cycle of irradiation in Trojan Cycle 13.

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Table 4-1. Fuel Desinn Parancters ard Dircrisions - Mark EM Ncaninal Fuel Rod O.D. (in) .374 Ncaninal Fuel Rod I.D. (in) .326 Ncaninal Active Nel Inrgth (in) 144.0 Ncxtinal Fuel Pellet 0.D. (in) .3195 Fuel Fellet Initial Dens.ity (%) 96 Initial Fuel Enrichment twt % % ) 3.55 Average Durnup IOC, (mwd /mtU) O Estinated Residence Tirac DOC, EFP11 6000 C1addity Collapoo Time, EFP.I >30,194 Claddirg Cbilapso Durnup, mwd /mtU >60,000 Ncaninal Linear lleat Rate, ()M/ft) 5.43 Average Fuel Muperature at Ncaninal UIR IDL, (*F) 1230 ninimum um to Molt, ()df/ft) 21.86 1

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5. IUC11AR DESIQ1

11. Ihvnics Omracteristics Tabte 5-1 provides the core physics parameters for Cycles 5 ard 6. The Cycle 6 Values were generated usirg the 1100D2' code (see Section 5.2) and are valid for the design cycle length (350 LTTO) plus 10 EF70. Figure 5-1 illustrates a representative relative power distribution for the beginning of cycle 6 at full power. 7his case was calculatcd by the PCQ07" ccdo (see Section 5.2) as part of the design depletion ard contained equilibrium xenon ard nominal rcd positions.

Durirg verification of the contml rod insertion limits specified in the CDE, calculated ejected red worths and their adherence to acceptance criteria were considercd. The adequacy of the shutdown margin with Cycle 6 stuck rod worths is denonstratcd in Table 5-2. The shutdown margin calculations include allowance (N for potential flux redistribution ard a 10% uncertainty on not rod worth. Flux

( redistribution was accounted for separately since the shutdown analysis was performed usirq a two-dimensioml mcdel. The shutdown calculation at the end of Cycle 6 was analyzed at 360 EFPD.

E.2. Chanags in 11uclear Desian The core design charges for Cycle 6 include the use of Mk-IE fuel assemblies and BPRAs with variable numbers of poison rods as described in Section 3.

Additionally, the Cycle 6 design lifetime of J50 EFPD represents an increase compared to previous cycles.

7hc Cycle 6 physics parameters appearirg in this report were calculated with the PDQ07, FIR 4E3', and 1100DE codes. These ccdes were run in either two or three dimensions deperdirg on the amount of nodelirg detail required and the characteristics of the irdividual codes. The PDQ07 calculations were performed in two dimensions; the FIAME3 calculations were performed in three dimensions, ard the 1100DE calculations were carried out in both two and three dimensic,ns.

The Reactor Protection System (RPS) limits (Technical Specification changes; O\ 5-1

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verified by aralyses for this fuel cycle are presented in Section 8.

Operational limits for the core are provided in the CDIR; revisions to the COIR for Cycle 6 are presented in Section 8.

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Tablo 5-1. Ihysics Parametern(*), Catawba 1 Cycles 5 an:16 Cvele 5 Lycle.6

( Ibsign cycle length, EITD 300 350 Design cycle burnup, mwd /mtU 12,520 14,145 Design average core burnup - DOC, ?U l/mtU 25,729 26,216 Design initial coro loading, nfJJ 81.7 84.1 Critical boron - DDC, ppmb, no Xe*

IIZP, all rods out 1,483 1,699 11ZP, bank D inserted 1,412 1,630 IIFP, all rods out 1,358 1,543 Critical bortn -- DOC, ppnb, eq Xe ICP 263 290 IIFP 15 16 Control rod worths -- DX, IIIT, eq Xe, pcm Bank D 647 639 Dank C 1282 1117 Control rod worths - DDC(*), IIPP, eq Xc, pcm Ihnk D 576 703 lbnkC 1161 1187

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(*) 355 DDC (D12)

(*) 473 EDC'*) (D12)

Max stuck rod worth - IIZP, pcm DOC (F8) 1065 867 EOC(*) (F8) 1173 912 Dwer deficit - IIZP to IIFP, pcm BDC -1716 -1564 DDC(*) ~3058 -2752 Doppler coeff -- HFP, pcm/ F DOC, no Xe -1.42 -1.57 DDC(*), eq Xe -1.71 -1.84 Modcrator coeff -- IIIT, pWF DOC,1670 ppub, no Xe -1.53 -1.70 EOC(') , O pptb, eq Xe -29.38 -32.27 5-3

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Table 5-1. Miysics Parancters), Catawba 1 Cycles 5 and 6 (Continued)

Cycle 5 Cvele 6 Doron worth - IFP, panVppmb DDC -9.06 -7.85 DOC -10.25 -9.40 Xenon worth -- ITP, pcm LOC (4 EFPD) 2728 2665 EOC'*) (equilibrium) 2991 2837 Effective delayod neutron fraction -- IIFP DDC 0.00609 0.00626

) 0.00525 00C(*)

(*) Cycle 6 data are for the conditions stated in this report; the Cycle 5 values given were provided by Duke Icaer Campany fram their analysis.

" IIZP denotes hot zero power (557F T,); IIFP denotes hot full power (593F Tm).

(*) EOC physics parameters calculated at design EOC plus 10 EFPD.

")

Ejected rod worth for banks D, C and B insertcxi.

) These values were not generated by Duko Ioser Company for Cycle 5.

5-4 B&W FuelCompany

Table 5-2. Shutdcun Kirgin Calculation for Catawba 1 Cvele 6

\

Availablo Rod Worth LQC, %Ac DOC'*). %Ao

1. Total rod worth, HZP 6.57 7.05

2. Maximum stuck rod worth, HZP .-0.87 -0.91

3. Not Worth 5.70 6.14

4. loss 10% urcortainty -0.57 -0.61

5. Total available worth 5.13 5.53 Required Rod Worth

6. Power defect, HFP to HZP 1.35 2.06

7. Max allowable inserted rod worth (b) 0.43 0.65

8. Flux redistribution _DJJ 0.70

9. 'Ibtal Inquirad worth 1.99 3.41 7

% 10. Shutdown Mari 9 n (total avail. varth

( minus total required worth) 3.13 2.12 N

EZ3: . Required shutdown margin is 1.30%Ap.

")EOC physitt parameters calculated at 360 EFFD, i.e., desicJn EOC plus 10 EFPD.

(b) Includes allcwance for consideration of RCCA positions as fully withdrawn at 222 steps withdrawn.

5-5 B&W FuelCompany

h Figure 5-1. B3C (4 EFPD), Cycle 6 7VO-Dimensicnil Relative kver Distribution - IEP, Equilibritto Xenon H G F E D C B A 8 0.87 1.12 1.19 1.30 1.20 1.23 1.08 0.88 l

9 1.17 1.29 0.95 1.25 0.99 1.23 0.71 10 0.99 1.21 0.92 1.22 1.06 0.00 11 1.17 1.27 0.97 1.12 0.46 12 1.15 1.17 0.55 13 O

0.69 0.26-14 15 X.XX Relative power density O

5-6 B&W FuelCompany

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6. THERMAlrHYmAULIC DESICH The thermal-hydraulic design evaluation supportirg Cycle 6 operation was performed with the - statistical core design (SCD) analysis method, which incorporates the SW WF oorrelation. The SCD method and SW have been demonstrated to be generically applicable in References 10 and '11, respectively.

Cycle 6 is the first transition cycle to the }hrk-Si design at Catawba 1 and the first cycle to_ utilize SffC's SCD methods. Core safety limits for Cycle 6 are based or, a full core Mark-34 analysis with a 1.55 design 15. In the mixed core,-

those limits have been confirmed to be applicable to the Westin@.ouse OPA with

-a design 6 of 1.'49. Table 6-1 provides a summary of' the thermal-hydraulic design parameters used to evaluate Cycle 6.

The SCD method that was used for this reload evaluation treats un:ertainties in >

( design inputs statistically. By doirs this, a statistical EtHR ' design limit is

\ determined that is greater than the SM GF correlation limit documented in Reference 11. 7b provide additional design flexibility,- a thermal design limit i

is established that incorporates thernal margin. For the Catawba 1 core the statistical design limit-(SDL) has been calculated as 1.345 (SW) based on the plant specific uncertainties listed in Table 6-2. Other -generic and fuel 30 dependent uncertainties are the same as those presented in BAW-10170P-A . For the Catawba 1 Cycle 6 analyses the thermal design limit (TDL) is -1.50 SW.

The thermal margin based on these values is as follows:

1.50 - 1.345 Thermal Margin (%) = X 100 = 10.3%

1.50 Table .6-3 outlines the penalties and offsets that must be assessed -against the thermal margin included .in the TDL.

-( 6-1 m

B&W PJet Company

Table 6-1. Nanimi 'Ihcrml-Hvdraulie Desian corrlitions Cvele 6 Design Power Invel, MRt 3411 Com Dcit Pressure, psia 2280 Harnimi Average Tcr.perature, *F 590.8 Reactor Coolant Systan Flw, gpn 385000 Core Dfpans F1 w , % 7.5 IMR ttdelirg ScD Refercrce Design Ruiial-local Mark-IM 1.55 kwer Peakiry Factor OFA 1.49 Reference Design Axial Flux Shape 1.55 Casine Active Fuel largth, in 144.0 QIF Cormlation EMCMV Statistical Design Limit (SDL) 1.345 Thermal Design Limit (7DL) 1.50 0

6-2 B&W FuelCompany

- - . - . . . - . . - - . . ~ . . . - - .-. -... - . ~ . . . - . . . - . . . . - .

.,O . Table 6-2. Statistical Core Desian Armlication Suttnarv -

-\ > % i.snt Unoartainties ;

-Variable! Name Uncertainty Distrikrtion

-l r

Q Core Power 2% Normal W- Core Flow 2.2% Unifom

-P- Core Pressure 30 psi Unifom

i' T' Cbre Inlet 'Rmperaturu 4*F ' Uniform R Measured FI 5% Normal ligg Analysis Urcertainties Variable Name Unoettainty Distribution

.)

W Core Bypass Flow 1.5% Uniform

(

R 'Ilot Channel Factor.

. 3%(*) Normal N

.~I.

(*)Also applies to the Westinghouse OFA j-o t

6-3 A

B&W FuelCompany

Table 6-3. Statistical Core Desicm Ardigadgn Stumarv Penalties & Offsets to be Assessed Against 7he Thermal Margin Ircluded 1-) the Therml Design Limit I

Statistical Design Limit (SDL) 1.34S Theml Design Liait (7DL) 1.50 Percent Margin Available 10.3 Penaltv/ Offset yAhtg li' irk-IM QEb Transition Core 0% 0% (*)

Rod DN 0% 2.7%

Flw Anxaly *) 3.5% 3'.5%

Instrumentatiorg/Hardwaru 4.2% 2.9%

7btal 7.7% 9.0%

Available IMR Margin 2.6% 1.3t

( OFA evaluations are based on a desity, fa of 1. 49.

) The f1w anomaly penalty is applied to ocnpensate for en armalous f1w c:niition that has been detected in the Catawba units. -The fim anomaly results frun a vortex fermed in the lower reactor vessel internals due to the flw distr.ibution .md internals configuration.

6-4 B&W Fuel Company

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, 7.0 ACCIDDTI' AND TRANSIDU ANALYSIS 1

].1 General Safety Analveig In ' order to detemine the effects of thic reload ard to ensure that the themal performance 'daring hypothetical. Incidents is not degraded, each IEAR accident analysis has been evaluated..

. The - safety l analysis evaluation ws p7esented .in: topical report BAW-10173II, !

Mark-N Reload Safety Analysic for Catawba and McGuire. MW40173P demon,'trates

' thatithe use of Mark-N fuel'in these plants dcas not Itduce the existing safety marrgin. Table c3.4-1 of . MW-1017.1P, entitled Input Parameters and Initial.

Conditions for Transients, presents a comparison of .the values used in the topihal report analysis and in the Catawba ard McNire ISAR analyses, he key parameters that:have the greatect effect on detemining the outcame of q

a transient were determined in Section 6.0 of RW-10173P. Ompa.risons of these- ,

" g;' key parameter values to the parameter values for Cycle 6 are shown in Table}

. i

1- and Figure 7-1.

%e Catawba 11 Cycle' 6 calculated parameters are all within the limiting values

. die'NW tin BAW-10173P.. We cycle specific evaluations of several transients

.are presented lin the following. sections.

i 17.1.1- Steam Line Break Analg ig-EAW-10173P provided a steam line break transient analysis-for the offsite pomr.- I 4

availableH case. ~ Revision' 1" and Revision 2" of BAW-10173P provide the

- conuspording steam line' break transient analysis for the offsite power not:

i available case. Cycle specific statepoint analysco-were performed to confim -

that these analyses apply to Cycle'6. Se reactivitics of the statepoints were fourd to be less than the values used in the analyses. 20 minimum NBRs of the:

.-offsite power available and offsito power. not available statepoints were 1.'37

'and 1.88 respectively. Rese _ results verify that the existing offsite dose analysis for steam line break is applicable to Catawba 1 Cycle 6.

7-1 l

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B&W FuelCompany u___ .u. __ ._ _ . ,_ _ _

I e

MA. MJGEiN

'PO cyr,la s1 cdile },weters were ebeckcd with the limiting parameters for the

. i tod ejectIrm twolent (nd were acceptable. %ese parancters are shown in Table 7 9.

2d,*L RCCA Misemratiqu Is irdicated in BAW-10173P, cycle specific evaluations wo e perforted for RCCA misoperations for dropped rods, red misalignment, and sirryle red withdrawal pin j census. 'Ibe results of the dropped red evaluation as describcd in DAW-10173P h are shown in H gure 7-3. All the possible dapped rod combinations withit. cach

- individual gru.tp or bank were analyzed, ard the peakirg for these cases was less ,

than the flIBR peaking limit.

A ' statically misaligned alrgle RccA of control bank D could either tu N11y (

[.

iraerted .,r fully withdrawn. %e peakin; for a fully inserted RCCA is th ne as the peakir,J for a ch:pped bank D RCCA ard has been afelyzed. A ftdip withdrawn single bank D CCA with the remaining ruds a!. the inse# ion lisit was analyzed at Br ard IDC. We peaking increasts for the full 9 wi@ra% <ases

- were less severe than the peakiry increases for the droppcd red cases, aM were h therefore a)ro acceptable.

- 1 Ee peakire for Catawba 1 Cycle 6 is itss severe than the power distribution

{ used to determine th9 numbe: of pins failed in the t,irgle ROCA withdrawal analysis in bW-10173P. %erefore, the rur.h e of pins failed is bourded by the v=due of 5 percent in BAW-10173P.

( hi IIxiked. Rotor

%e locked rotor analysis in &%10173P was pedarmed with a design peakirn distributico that is typically conscrvative when canpared to an actual fuel cycle design. Ecwever, the WIPC cerc design methodology uses maximum allowable peakin (MAP) limits to prnait trtdo-offs between radial ard axial peaks in settirrj AFD operatirq liudtw. %ese WG -li tits define combinatiors of radial ard exial peaks that maintain DiBR eqdralence to the design peaking distrihation. For the cycle erteifAc evaluation of the locked rotor event, the

_l design MP limits (i.e. thoso PAP limits with IIIB equivalence to the design

. MaPJ.In dietriWtion) win rtducCN so that the allcwable radial peak at the 7-2 B&W FuelCompany m

.----_a__ _ _ _ _ _ _ . _ . m . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

L 1

design axial (2.55 at 0.5 K/L) was agalvalent to the hot pin radial peak thidi l

{

\ pnduced an acceptable DER durig the 109e3 mtor trancient. 7his radial pahim value had been determined by evaluatirg the lo:ted mtor transient at I

successively lower peakirg values until the hot pin pak that prechded DG was determined.- 7b detemine the pertentage of pins in Dm, pmdictad peakirn distributions for limitig operatirg conditions of Catasta 1, Cycle 6 wen

$' conpared to the adjusted )RP limits. All pins with peaks that exceeded the 4

adjusted 1%P limits were then assumed to be in DS.

l An evaluation of the Catawba 1, Cycle 6 core design has been performed ard the results irdicate that the percentage of Qins in Da is less than 3.3%. This amount of pins in DS is less than the assunption of 10% failed fuel used ja the l radiolcgical consequences evaluation for the locked rotor accident. Also, since the peak clad surface tanperni.ure is less than 1800*r the core will remain in I place and intact with no loss in coolig capability.

7.2 ECts Analysig A loCA analysis, applicable to the Westinghouse designed nuclear plants operated i j' by the Duke Power Ocarpvfy, McGuire Units 1 ard 2 aid Catawba Units 1 ard 2, has 4- ; been performed by IETC. The analysis supports operation of the four Daka units sith Mark-IM fue1, and is h=nted in topical rep:st BAW-10174". Methodology I

uployed in the analysis is in accort! ylth 10CITt50 Appendix K and is docunented 2 The toCA evaluation considered both in tcpical report IAW-10168P, Revision 1.

large ard small breaks, 'ard ' transition cores containing mixed Mark-!M and OTA fuel. 7he . evaluation carcluded that the small break 10CA (SBLOCA) fSAR analyses, performed by Westinghouse, zumain valid for plant licensite during the transitico cycles ard even after the ome is loaded with IETC-supplied fuel.

It was further concluded that, under mixed core operation ~, the Westirghouse FSAR analysis remains valid for OFA licensirg. Cors IDCA limits, resultirq frun the evaluations presented in !E10174 ard BAW-10174, Revlsion 1", axe given in the j Core Operating Limits Report for Catakba Unit 1 Cycle 6. All 10CA l configurations were fourd to be in c.Snformarm with the five criteria of 10CFR50.46, thus demonstratirg conservative results for the operation of Catawba

! Unit 1 Cycle 6.

p 7-3 p

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73 Radiolcnical ConncGE2D203 Lach ITAR accident aralysic hss bcen cualuated to determino the effccts of Cycle 6 cycration, and to ensure tMt the radiolcgical consogannoes of h)Tothetical accidents are witnin applicaolo mgulatory guidelires and do not adversely affect the health and safety of the public. The design basis LO3 evaluations assessed the radiolcgical inact of dif ferencos betwc3cn the Mrk-B4 fuel and Westinghouse OTA fuel fission prtduct oore irNcntorics. Also, the dose calculation cfracts fmn non-ID3 transients manalyzcd by Bfir utilizing Cycle 6 characteristics were evaluated. Differenas in the current IWJ1 dose values that are not relattd to the insertion or Mark-%f fuel reflect the application of the latest revisions to Standard Povica Flan ckt.c assessnont rcthodolcqy.

A brief dincassion of each ac::idant analyzcd is prwidcd beltu. A sunrary of the calculated radiolcgical consequences is providcd in Table 7-3, 7ho calculated radiolcgical consequerces are all within cpccified regulatory guidelines and contain significant Icycis of ratgin. ,

7.3.1 Icg;s of Ccolant Accidents 7ho offsite radiolcgica cons <.quences of a design basis 103 are calculatcd utilizing the applicable asstrptions containcd within Regulatory Guide 1.4 and Standard Review plan Sections 6. 5. 2, 6. 5. 3, 6. 5. 4, aid 15. 6. 5. The control room radiological consequenocs are calculatcd utilizirg the additional assunptions within Stardartl Review Plan Section 6.4 that are applicable to the cataktaa Cbntrol Rcan Ventilation design. The rod ejection accident offsito dose calculatior. is bastd on assumptions providcd in Regulatory Guido 1.77 ard Standard Review Plan 15.4.8. Chapter 8 contains applicable IWJ1 Chapter 15 pages appropriately revised to reflect the assumptions ur4d in the ICCA consequence analyses.

Usirn similar methcdolcgy, fission predact cc:? inventoric.s were calculated ascuming a reactor oore containirq Mark-DY fuel and one with Westirghouse OFA fuel. The results providcd in Table 7-3 demonstrato that utiliziry the Mark-31 fuel design as a replacement for the Westinghouse CFA fuel design produms differences of less than one percent in the calculatcd doses associatcd with the design basis LccA analyses that are in the 1989 Catawba FSAR. !kuever, there are several important assumptions which have bocn revisod in the analyses

'N l

B&W FuelCompany

describcd in this report that result in nore significant differerecs in the current F4 dose values, s

We most inportant conservatism that has boon ad:kd addresses the manner in which mixirg and filtration are assuzxd to oo::ur within the Annulus. We analyses presented -in the ITN1 assume mixirs of Contairment leakage in 50 pcIrcnt of the Annulus volume prior to filtration by the Annulus Ventilation filters. The analysis presented in this report assumes that Containment Icahage is processcd directly by the Annulus Ventilation syste filters prior to mixirg in the Annulus. We not effect of this conrarvatism is to decrease the calculatcd hold-up tine for radioactive decay within the Annulus and, thus, ircrease the calculated radioactivity releases to the environment. 21s assumption is omsistent with SteJdartl Review Plan Soction 6.5.3.

Another inportant conservatism added to the offsite dose analysis affects the assumod post-Locident leakage of ESF cmponents outside Containment. We analysis contained in the ITAR assumos that the maximum operational leakage occurs throughout the accident. The analyaid provided in this report also includes the leakage from a groca failure of a passive ocuponent. W e leakage

/ is conservatively assumod to be 50 gallons per minute, starting at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months i \ after the IDCA ard lastirg for 30 minutes. Although safety-related portions of 4

the Auxiliary Dailding Ventilation Syste service those areas of the plant Wern such a grecs failure is most likely to occur, no iodine removal credit is assuned for the Auxiliary fulldirn Ventilation System filters. W e not effect of this conservatism is to increase the calculated release of radioactivity to

_ the environment. Bis assumption is consistent with Standart1 Revics Plan ,

Section 15.6.2, Apperdix B.

An overly conservative assunption that has been rem 3ved deals with radioactivity n.dease pathways. %e IOCA dose analyses currently within the ITAR assume total failure of ' the redundant, safety-related hydrogen reacnbiners located in Contairment. This assunption necessitates the assumed use of the Hydrtgen Purrje System to maintain Containment hydrrgen concentrations belcw the theoretical flammability limit of 4 v/o. Such an assunption clearly requires two active failures, an assunption that is not consistent with other accident analyses

[

evaluated in the FSN1. he analysis provided in this -report assumes a single active failure that disables one hydrogen rectnbiner. We remaining operable 5 7-5 B&W FuelCompany

)

hydrtgen rocmbiner is able to mintain the Contaiment hydttgen concentration well belcu the 4 v/o flanmbility lintit, thus obviating the recd for the assumcd use of the flyditgen htrge System. The not offect of this anstmption is to rtduce the calculated release of radioactivity to the environ-ent. This assumption is corcistent with 10 CPR Part 50.44.

The total not ef fcct of those additional assu"ptions when ccrparxd to the IOCA analyscs pitnided in the 1989 ITM is to increano the calculated khole bcdy and eAin doces, ard to rcduce the calculatcd thyroid doses. As rentioned previously, in all cases the applicable regulatory guidelines contained within 10 CPR Part 100 ard General Dasign Critoria 19 are Uct with significant levels of mrgin.

7.3.2 lockcd Rotor Accident 1ho radiological corsoquences frtxn a reactor ecolant punp rotor seizure were reanalyzos due to the results frca IAW-10173P that predictcd 3.3% of the pins in the core would be in END. Regulatory guidance given in Standartl Review Plan 15.3.3 providcd the basis for the offsite doce cinsequence assessnent from a Iccked rotor accident. The calculated doses Are presented in Table 7-3.

Technical Specification lintits on primry ard socordary ecolant activities limit the potential doses to a small fraction of the 10 CPR Part 100 exposure guidelinos. The calculated doses are within 10 CPR Part 100 exposure guidelincs even -if the accident should ocuir with an icdine spike.

7.3.3 Simle RCCA Withdrawal at Pther The nost limitirg rod cluster control assembly miscperation, accidert.al withdrawal of a sirgle RCCA, is predicted to result in less than 5% fuel clad damge. The subsequent reactor ard turbino trip would result in atmospheric steam dunp, assumirs the condenser was not available for use. The radiolcgical corcequerces fram this event would bc loss than the lockcd rotor event, analyzed in fSAR Section 15.3.3 and Soction 7.1.4 of the Reload Report.

7-6 B&W FuelCompany

.--..~-...~.-_.-....~~.-- .

- . . - . . . . . - ~ . ~ . _ . . ~ . - - - . _ . . - . _ . .

Table 7-1. Safetv Analysis Chacilist for Ihysics Data

'l 1

. Du32tfdK _Ilmit Value Cvele 6 Value 4

Moderator Coefficients, pcn/F h2P, Maximan < +7 +4 HFP, Mav4== < .0 -2 All, Minim a > -32 Power Oxfficients, pcn/% power see figure 7-1 Pstf, 5 Mav1= = < .75 .63 Miniman > .44 .52 Trip Tinactivity, pan. 4000 see figure 7-2 i

Shutdown Margin, pam > 1300 2120 Mavi== Differential Rod Worth, panVsec < 63.75 40.3 Fui at HZP < 2.48 2.12 i

t.

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7-7

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B&W Fuelcompey 1

Table 7-2. Red dection Farraneters nge Limit evele 6 Value O1l Mu Ejccted Itxi ibrth, pat IIC, IGP $ 750 408 BOC, IDP $ 230 82 IDC, IEP $ 900 544 l IDC, IIFP $ 230 110 Mu Fn after ejection

BOC, lEP S 11.0 7.2 DOC,IFP $ 4.5 2.0 EOC, FEP $ 19.0 15.5 IDC, IIFP S 5.9 3.1 Dett, t IOC > .55 .62 IDC > .44 .52 Pin Census, %

BT, IEP $ 10 0 DOC, IIFP 5 10 <3 100,IEP $ 10 0 DOC, HFP 5 10 <3 Minimm Trip reactivity, pan DOC, IGP 2 2000 2400 IOC, HFP 2 4000 4720 IDC,IEP 2 2000 2500 IDC, }IFP 2 4000 4950

Fo prior to ejection is less than 2.32 set by the IIDs.

l 7-8 88W FuelCompany

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Table '7-3. PnMolocrical 00n _*_m Ebse Results Surmary

(

- I. DBA Offsita Dome with ECCS Leakage Sqm3;38 (Rem) d hclusicri Area Bourda.ty IIM Population Zone Whole Body myroid Enole Body 'Ihyrols}

A. Westinghouse OFA Core Inventory 9.08 127.0 1.14 32.1 B. Mark- N Fuel-Core Inventory 9.10 127.0 1.14 32.1

- II. DBA Offsite Dose without ECC9 Lea); ace Sauraas (Rem) hclusion Area Boundary Irw Fopulatjen Zone i Whole-Body %vroid Whole redv h yroid A. ' Westirghouse .0FA Core Inventory 9.05 118.0 1.12 13.4 5

- B.- Mark-W Fbal Core Inventory 9.07 118.0 1.12 13.4 M.

III._ Control Roam Operator Dose (Rem)

Whole ~ Body ghiD hyroid A. Westirginase OFA Core Inventory 1.63 32.1 14.2 0:

.B. Mark-BW Fuel Core Inventory- 1.64 32.0 14.2- ..

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Lable 7-3. Radiologigal Corran)erros Ibe.e Rggits Surrard (cont.)

P/. Red IMection Accident Of fsite Dr.g (Rem)

O Delusion Area Intni'iry los Tbpulation Zone M1 ole BMy %vrei_d thpig ~)pjy 'Ihyroid A. Westirghouse OFA Core Inventory TYimry 2. */ SE- 1 5.91 3.22E 2 6.72E 1 SOC TM 2.20 1.77E+1 1.57E-1 5.95 B. !brk-!M Fuel Core Inventory N O' 2.75E-1 5.90 3.22E 2 6.71E 1 SOT *O' 2.20 1.76E+1 1.58E-1 E.95 V. Isshed Rotor Accident Offsite Lbr.o (Rem) hclusion Area Dourdary 104 Tqa.11ation Zono Whole Ibdy 'Ihyroid Whole Itdy 1hyroid

1. Case 1 4.41E-1 3.63 3.10E-2 1.20 (11o iodina spiko)

2. Case 2 A.41E-1 3.67 3.16E-2 1.21 (Pre-iodine spike) 7-10 B&W FuelCompwiy

. _ _ . . _ ._ _ _ _ _ _ _ _ . . _ _ _ _ - _ .. _ . _ _ . _ _ .~

N Rgure 71 Doppler Power Coefficient I

i i

toast Negative Doppler

8.0I i

g *

,7 : "

I l 1 K 12.0- -

1 1

4 i

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l 16.0-Most Negative Doppler .

4

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20.0 .

o- so 4e so ao . en l

% Power I } Rahpe et Values for Catawba 1 Cycle 8 w

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7-11 B&W FuelCompany

4 Figure 7 2 l Scram Curve '

5.00 l

4.00- '

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3.00-f' u

Catawba 1 Cycle 6 3 2.00-n:

1.00-

SA Trip Worth

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0.00 g y 4 s s ioo

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7-12 B&W FuelCompany .

4 Figure 7 3

( Catawba 1 Cycle 6

Dropped Rod Peaking 28 --- -- --

l l

800 E00 DN8 LWt 24- xxx ooo - ... -

20-

x 3 16 -

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x .

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o j' 40 IO NO /0 80 $0 150 10 12 0 l Maximum Achievable Power, %

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v av ---e N--- Tr V '*-se y,..Idi e-wwwa't-'*'*r' W y N v pp .e -vP='w' = * = Tr -w-w tr w -q ?

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8. HOICSED lODITICATICriS TO TD3NICAL SPECIFICATICt1S AND COIR 7he %::hnical Specifications ard Core Operatirg Limits Ro;rrt (COIR) have been i reviscd for Cycle 6 operation to ac.umalate the influerce of the Cycle 6 core a design on power oeakirg, reactivity, ard control rod worths. W e Technical Specification limits ard COLR limits also reflect charges in reload analysis methodology" beginnirn with this core. We cycle 6 design analysis basis includes a low-leakage fuel cycle design ard a mixed core containirg both IEW Mark-IM ard Westinghouse ora fuel assemblics.

A Cycle 6-specific power distribution analysis of the final core design was

~

corducted to generate the f(6I) limits for the Overpur Delta-T ard Overtenperature Delta-T trip functions ard the Limitirg Corditions for Operation (control bank insertion ard axial flux difference). We f(AI)-

limits preserve the centerline fuel melt ard steady-state WIR limits, and the Limitirg Conditions for Operation preserve the maxir.um allowable IDCA ard initial cordition INB peaking limits, ejected rod worth reactivity limits, ard the shutdown margin reactivity limit. W ese limits were developed based on the NRC-appivved methodology described in Reference 17. - A peakirg penalty for quadrant power tilt was taken in the analysis so that the resulting limits acc.wm4ste qt adrant power tilt ratiosj up to a value of 1.02.

We nuxims Alowable LOCA peaking limits shown in Figure 4 of the COIR _are based on the IhTO ECCS evaluation described in Section 7.2. A camposite K(Z) limit was developed based on both large ard small break analyses. Separate camposite limits applicable to Mark-IM end OTA fuel were used in the power distribution analysis, and are specified in the CDIR. W ese limits were used directly in determination of the contiti rod insertion ard axial flux Ldifference operatirg limits given in Technical Specifications 3.1.3.6 ard-3.2.1. Technical Specification 3.2.2 provides the nuclear heat flux hot channel (ro )-peaking limit.

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= _ _ _ . .. -.

We initial coniition DIB mximum allwable peakirq (!%P) limits chcun in Table 3 of the COIR are bascd on core referenm design peakiry factors. Wo

!%P lbnits prwide allcwble a:nbimtions of peakirg factors that preserve DIBR performnm (quivalent to the uesign pur distribution for a limiting Icss of coolant ficM transient. 7ho initial cordition WJu aru uscd as described in Refererm 17 to calculate DID ptirg mutJ ins for decermimtion of the control red positicn ard to:ial flux diffenice operatirg limits given in 7tchnical Spccifications 3.1.3.6 ard 3.2.1. 7tdmical' Specificatico 3.3.3 '

provides the nuclear enthalpy rise hot chanrel (r a ) peaking limit.

The methcdolcgy for survei14ance of tho core hot chinnel peakirg factors, Fo(X,Y,Z) ard Fm(X,Y), is describcd in Referonx 17. In this application of the mothcdology, Duke Ptuer Capany has electcd to bypass the first tier surveillance (caparison of measured peaking to prodicted design peakirg), aid to perform the peaking mrgin calculation directly shonever an ircore flux map is taken for surveillanx nonitorirg. This is a conscIvative application of the nonitorirg methcdolcgy ard is therefore acceptable. Specifications 4.2.2 ard 4.2.3 have boon written in a forn that prwides this capability, ard only the grancters required by this application of core monitorirg are pIwided in the CDUR.

The core operatirg limits are provided in the Core Operatirg Limits Report, in accottlance with Imc Generic lettcr 88-16 aM Technical Specification 6.9.1.9.

Table 8-1 lists the 7bchnical Specification changes required for Cycle 6, khile Table 8-2 lists the charges to the core operating limits contained in the CDIR. These charges are beirg submitted to the imC urder separate covers.

Patnmeters related to nonitorirg the core pcuer distribution are defined in Reference 17, ard are used by the plant camputer software. These parameters will be supplied for irclusion in the CDUt.

Based on the analyses ard revisions to the Technical Specifications ard COLR described in this report, Cycle 6 of Catawha Unit i vill operate within the Final Acceptanco Criteria In:S limits ard within the thermal design criteria.

The folicwing pages contain the required 7bchnical Specification revisions and the ruvisions to the core cperatirq limits specified in the COIR.

8-2 B&W FuelCompany

9 (N D ble B-1. 7tqhnical _ Oprifiratigm_Qaang f

14plicable BL%_SICL lh .

EcasoILf91_QWrlp 2.1.1 changed C.T correlation increased Fai for Mark-IM fuel reduced minin mamtrrxl Tys ficv 2.2.1 opmplettd RID bypass rcmruni rodu:xd minimam masurcd 2 ficw increascd error allwanoes nn certain reatt. or trip instrunentation 3/4.2.1 deleted baseload operation 3/4.2.2 chargcd Fo mthodology 3/4.2.3 chargcd Fm mthodolcgy ocparatrd RCS flcu and Fu relationship 3/4.2.4 increased the tilt ratio at which a p:. wor rcdwtion is rcquired rewrot.c IID to be consistent with Westirghause FIS

(] j 3/4,2.5 ircorporated RCS ficw as a t2iB paramter j remvod pcuar/ficw trndooff dependence on R

't/ Irxtuccd minimum measurcd RC3 ficw 4.5.2 charged ficw and developcd pressure rcquirements to be consistent with reviscd accident analysis f1cu assunptions 6.9.1.9 . reflccted the charge to IMV operatirg limit methodology (3

I 8-3 l

B&W FuelComparty

Igbie 8-2. Cerq_Gryfrglin7 limits Repathgs, Applic2d>le Irgh_fpar, .No.t kcanon for Ch3D22-O '

_ .s b

3/4.1.3.5 revisod cafety terJ: irr.crtion limits to reflect a red withdrawal limit of 222 r. tops 3/4.1.3.6 revicod control kan); insertion limits to reflect a red withdrawal limit of 222 steps 3/4.2.1 revisod AFD limits for Cycle 6 operation 3/4.2.2 reviscd for Cycle 6 operation to reflect a chargo in the Heat Flux ilot 01annel factor Fe(X,Y,Z) nothodology 3/4.2.3 revincd for Cyclo 6 operation to reflect a change in the Nuclcor E:nthalpy Rico }iot Ounnel factor Fw(X,Y) rethcdology 0

8-4 B&W FuelCompany

i i

4 t

Owges to 7bchnical Specifications N

J l

l

! 8-5 l

l l-B&W FuelCompany

['

5

\ 660 g_,, , , i ; l t i ;, ,

p l N I ! I i l l 1 i #i- I i

j g, j i ; i UN ACCEPTABLE F OPERATigN . .f

, i lixi i i i ,

i ji j

Ni i 6 i i i l i i j i i .i , g24,00 PSI A , i l ;

6

i iN i i IN i i t- t I i 640 i i\I ) ! l\ ' i 1 1

, ; , \; , ; ; ;\ , ; ;

i , i l

~

N250 PS(A ' \; t' q I 1 i i N i. i i :\:

N i i .

Ni i \

i N 1i i i !

N kt ..\ n

! iN I I i i i Nii\ i r it i i Nit . i i N i\ c C i i t ; '

N. .

l 1.

. \i \"

620 ; : i N2000, PSI A l , \ \$

i i

> Ni '

l I i i\i i i :- I i\'\l

/ K; I I i i NI i I.. lt .i\'

th i N : 1 i i i A1- 1 I \

Og N i i N. I i i I AI "\'

/ i iN i ! j _

i '

IrN -

i i iN.I i .. t. . i . . ..F - N

i i ! - A,1775, PSI A .

g.; p jgi 600 ,,,,.

, , .. g ,. g 4,g..y .

!- i i j l I

N -

i i.r l' 1\

i j . I iil l .

t i K l. a ! i.. '

I i i I i i i NI !

i ACCEPTABLE i . i i i i i i N :

I I ' I 'I I ' I OPERATION l i . .

} l l I .

} I i 'Ih '

j j i j i i4 i i i I 580 g i ; ! ; , ; , , ; ;; g i

i i i i1 I l i l I .-i i t' j i i i i i l l t i . I 1 ii

{ l , i I i i i I t i i i i 0 0.2 0.4 0.6 0.8 1.0 1.2 FRACTION OF RATED THERMAL POWER q FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION CATAWBA-UNIT $$f$$2 2 '.

B-b

j-I a

1 l Figure 2.1 1 Reactor Core Safety Limits Tour Loops in Operation i

I 665,

' 660 Unacceptable Operadon 655i.

650:- 2400 pela >

6451 2250 pela 640 . -

635i 630i E-w p625:- 2000 psia

. m.

W m 620 3 0 s at: .

015:-

610 -

1915 pela

- 605--

' 600i-l Acceptable t-595i Operation 590i- t 585:-

580- - . . . -

0.0 0.2 0.4 0.6 0.8 1.0 1.2 l Fraction of Rated Thermal Power l

- CATAWBA UNIT 1 8-7 w--- -y- ge y - --y y n g +e- g gv-we ey- v--yv yy pd wgy

,<weig--+;-e.'mm-%a4-e*mme-e m--wwa--s.--eve - + e wwww.e e s-m* w - e tse w m esi-we'eew - w w N- m-mm'-We--'pM== N- ' nee **** + - "

F -"' *

/

,m ~

p),

L. Q).

TABLE 2.2.-I S

REACIOR TRIP SYSTEM INSTRUMENTATION TRIP SEIP0lNIS lii

(, 10iAL AttOWANCE SENSOR ERROR AiLOWABLE VAtut (5) IRIP SEIPOINT (IA) Z h FUNCil0NAL UNil N.A. N.A. N.A. N.A. N.A.

N 1. Manual Reactor Trip gio,q w

2. Power Range, Heutron Flux 0 $LLL-It of RIP *

, 7. 5 3ACS.97,. $109% of RIP *

a. High Setpoint 0 $25% of RTP* $27.1% of RIP *

b. Low Setpoint 8.3 34 ( S.92-0 <5% of RIP

with <6.3% of RIP

with Power Range, Neutron Flux, 1. 6 0.5 i time constant

3. a time constant liigh Positive Rate > 2 seconds

~

3 2 seconds l 1.6 0.5 0 $5% of RIP

with $6.3% of RIP

with

4. Power Range, Neutron flux, a time constant a time constant liigh Negative Rate ~>2 seconds , 32 seconds m ~

fu 8.4 0 $25% of RIP * $31% of RIP *

A 5. Intermediate Range, 17.0 Neutron Flux $1.4 x 105 cps )

17.0 10 0 $105 cps

6. Source Range, Neutron Flux See Note 2 See Note 1 4,A 2,83

7. Overtemperature al J # See Note 3 See Note 4

,4AY4.9 ) JM{1.241 17 2(1.7 )

8. Overpower AT 31938 psig***

4.0 2.21 1.5 31945 psig jrj p 9. Pressurizer Pressure-tow $2399 psig Pressurizer Pressure-fligh 7. 5 ,4/ J ( O 'll g l.S $2385 psig 10.

1.5 $92% of instrument $93.8% of g 11. Pressurizer Water level-fligh 5.0 2.18 span span gginstrument g,q g o

0.6 390% of loop 3 "' " (M of loop 5

?f 12. Reactor Coolant Flow-Low 382.92 minimum measured minimummeasurediIow'*l fIow** l v e*s~

E U:

s 3g 'RTP = RATED THGfMAL POWER ? ym l l w ** Loop minimum measured flow = %g';00 gpm

~

- - iime constants utilized in the lead-lag controller for Pressurizer Pressure-tow are 2 se and I second for lag.

1 l

dues.

F -licehle .upon deletion of RID DypesL5ysica.

i

\ f%

i

~

x_/ J .

i 2

o TABLE 2.2-1 (Continued)

. IA8tE NOIATI0lG

>- - f 6 NOTE 1: OVERTEDFEMATURE AT AT I 12 5)y AT, M - 32 g l ,;3) U (3 f t 3) - T*] + K3 (p - p') - f,(al)) ,_

T5 l i ! = Measured AT by RTS Manifold Instrumentation; I d idhere: AT '

I

". 1 + 1'5 = tead-lag compensator en measured AT; e y,,$ ,

N t i 3 Nr 3

i Is, 12 = Time constants utilized in lead-lag compensator for aT,1 = $ 12*1 s, l 1 =3s, i

1 = tag c - nsator on - as. red aT 1 . ,,s 13

= Time constant utilized in the lag compensator for AT, r3 = 0; l

! ?

e ",' AT = Indicated AT at RATED THEIWmL PoldR; j . o I K =

,M$1.38 ; g 8

K = 0.92401/*F; ;

i I * ** = The function generated by the lead-lag compensator for Tavg i 1 + ts5 dynamic compensatlon;

, ;,3 s,

t., 15 = Tlee constants utilized in the lead-lag compensator for T,,g, r, =%22

,k

i 1s = 4 5; l 3% = Average temperature, "F; l ;: :: T

c,<>

I l t:*;

t y , ,,$

= tag c @ sa w a w as W T ,,,; ,

! 22 =- Time constant utilized in the acasured T,, lag compensator, is = 0; ELEL 1 l

l

l!. ,j f f ! i I(li1 i7 ! i' 3';! -, tI*3!>

!1l1l!}f !il e , - .

l p

N i

I r R E

W y e

I O h m f A P b t o o e t t h D e L n

f t t t t E h A one o n ,TA t M i o

b e R

- sm c %R 5 % HT

. E t

p ru d r n e a pe or t 5 s tt 2 a cs 2 6 D p en p n - e - E ti o i u T i r

e t nl n A dd R a a a R h

T e h v mr ou kEOW t

t s

t t a

d e

P e t e t t s t a n v i v e u

i u p

) oe i L A

i t f t l m s bm R a o i a o i

dn M g% s v c

. s y na e o s P st l

a )

a ped NK I n

e 1

1 5

8 i t

i n e os r . f A a o 3 8 f d t

y r

u s

t a n

b mlo t m s b y 8 o S %1 e

e cx f

e s e

ed ee wt Ets 1

i d

I 4 e

) a r geu d S p t c ee D i

c %61 t o

) e d =

e in b y g n 8

bl e

E I g . e d -

y n _

es i

u n t n

d e

i t

a s c ne

+

g gry g tb d

l l

a t o w ; r , eb g f l f e h n C o g e r r o la o cu s o( t p e ee .

C (S

l l

a s

p S

o t a

r f l :

f isa t

5 md pa n e c d i e d d e u r t

n

_ , 1 O N

e, C e p

dnh 6 , t u t a t i

o

- I

- N it e y m i y p

_ 2 T M r o da .

d r i

n o n l t

. A T u l e gh c n a ev l

g t g l e 2 T s a e t a a u a a S O l s n r ahu ma m c E N a m

e r

i m f e ct s ii % git e e e t i

i p

L p o s dws 5 c h a r B E A L l

e N n n t e p

h b t t m T TB e r ( a i;s 2 s l o .

A I e r se 2 e t l t t T

I

( ; z g t ert -

tr ; a a a u

_ 9 i i he n g E h h h a T 8 r s e tbP t s d t i

u p c mU e e E n e x 8

1 1 s a f aI e w re o e rO W t t a t b a3 0 s 5 l p

ohR c P n n n m 0 0 e 3 a n cA t e h e i ; e l s

9 r 2 T c o R c l

_ 5 0 P 2 t onS b w e hLA r p E r a io b , t M e t W e h 'l tit R p e O p s e

<- = = = = c nna n q 0 f E 5 P nn na o H h h t uol f rp

- =

T c p t c a i n ah ht t ) s a i A e op c

_ aug q l e D e r t e e

~

en r a vl E

(, a I r T l E r t er sni oT H o e he

_ i r o T e

)

T K 3

p 'P 5 ) gd eu f f h M F A T F S d l n e aae

_ u i

n ( ,rs - n t f ro n ep ) )

d ws )

_ o noe 1 i i C apr ( i i

_ ( ( i

. (

2 1

E I

E O i

N rC o N

, c, o*eE E h z.d

- l .- e- en

=

. E ip[h_

%E .

l ; - , :

N s A

.-l 6 i

r TABLE 2.2-1 (Continued) n TASTE IIDIAllOIS (Centinued) j m -

g E pe01E 3: OVERPOWER AT--

c e

ai (1 + t,5) ( 1 ) I*~

( t,5 ) '((1 1+ toS)

) 1 T - Ko (I (gg , ,,5)) - I.] - f (al))

i (1 + 125) (1 + 1 35) 1 A o 5 (1 + t,5) .

. t

e. Where: AI = As defined in Ilote 1 I * $ = As defined in plate 1, 1+152 -

.r t

= As defined in plate 1.

ts. T2 i 1 = 'As defined in M e 1, i g , ,ag i

+

= As defined in Ilote 1, [

r3 !

c.o m.

4.0 ,

= As defined in IIote 1,

- AT, i f

1.0704,

\

=

K. !

+

K3 = 0.02/*F for increasing average temperature and 0 for decreasing average temperature, ,

The function generated by tine rate-lag controller for i ava dynamic d

125 = i j

> 1+:3 5 compensation, t i, = Time constant utilized in the rate-lag controller for T, ,sy = 10 s,-

.l 1 = As detined in 16ete 1* ,

1.+ ts5

s. = As defined in Ilste 1 :

l

+

e _ . -

/ \ /

N s

(

v/

i \

%/

) ('QJ l n TABLE 2.2-1 (Continued)

D TABLE NOIATIONS (Continued)

?E E NOTE 3: (Continued)

E K. = 0.001707/*F for T > 590.8*F and Ks = 0 for I i 590.8'F, f.

d T = 7- defined in Note 1, e

e. 1" =

Indicated T, at RATED TIIERMAL POWER (Calibration temperature for aI instrumentation, 5 590.8'F),

S = As defined in Note 1, and f 2(AI) = 0 for all al.

NOTE 4: The channel's maxi Trip Setpoint shall not exceed its computed Trip 5etpoint by T'?

w more than 226%2.

{

t?

aa 34 11 ee t:E 22 n "-

App t i~ cal 23poperaR1stMti 23Mt1B4ypasst4fs,temy i

I

a. 1 j

i 2.1 $AFETY LIMITS

'g BA$f5 L I

,]

i 2.1.1 REACTOR CORE 1 4

The restrictions of this $4fety Limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant-saturation temperature.

Operation above the upper boundary of the nucleate boiling regime could t result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer cosi'ficient. DNB is not-a directly measurable parameter during operation and therefore THERMAL POWER and ator Coolant Tempera) re and Pressure have been related to DNB through the correlation. The ONB correlation has been developed to predict the NB flux and the at on of DNB for axially uniform and nonuniform heat lux distributio . The local DNB heat flux ratio, (DNBR), is defined as the ra io of the he lux that would cause ONB at a particular core location to he local he) flux, and is indicative of the margin to DNS. g g/

'The DNB design bas is as follows: there must be at least a 95%

/ probability that the nimum DNBR of the limiting rod during Condition I and t II events is gr han or equal to the ONBR limit of the DNB correlation U being used (the correlation in this application). The correlation DNBR limit is established based on the entire applicable experimental data set such that there is a 95% probability with 95% confidence that DNB will not_ occur when the minimum DNER is at the DNBR limit, n de. EWC AV I4 corretJIQ In meeting this design basis, uncertainti'es it) plantoperalingparameters, nuclear and thermal parameters 4 4 fuel fabrication parameters 4are considered statistically such that there is at least a 95% confidence that the minimum DNBR for the limiting rod is greater than or. equal to the DNBR limit. The uncertainties in the above +4* parameters are used to etermine the plant This ONBR uncertainty rc::Hd W the cerrchtien OZE is used Yo DNBR uncertainty.

Phnitt establishee a design DNBR value which must be met in plant safety analyses using values of input parameters without uncertainties.

.The curves of Figure 2.1 1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature below which the

. calculated DNBR is no less than the design DNBR value, or the average enthalpy at' the vessel exit is less than the enthalpy of saturated liouid.

O V

CATAWBA UNIT $ 1 ) { B 2-1

. ,e-13

t j

t i

b C

p 2.1 SAFETY LIMITS BASys N

curvalMbased on a nuclear enthalpy rise hot channel (Actor,

) Fg ,_of 1.49,and a reference cosine with a peak of 1.55 for axial power shape, i An allowance is included for an increase in F N at reduced power based on the expression: 14, g,; p , n ,4 ct ,

o g ,. g 4 Tj,c g f; (opa f) % Q N % J.2 5 4 r W Fati s Fg = 1,49 (1 + 0.3 (1 P))4' n,km*bh!"C

! N q t. ..A s <o tn ~M A 'c es

~

[ Where P is the fraction of RATED THERMAL POWER.

These limiting heat flux conditions are higher than those calculated for

, the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance is within-the Ifmits of the ft (AI) function of the Overtemperaturentrip. When the axial power imbalance is not within the tolerance, the axial temperature AT trips will reduce the Se$cwer t ointsimbalance to provideeffect on the consistent protection Over-with core Safety Limits. (

2.1.2 REACTOR COOLANT SYSTEM PRESSURE

(

/ The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization a.nd thereby prevents the release of A

x radionuclides contained in the reactor coolant from reaching the cotaainment atmosphere. '

{ The reactor vessel, pressurizer, and the Reactor Coo'lant System piping, valves, and. fittings are designed to Section III of the ASME Code for Nuclear Power Plants which permits a maximum transient pressure of 110% (2735 psig) of design _ pressure. The Safety Limit of 2735 psig is therefore consistent with I- the design criteria and associated Code requirements.

The entire Reactor Coolant System is hydrotested at 125% (3110 psig) of design _ pressure, to demonstrate integrity prior to initial operation.

h * ?

l a

l I

l l

t i

V e

CATAWBA-UNIT $1$$ Bg 2-{

3/4.2 POWER DISTRIBUTION LIMI15

( / 3/A.2.1 AXIAL FLUX DIFFERENCE (AFD)

LIMITING CONDITION FOR OPERATION ,

(+h e a cc eptabie (;mic.e sxcined in +he ccd N A'Nnv6 t.imoN SrcO(<sk.)

3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) thall be maintained withih Ag [a. the allowed operational space as specified in the CORE OFERATING LIMITS REPORT (COLR) for RAOC operation, or

b. witnin the target band specified in the COLR about the target flux difference during baseload operation.

APPLICABILITY: MODE 1, above 50% of RATED THERMAL POWER.*

ACTICH: p,,.d e-

6. For AOC operation with the indicated AFD outside of the limits spec ed in the COLR,

1. Either restore the indicated AFD to within the COLR limits within 15 minutes, or

/ 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER (v ]) within 30 minutes and reduce the Power Range Neutron Flux-High Trip setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

(b. For Base Load operatiun above APLND** with the indicated AXIAL, FLUX DIFFERENCE outside of the applicable target band about the target Ak flux difference:

" 1. Either restore the indicated AFD to within the COLR specified I target band limits within 15 minutes, or ND 2, Reduce THERMAL POWER to less than APL of RATED THERMAL POWER and discontinue Base load operation within 30 minutes,

b. (, THERMAL POWER shall not be increased above 50% of RATED THERMAL P0WER unless the indicated AFD is within the limits specified in the COLR. l

$let e-

See Special Test Exceptions Specification 3.10.2.

I **APL"U is the minimum allowable (nuclear Gesign) power level for base load (d ]peration and is specified in the CORE OPERATING LIMITS REPORT per Specification 6.9.1.9._ f 3/4 2-1 Amendment-Hc. " (Urdt 1)-

l CATAWBA-UNIT {1] 8 15 -Amendment-He-68 (Unit 2)-

l

, s.

TN POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION ,_

SURVEILtANCE REQUIREMENT 5 _ ,

4.2.1.1 The indicated AFD shall be determind to tu within its limits during l POWER OPERATION above 50% of RATED THERMAt P WcR oy:

a. Monitoring the indicated AFD for each ODERABLE excore channel:

1) At least once per 7 days when the AFD Monitor Alarn is OPERABLE, and

2) At least once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after restoring the AFD Monitor Alarm to OPERABLE status,

b. Monitoring and logging the indicated AFD for each OPERABLE excore channel at least once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and at least once per 30 minutes thereaf ter, when the AFD Monitor Alarm is inoper-able. The logged values of the indicated AFO shall be assumed to exist during the intervn1 preceding each logging,

c. The provisions of Specification 4.0.4 are not applicable.

/

! 3< 4.2.1.2 The indicated AFD shall be considered outside of its limits'when at O leart two OPERABLE excore channels are indicating the AFD to be outside the limits.

[4. 2.1. 3 When in Base Load operation, the target axial flux difference of each OPERABLE excore channel shall be determined by acasuremont at least once per 92 Effective Full Power Days. The provisions of Specification 4.0.4 are not applicable.

4.2.1.4 When in Base Load operation, the target flux difference shall be updated at least once per 31 Effective Full Power Days by either determining the target flux difference in conjunction with the survelliance requirements of Specification 3/4.2.2 or by linear interpolation between the most recently mea-sured values and the calculated value at the end of cycle life. The provisions of Specification 4.0.4 are not applicable, e/et

('N

\

i CATAWBA - UNIT 3/4 2-2 -Ameneent-Me. 39 (UM t 1}-

1Q 8-16 4eereent-Her-&-@W.r-4) ,

_ _ _ _ _ _ _ _ _ _. _ _ . ._ _ .m _ _ . .

90WiR O!!tRIBUTICN LIH!1S 3

E4. 2. 2 HEAT FLUX HOT CHANNEL FACTOR - [

j-D :

[1MITINGCONDITIONFOROPERATION

! ' T_L,2 . -

[ @,r, t h .

GymnD f 3.2.2 f shall be limited by the fo11 ewing relationships:

'Q(Z)

W I f K(Z) for P > 0.$

l MA P l

q G,4 a)

Fh<F g

RTP K(Z) for P i 0.5 l

. Where

FhTP=theF g Limit at RATED THERMAL POWER (RTP j specified in the CORE OPERATING LIMITS REPORT-(COLR), ,

7 > :

P = THERMAL POWER , and >

RATED THERMAL POWER r7 (,xs (d i

fg K(Z) = the. normalized F (Z) $;r : g M n ::r: h:i ght-

) spectfied in the R

.g , m u p w,.;ge.(ed 9 pes, i'

APPLICABILITY:

MODE 1.

Lap ue win t ACTION: ~g j(yg +) g g ,,,y,e f ,

With Fg (Z) exceeding.its limit:' ./

a. _l Reduce. THERMAL-POWER at least 1% for_each 1% F (Z) exceeds the limit i 9

,'. .within 15~ minutes and similarly reduce the Power. Range Neutron t Flux-High Trip $etpoints withi_n.the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ;1 POWER OPERATION may proceed for up,to a total of.72. hours; subsequent POWER OPERATION:

, .may proceed provided.the Overpower AT Trip Setpointa (value of-K4 ) have i been reduced at least 1% (in AT span) for esch 1% n F (2) exceeds the ]; 3' limit. andy -

' 4)l. Identify and corrtet,the cause of the out-of-limit condition prior.

-to . increasing THERMAL-POWER above the reduced-limit required by ?

0N a;, above; THERMAL POWER may then be increased provided '

AQ(Z))is demonstrated through incore mapping to be within its limit.

~

^ /E (4Gd) '

( ~

g * ('x,g g) : +4e m e nwc o' h e

6 fl'"

a g, g, t a,. a n t s s o rf ccific'I In N '?

h*A 'b"""'I '

CATAWBA' UNITS 1h 3/4 2-5 5:n ecnt Ns. ?? (Unit-1t -

8-17 .A,.enoment No. $0 (Unit. 2)--

r 1_

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

- - . _ . . . . - . . . . . - . . _ . _ - . . . - ~ . _ - . . . - . . . - - - . - . - - - . - - - - . - - - . . _

l for Specification 3.2.2 I

i 1- - l 4

iAttachment 1:

1

,. a. Reduce TIIERMAL POWER at least 1% for cach 1% Fn*(X,Y,Z) exceeds the

!: limit within 15 minutes and similarly reduce the Power Range Neutron Flux High Trip Setpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , and l

b. Control the AFD to within new AFD limits which are determined by reducing the allowable power at each point along the AFD limit lines of Specification 3.2.1 at least 1% for each 1% Fn*(X,Y,Z) er eeds the limit within 15 minutes and reset the AFD alarm setpoint co the I modified limits within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , and l

.c. ~.JWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; subsequent POWER OPEPATION may proceed provided the Overpower 6T Trip Setpoints

-(value-of K ) have been reduced at least 1% (in 6T span) for each 1%

4 i

fn*(X,Y, Z) exceeds the lirr ' t , and a

l ..

I,

~

L l

l' h

8-18

_ . _ _ _ _ . . _ . . - _ . . . . _ . - _ _ _ . . . - - - - _ . - - . _ ~ .-. -__ _ . _

i q

POWER DISTRIBUTION' LIMITS O

SURVEILLANCE REQUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4(gre not applicable. @^M)

Fq%M 2) wAsMa 4.2.2.2 ForRA00oper: tion,F(z)shallbeevaluatedtodetermine[44F(2) is within its=limi_t by: 9 9

a. Using the movable incore detectors to obtain a power distribution

-map at any THERMAL POWER greater-than 5% of RATED THERMAL POWER.

(b. Increasin0 the measured F (z) component of the power distribution 9

map by 3% r,o account for manufacturing-tolerances and further in-

]

g creasing the value by 5% to account for-. measurement uncertainties.

. Verify the requirements of- Specification 3.2.2 are satisfied. -)

c. Satisfying the.following-relationship:

RTP M F Q (7) i QP-x--W(z) x K(z) for P >.0.5 RTP' '

N F

L k. Fg ',z) 1 Q x'K(z) for P ( 0.5 W(z) x 0.5 -

where F'(z)'is-the measured Fq (z) increased by the allowances for manufactur_ingtolerancesandmeasurement. uncertainty,FhTP is:the F

q limit 1

K(z)-is Lthe normalized Fq (2) as a function of . core height,

-P;is-the relative; THERMAL POWER,-and W(z) is-the cycle dependent functi_on that" accounts for power distriaution transients encountered duringinormaloperation.'F TP .. K(z),. ana W(z) are specified in the

CORE OPERATING LIMITS _ REPORT per Specification 6.9.1.9.

d. LMeasuring.F (z)'according to'the_following. schedule:

- 3 1.- 'Uponl achieving equilibrium conditions after exceeding by 10% or

"' .nore of-RATED THERMAL' POWER, the THERMAL POWER at which F (z) 1 j3 ,

was last determined,* or 9

2. A't least ^ .ce per 31 Ef fective Full Power Days, whichever occurs e f i rs t.~

L

During power escalation at the beginning of each cycle, power level may be j

} . increased until a-power level for extended operation has been achieved and a z

f

( power distribution map obtained. 7 CATAWBA-UNIT {1${ 3/4 2-6 -Amendment "c. " (Unit 1>

8-19 A.Tendment-No. 50 (Unit 2)

- a. - -- . - .- - . - - ._ - .. .. - . .. -

for Specification 4.2.2.2 Attachment-2:

) f

b. Measuring Fo H (X,Y,Z) at the earliest of:

1. At least once per 31 Effective Full Power Days, or

2. Upoh reaching equilibrium conditions after exceeding by 10% or more of RATED THEPJiAL POWER, the THERMAL POWER at whi::h oF "(X,Y,Z) was last deteninedm, or 3, At each time the QUADRANT POWER TILT RATIO indicated by the excore detectors is normalized using incore detector measurements, mNo additional uncertainties are required in the following equations for FgM (X,Y,Z), because the limits include uncertainties, mDuring power escalation at the beginning of each cycle. THERMAL POWER may-be increased until a power level for extended operation has been achieved and a power distribution map obtained.

A k

8-20

POWER DISTRIBUTION LIMITS fN SURVEILLANCE REQUIREMENT 3 (Continued) i i

( ,/ With measurek.,_ents indicating fe.

maxitoum

~

F (z) over I K(z)

D has increaset since the previous determination of F (2) either of Repla c*- the folloding actions shall be taken:

win AS.chet

+ 1) F (z) shall be increased by 2% over that specified in Speci fication 4.2.2.2c. , or

2) F (z) shall be measured at least once per 7 Effective Full Power Days until two successive maps indicate that maxit.ium

~

F (z) is not increasing, over 2 K(z) ,

With thelelationships specified in Specification 4.2.2.2c a5ove

[ f. not being satisfied:

t 1) Cal:ulate the percent Fq (z) exceeds its limit by the following expression:

r 1 maximum (z) x W(z)* 1 x 100 for P > 0.5 gwcwt over z p RTP /

3 -

x K(z) r'. - -

maximum. (z) x W(z -

x 100 for P <'04 5 over I p RTP

.5

K(*) , j

2) One of the following actions shall be taken: ;

a) Within 15 minutes, control the AFD to within new AFD limits which are determined by reducing the AFD limits of Specification 3.2.1 by 1% AFD for each percent F (z) exceeds l 9

its-limits as determined in Specification 4.2.2.2f.1).

Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , reset the AFD alarm setpoints to these mod-ified limits, or b) Comply with the requirements of Specification 3.2.2 for j g(z) exceeding its limit by the percent calculated above, or j

[,

y ) c) Ve'rlfy that the requirements of Specification 4.2.2.3 for Base Load operation are satisfied and enter Base Load operation.

CATAWBA - UNITS 1Q S/4 2-7 -AmendmeM-Nc . ( U n i t -D-8-21

6. _____-__.____________-______m____ _ _ _ _ _ - _ _ _ _ _ __ _ _ __._____ _ _ _ - .. ..M

for Specification 4.2.2.2 Attachment 3:

'N c. Performing the following calculations:

1. For each location, calculate the '% margin to the maximum allowable design as follows:

Fn"(X,Y,Z)

% Operational Margin - ( 1 - ) x 100%

[ Fo'(X , Y , Z) )"

% RPS Margin - ( 1 - 0( ' ' )

[ Fn'(X , Y , Z) ]"# ) x 100 %

L where (Fn(X,Y,Z)0P and ( Fn'(X,Y,Z]Res are the Operational and RPS

' ~

design peaking limits defined in the ColR.

2. Flad the minimum Operational Margin of all locationc examined in 4.2,2,2.c.1 above. If any margin is less than zero, then either of the following actions shall be taken:

(a) Within 15 minutes:

(1) Control the AFD to within new AFD limits that are determined by:

=

(AFD Limit) edu , at - (AFD Limit)n ive

+ (NSLOPE/3) x Margin " ] absolm vh (AFD 1.imit)#" p, g - (AFD Limit)Cp t ve (PSLOPE/8) x Margin " ] absolute value where Margin " is the miniir.um inargin from 4.2.2.2.c.1, and (2) Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , reset the AFD alarm setpoints to the inodified limits of 4.2.2.2.c,2.a, or (b) Comply with the ACTION requirements of Specification 3.2.2, treating the margin violation in 4.2.2.2.c,1 above as the amount by which Fo"^ is exceeding its limit.

mDefined and specified in the C01R pet Specification 6.9.1.9.

s 8-22 w___- _ _ _ - _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ __________

.. _ . . -. ._.m. . _ . . _ . . . . _ . _ _ . . _ _ . _ . . _ _ _ _ . _ . . - _ _ _ _ . _ . _ . . . . _ . . __ ._ . - . _ . . . _ -

y for-Specification 4.-2.2.2 i

Attachment 3 (con't): 1

3. Find the minimum RPS Hargin of all locations examined in 4.2.2.2.c.1 above. If any margin is less-than cero, then the following action shall be taken:

k'ithin 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , reduce the K3 value for OTAT by:

I"

- K3 *dJust *d . g3 (4) .

'KSWPE(2) x Ha@ "PS) R absolute value where MARGIN 3

is the minimum margin from 4.2.2.2.c.1. ,

i mDefined and specified in the COLR per- Specification 6.9.1.9.

(*)K3 value from Table 2.2 1.

l.

~

L

-.i

-4 8-23

if - .

for Specification 4,2.2.2 e

_ Attachment 4

@ . d. _

Extrapolating-the two most recent measurements to 31 Effective Full Power -Dg's beyond the most recent measurement and if:

-]'

' H L

[ Fn (X,Y, Z) , (extrapolated) 2- (Fn (X,Y,Z))oP (extrapolated), or

l. ' (Fn "(X,Y,Z)] iextrapolated) -2 [Fn k

(X,Y,Z))* (extrapolated),

t 4

either of-the fellowing actions shall be taken: ,

- 1., FnM (X,Y,Z) saall be increased by 2 percent over that specified in -

J 4.2.2.2.a. an ' the calculations ' of 4. 2.2.2. c repeated, or

I

l. ' 2 .~ A movable incore d=tector power distribution map shall.be E obtained,-and the calculations of 4.2.2.2.c.1 shall be-performed

' : no later than the time at which the margin in 4.2.2.2.c.1 is extrapola^ted to be equal to zero.

l

I I

eO j

..t l

n .

1 nt

!l --

l

~!

O 8-24

= y*,.<=w ..,,v+*

. _-.-- - - . . - . . - ~- - . - - - . . . - - . -

POWER DISTRIBUTION LIMITS r

SURVEILLANCE REQUIREMENTS (Continued)

.Tg . The -limits specified in Specifications 4.2.2.2c. , 4.2.2.2e. , end l gi ) 4.2.2.2f. , above are' not applicable in the following core plane n d ,. regions:

Ag&*M ' 11. Lower Core region:from 0 to 15%, inclusive

' 2. Upper core region from 85 to 100%, inclusive. _ ;

'4.2.2.3- -Base Load-operation is permitted at powers above APL # if the- following

-conditions are satisfied: -

a.- Prior to entering Base Load cperation, maintain THERMAL POWER above ND APL and less than or equal to that allowed by Specification 4.2.2.2 for at ,least .the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .- Maintain Base Load operation surveillance (AFD within the target band.about the terget. flux differ-gle.tc ence of Specification 3.2.1) during this time period. Base Load' operation is then permitted providing THERMAL POWER is maintained ND between APL $dAPLOL or between APL ND and 100% (whichever is most limiting) and FQ surveillar.ce is maintained pursuant to Specification

BL 4.'2.2.4. =APL is defined as:

p RTP APLOL = minimum g ;Q x K(Z) )x100%

over-Z

-F (Z) x W(Z)BL O where: F (z) is the measured Fq (z) increased by the. allowances'for

( manufacturing tolerances and measurement uncertainty. F RTP q

is the

-Fq_ limit,:K(z)~is-the-normalized Fq (Z) as a function of core height.

l W(z)gt is the cycle dependent- function tnat' accounts for: limited -power j distribution- transients encountered -during Base Load operation. '

RTP F

g , K(z), and W(Z)BL are specified.in the CORE OPERATING LIMITS ;

LREPORT per Specification 6,9.1.9.

( b .- Durin'g Base Load-operation, if the THERMAL POWER is decreased below ND

-APL then the' conditions of 4.2;2.3a shall be satisfied-before

-re-entering. Base _ Load operation.

'4.2.'2;4Ouring_ Base Load 0perationqF (Z) shall be evaluated to determine if-Fq (Z) is within-its limit by:

a; Using the movable incore detectors to obtain a power distribution map;at any THERMAL POWER above'APL ND ,

b. IncreasingJthe measured qF (2) component of the power distribution map by:3% to account'for manufacturing tolerances and further increasing.

the value by 5% to account-for-measurement uncertainties. Verify the requirements of Specification 3.2.2 are satisfied.

ND

APL is the minimum allowable-(nuclear design) power level for_ Base Load operation in Specification 3.2.1.

CATAWBA -UNITS 1 {d 3/4 2-7a -Amendment tio. 74 (Unit 1} -

8-25 -Amendment !?e. SS(Unit 2)

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

1 for Specification 4.2.2.2 j

l' s

Attachroent $:

j .- -

c. The limits in Specifications 4.2.L 2.c and 4.2.2.2.d are not applicable I: in the fol' lowing core plane regions as measured in percent of core i height from the bottom of the fuel:

i

[. 1. Lower core region from 0 to 15%, inclusive, i

Upper core region from 85 to 100%, inclusive, 2.

i.

l l

l s

i-8-26 l

POWER DISTRIBUTION LIMITS

/, -~ \

( SURt21LLANCE REQUIREMENTS (Continued) h Satisfying the followilg relationship:

RTP J)c.letc F (Z) $ -[ for P > APL ND M TP where: Fg (Z) is the measured Fq (Z). F is the F q limit.

K(Z) is the normalized Fq (Z) as a function of core height. P is the relative THERMAL POWER. W(Z) is the cycle dependent function that accountsforlimitedpowerdibributiontransientsencounteredduring Base Load operation. FhTP , K(Z), and W(Z)gg are specified in the CORE OPERATING LIMITS REPORT per Specification 6.9.1,9.

t d. Measuring F (Z) in conjunction with target flux difference deter-mination ac.crding to the following schedule:

1. Prior to entering Base Load operation after satisfying surveil-lance 4.2.2.3 unless-a full core flux map-has been taken in the previous 31 EFPD with the relative thermal power having been maintained above'APL ND for the 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to mapping, and

/

}

YM 2. At least once per 31 effective full power days,

e. With measurements indicating

-maximum (z) over I b K(z) 3 has increased since the previous determination bIgF (Z) either of the following actions shall be taken:

1-. F"(Z) q shall be increased by 2 percent over that specified in 4.2.2.4c, or 2; FqM (Z)-shall be measured at least once per 7 EFPD until 2 successive maps indicate that

-maximum F (z) is not increasing. '

over z K(z) 3

f. With the relationship specified in 4.2.2.4c above not being satisfied, either of the following actions shall be taken:

A-( 1 1. Place the core in an equ'ilibrium condition where the limit in 4.2.2.2c is satisfied, and remeasure F (Z), or CATAWBA - UNITS 1$4 3/4 2-7b -Amendment-Ho. 74 (Uni t 1) 8-27 emendment NO. 00(Li; 2)

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

5;

t r "

1--

+ _i

-POWER

DISTRIBUTION LIMITS--

l SURVEILLANCE-REQUIREMENTS (Continued)

-2.; Comply with the ~ requirenients of Specification- 3.2;2 for i heletc .

Fq (Z) exceeding i.ts limit by the percent calculated with

.the following expression:-

((max.-over.z of.[ -~F (Z) x W(Z)BL ) )-_ -I ] Nx 100 for Pj FfP x K(Z) '[

P d

, _ .g.

The limi ts speci fied in :4. 2. 2. 4c. , 4'. 2. 2. 4e. , and 4. 2. 2. 4f. 1 above are not applicable,in the-following core plan regions:

/' , 1. Lower core region Of-to 15 percent, inclusive.

- 2.. : Upper. core regionJ 85 tc~,100 percent,' inclusive.

2,2;5 When1Ff(Z))is measured:for reasons other than meeting the- requirementsh. d g ,{ of Specification;4.2;2.2 ?an overall measured F q(z);shal.1-be obtainedlfrom a?poweri j y-distribution-~ map <and~ increased'by.3% to account for manufacturing; tolerances- 4 7 and furtherJincreased by 5% to-account for measurement uncertainty. .!

a3 Acykes:wth1.

Acc4me,rc & j l;

i i,

M i

1 21 1

-i j

-,b

U -

3/4 2-7c Amemi ment-- N c. "(Unit it FCATAWBA-UNITb1kf- 8-26 f.mendment-Ho.SS(Unit.2)l 31# -+.+.y s. , m . .. - - , - w w?cw.-, m _ m-__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ . _-

_ .._______-.____._._.._.__________.__._......_._._____._.__.___.,-____..._m.

for Specification 4.2,2,3' .

Attachment 6: j 4.2.2.3 When a full core power distribution map'is teken for reasons other than meeting the- requirements of - Specification 4.2.2.2 -' an- overall- Fn"(X,Y,Z) shall.be determined, then increased by 3% to account for manufacturing F . tolerances, further increased by 5% to account for measurement uncertainty, and further increased by the radial-local peaking factor to obtain a maximum i local peaki This value shall be compared to the limit in Specification 3.2.2.

1 b

I

?G l.

i I'

l.

i i

l l-l 8-29

POWER DISTRIBUTION LIMITS 73 i )

DC SURVEILLANCE REQUIREMENTS __

4.2.2.2.1 The provisions of Specification 4.0.4 are not applicable.

4.2.2.2.2 Fxy_ shall be evaluated to determine ifq F (Z) is within its limit by:

a. Using the movable incore detectors to obtain a power distribution map at any THERMAL POWER greater than 5% of RATED THERMAL POWER,.

b. Increasing the measured F x component of the power distribution map by-3% to account for manufacturing tolerances and further increasing the value by '5% to account for measurement uncertainties, C

c. Comparing the F xy computed =(Fx ) obtained in Specification'4.2.2.2.2b.,

above to:

1) _The F xy limits f r RATED THERMAL POWERx(FRTP) for Ge appropriate measured core planes given. in Specification 4.2.2.2.2e. and f. ,

below,-and

2) The relationship:

,s -F *ERTP [1+0.2(1-P)],

x x i Where F l

is the limit 'or fractional THERMAL POWER operation express as a function of F xRTP and P is the fraction of RATED THERMAL _ POWER at which F xy was measured.

d. Remeasuring F xy according.to the following schedule:

1) When F is greater than the F,RTP limit for the appropriate. )

measured core ' plane but less than the F relationship, additional powerpistributionmapsshallibetaken

F *YC' compared Yto F*RTP and F xy either:-

a) Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding by 20% of RATE 0 THERMAL C

POWER or greater, the THERMAL POWER at which F was x

last determined, or b) At least once per 31 EFPD, whichever occurs first.

f Deleh Ilp \-

\_.)

CATAWBA-UNITj1&j 3/4 2-7e -Amc adme n t llc . 1^ (UaH-1-F 8-30 -Amendment No. 6 (Uni t 2)

POWER Di$1RIBUTION LIMITS SURVEILLANCE REOUIREMENTS (Cont-inued)

\ - . _ ,

2) When the F is less than or equal to the F limit for the x

appropriate measured core plane, additional ' power distribution raaps shall be taken and F compared to F E and F at least x

once per 31 EFPD.

R

e. The F,y' limits for RATED THERMAL POWER (F x ) shall be provided for all core planes containing Bank "0" control rods and all-unrodded core planes in a Radial Peaking Factor Limit Report per Specifica-tion 6.9.1.9;

f. The Fxy limits of Specification 4.2.2.2.2e., above, are-not applicable in the = following core planes regions as measured in percent of core height from the bottom of the fuel:

1) Lower core region from 0 to 15%, inclusive,

2) Upper core region from 85 tc 100%, inclusive, i p 1

! .3) Grid plane regions at 17.8 1 2%, 32.1 1 2%, 46.4 2%, 60.6 2 2%

\ and 74.9 2%, inclusive, and ,

4) Core plane regions within 2% of core height (1 2.88 inches) about the bank demand position of > the Bank "D" control rods.

g, -With F exceeding F ,- the effects-of F on Fq (Z) shall be evaluated xy

to _ determine ifq F (Z) is within its limits.

4.2.2.2.3 When F9 (Z) is measured for other than F xy determinations, an overall l

-measured qF (Z) shall be obtained from a power distribution map and increased by 3% to account for manufacturing tolerances and further increased by 5% to ount for measurement uncertainty. ..

3ejese l l

n (

CATAWBA-UNITS 1-(j 3/4 2-7f emendment-Hc.14 (UtrH-&)-

8-31 Amendtnent Nc. ' fUni; 2)

l l

POWER DISTRIBUTION LIMITS

[\ 3/4.2.3 -REAGTOR-000bANT-SYSTEM FLOW RATE ANE-NUCLEAR ENTHALPY RISE HOT CHANNEL F ACTOR - f:g-g)

LIMITING CONDITION FOR OPERATION F3 . 2. 3 The combination of indicated Reactor Coolant System total flow rate and D

, R shall be maintained within the region of permissible operation specified in

, the CORE OPERATING LlHITS REPORT (COLR) for four loop operation.

Where:

N F

Aeg/dce 3g R=

a.

w//4. RTP Fg [1.0 + MF3g (1.0 - P)]

(At6ht I D THERMAL-POWER ,

b* '

~ RATED THERMAL POWER c._ F H = Measured values of F g obtained by using the movable incore detectors _to obtain a power distribution map. The measured values of F g shall be used to calculate R since the figure specified in the COLR includes penalties for undetected feed- }

water venturi fouling of 0.1% and for measurement uncertainties

[ of 2.1% for flow and 4% for incore measurement of F g, V

d. FRTP= The Fh limit at RATED THERMAL POWER (RTP) specified in the COLR, and mfg = The power factor multiplier specified in the COLR.

APPLICABILITY: MODE 1.

ACTION:

WilhthecombinationofReactorCoolantSystemtotalflowrateandRwithin fa.

the region of restricted operation within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months reduce the Power Range l Neutron Flux-High Trip setpoint to below the nominal setpoint by the same amount (% R1P) as the power reduction required by the figure specified in the COLR.

.b. With the combination of Reactor Coolant System total flow rate and R within the region of prohibited operation specified in the COLR:

e 1. Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either:

p chunc 9, Restore the combination of Reactor Coolant System total flow a) rate and R to within the region of permissible operation, or

[m\

( j b) Restore the combination of Reactor Coolant System total flow

(' rate and R to within the region of restricted operation and comply with action a. above, or g -

3/4 2-9 Amendment-He, " (Unit 1)

CATAWBA - UNIT { 1 & j 8 - 3 ?. -Amendment No. SO(Ufn-t 2)

-l 7

l1 F Sor Specification 3)2,3 is > Attachment 11:-

t' .

i >

l :" - 3.2,3L Pg;(X,Y)Lshall be= limited by-imposing the following relationship n

- FallR (X,Y) $ FallR (X,Y)

H l

) ;>

h Where: 4 H

FAllR (X,Y)

._the maximum measured radial peak ratio as defined-in the CORE 0PERATING LIMITS REPORT ,

, L, '(COLR). ;

gr IJ- ~ ,

FAllR'(X , Y)!-- .tho' maximum allowable radial peak ratio as l h defined in.the COLR, g a,t : .3

's

,-..F T

I . . .

i f

! r a- s ;

, '1'.

d o.

It i

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h 1 ~i

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i 4 c.

aT 1 p

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8-33 ,

) :-

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.. . . _ , . . . . _ . . . , . . . . _ _ .; . .._.._m'

for Specification 3.2.3 Attachment 2:

gg1011:

With Pa(X,Y) exceeding its limit:

a. Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , reduce the allowable TilEM1AL POWER from RATED THERMAL POWER at least RRH%") for each 1% that FAHR M (X,Y) exceeds the limit, and

b. Within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months either:

1. Restore FAHR M (X,Y) to within the limit of Specification 3.2.3 for RATED THERMAL POWER, or

2. Reduce the Power Range Neutron Flux liigh Trip Setpoint in Table 2 . 2 - 1 a t le as t RRitt for each it that FAllRH (X,Y) exceeds that limit, and

c. Within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of initially being outside tha limit of Sp< aification 3.2.3, either:

1. Restore FoHR M (X,Y) to within the limit of Specification 3.2.3 for RATED TilEDIAL POWER, or

2. Perform the following actions:

Reduce the OTAT Kg term in Table 2.2-1 by at least TRll e) for t

(a) each it that FA11R (X,Y) exceeds the limit, and M

(b) Verify through incore mapping that FAllR (X,Y) is restored to within the limit for the reduced TilERMAL POWER allowed by ACTION a, or reduce TilERMAL POWER to less than 5% of RATED TilERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , u) RRil is the amount of TilERMAL POWER reduction required to compensate for L

each 1% that FAHR M (X,Y) exceeds FollR (X,Y), provided in the COLR per Specification 6.9.1.9.

(2)TRil is the amount of OTAT R 3 setpoint reduction required to compensate for

- each it that FollRM (X,Y) exceeds the limit of Specification 3.2.3, provided in the COLR per Specification 6.9.1.9.

8-34

._ __ m.

POWER DISTRIBUTION LIMITS CN 3/4. 2. 3 4GACTOR COOLANT SYST H-FtOW4AT~-A@ NUCLEAR ENTHALPY RISE HOT

.("'

CHANNEL F AC10R - Fas(e:r)

LIMITING CONDITION FOR OPERATION zncorpraced iro ^

MTION(Continuen O#"6 " #' y c) Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux - High Trip Setpoint to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ,

2. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of initially being within the region of prohibited 3 operation specified in the COLR, verify through incore flux mapping and_ Reactor Coolant System total flow rate comparison that the com-j l bination of R and Reactor Coolant System total flow rate are restored to within the regions of restricted or permissible operation, or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

d Dc/cLe

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m

POWER DISTRIBUTION LIMITS 5

LIMITING CON 0lT10N FOR OPERATION ACTION (Continued) , y ,d/o ,- c.a.]

dj.T Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced THERMAL POWER limit required by ACTION (b.1.c) and/or b.22, above; subsequent POWER OPERA-

-- TION may proceed provided that t+e-comb 4 cat 4cn-+f-R-and-4edleated-.

pyjgMp;'y)g eeeter Ceciant-Gysrtom-tetal -f-low-ee-te-eMdemonstrated, through incore flux mapping and-4eacton-Goohnt-Synem-t+t+M4ew-eate-- jf,,,,. e

-sompar4+on, to be within theleegens-of-eene4Meo-oe-peem4*4b4e aperat4on specified in the COLR prior to exceeding the following l THERMAL POWER levels:

/ f) -A-nennab 50% of RATED THERMAL POWER, a K) -A-nonneb 75% of RATED THERMAL POWER, and a p) Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of attaining greater than or equal to 95%

g*jg4 of RATED THERMAL POWER.

SURVEILLANCE REQUIREMENTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicable.

f4.2.3.2 The combination of indicated Reactor Coolant System total flow rate determined by process computer readings or digital voltmeter measurement and R shall be determined to be within the regions of restricted or permissible y operation specified in the COLR: (

O F/ " *' a. Prior to operation above 75% of RATED THERMAL POWER af ter each fuel

' 'M loading, and A%Mst 3 b. At least once per 31 Effective Full Power Days. _

((4.2.3.3 TheindicatedReactorCoolantSystemtotalflowrateshallbelerified r to be within the regions of restricted or permissible operation specified in the COLR at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the most recently obtained value of R, obtained per Specification 4.2.3.2, is assumed to exist.

4.2.3.4 The Reactor Coolant System total flow rate indicators shall be subjected to a CHANNEL CAllBRATION at least once per 18 months. The measurement instrumentation shall be calibrated within 7 days prior to the performance of the calorimetric flow measurement.

4.2.3.5 The Reactor Coolant System total flow rate shall be determined by precision heat balance measurement at least once per 18 months.

ZnCOryors\ t.Y lrt

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3/4 2-10 -Amendment-Ho. " (Uni t.1)

CATAWBA-UNITh1{} -Amendment-Nc. GS (Untt 2) 8-36

- ~ _ - _ _ _ . . _ -

for Specification 4.2.3.3 1

l Attachment 3:

6

< b. Measuring FallR H (X,Y) according to the following schedule:

1. Prior ' to operation above 75% of RATED THEPJiAL POWER at the beginning of each fuel cycle, and the earlier of:

I i

2. At Icast once per 31 Effective Full Power Days or 1

3. At each time the QUADRANT POWER TILT RATIO indicated by the excore

detectors is normalized using incore detector measureme"ts.

d 4

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8-37

t for Specification 4.2.3.2 i 1

i I

' Attachment' 4;. -)

4.2 ;1. 2 FAIIR M (X,Y) shall' be evaluated to determine whether . Fa(X,Y) .is within its limit by- i

a. Using the movabic incore detectors to obtain a power l< distribution map at any T}lEPXAL POWER greater than 5% of RATED TilERMAL POWER.

L p.

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I 9

a.

8-38

- ~ , , < s %, .l -J d. m4 . _ _ _ _ _ , , _ , . _ _ _ _ _ , , _ , , _ , , _ , , , _ _ , _ _ _

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

.i r _

! . for Specification 4.2.3~2. ( ,

a 7 Attachment $:-

m

c. Perfo_rming the following calculations:

- 1. For each location, calculate the 4 margin to the maximum allowable design as.follows:

}

4 4

4F.n t Margin - -(1 _ FAHR"(X,Y))_ x 100%

FAHR'(X ,Y) f

.e i

? .No-additional-uncertainties are required for FA11R"(X,Y), because FAHR'(X,Y) includes uncertainties. '

F E

2. Find the. minimum margin oftall= locations examined in .

4.2.3.2.c.1 above. If _ any margin is less than zero - t comply:with the ACTION requirements of. Specification

- 3.2.3.

-d. Extrapolating the two most'recent measurements to 31 F - ' EffectiveEFull power Days beyond the most recent measurement ,

and if:

FallRM . (extrapolated)' 2- FAHR' '(extrapolated) -

either of the following actions shall be taken:-

l5 -- 1.x 1 FAHR H (X,Y) . shallibe .: increased by 2 p'ercent over that t

]

b specified inl4 2.3 22.al: and: the _ calcul~ations of 3

J 4.2.3.2..c;repeatedi;or- _;

, n 2. A movableLincore' detector power' distribution map shall be obtained,.,and/theicalculations'ofi4.2.3,2,e shall.

- be- performed' no later than the time at which the

margin;in 4.2.3J2'.c.is extrapolatedito be equalsto-

~

.zero, y_- 7

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C' l

('sc POWER OlSTRIBUTION LIMITS i b N_j' 3/4.2,4 OUADRANT POWER TILT RATIO LIMITING CONDITION FOR OPERATION 3,2.4 The OVADRANT POWER TILT RATIO shall not exceed 1,02.fibove 50% of RAT

@HERMALPOWER, APPLICABILITY: MODE 1!N**

A ACTIONi

a. With the QUADRANT POWER TILT RATIO determined to exceed 1,02 but less than or equal to 1,09:

1. Calculate the QUADRANT POWER TILT RATIO at least once per hour until either:

a) The QUADRANT POWER TILT RATIO is reduced to within its limit, or b) THERMAL POWER is reduced to less than 50% of RATED THERMAL (3 i POWER.

() 2. Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either:

a) Reduce the QUADRANT. POWER TILT RATIO to within i',s limit, or b) Reduce THERMAL POWER at least 3% from RATED THERMAL POWER ,

for each 1% of indicated QUADRANT POWER TILT RATIO in 'l'

/, oa nd similarly reduce the Power Range ~ Neutron excess of'@ip Flux-High Tr Setpob,cs within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , . -

3; . Verify that the QUADRANT POWER TILT RATIO is within its limit:

within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding the limit or reduce THERMAL POWER to.less than 50% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High' Trip Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; and

4. Identify and correct the cause of the out-of-limit-condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified acceptable at 95%

or greater RATED THERMAL POWER, l \

l ^5ee Special Test Exceptions Specification 3.10.2.

]A -Q n ser c A ccae.h m en t I}

CATAWBA - UNITf 1 $ $ 3/4 2-12 8-40

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1 for Specification--'3.L 41 -a

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, *?Not-applicable.until calibration-of the excore detectors l's completed ji, m': subsequent to refueling! J l

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1 POWER DISTRIBUTION LIMITS

[ )) -

LlHITING CONDITION FOR OPERATION ACTION (Continued)

b. With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to misalignment of either a shutdown or control rod: ,

1. Calcuhte !.he QUADRANT POWER TILT RATIO at least once per hour :

until eithee: j a)- The QUADRANT POWER T!LT RATIO is reduced to within its limit, or b) THEkn,.. "WER is reduced to less than 50% of RATED THERMAL POWER. ,

i

2. Reduce THERMAL POWER at inst 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT f 0WER T'ILT RATIO in excess of i within 30 minutes; ,

,, 3. Verify that the QUADRANT POWER TILT RATIO-is within its limit-

% within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of ter exceeding the' limit or reduce THERMAL

, ) POWER to less than 50% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High Trip Setpoints to-less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; and

4. Identify and correct the cause of the out-of-limit condition 4 prior to increasing THERMAL POWER; subsequent POW'tR OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least

-once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified acceptable at 95%

or greater RATED THERMAL POWER.

c. With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to causes other than the misalignment of either 'a shutdown or control rod:

1. Calculate the QUADRANT POWER TILT RATIO at least once per hour l

until either:

a) The QUADRANT POWER TILT RATIO is reduced to within its limit, or

i. b) THERMAL POWER is reduced to less tht:' 50% of RATED THERMAL l POWER.

( )

v CATAWBA-UNITh1kk 3/4 2-13 8-42

)

L .. _

.I

POWER OlSTRIBUTION LIMITS

. LIMITING CONDITION FOR OPERATION ACTION (Continued)

2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High Trip Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; and

3. Identify and correct the cause of the out of-limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATEC THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified at 95% or greater

-RATED THERMAL POWER. ,

d. The proAsions of Sp Mic tton 3. c .y m ne t opgic uc. l SURVEILLANCE REQUIREMENTS 4.2.4.1 The QUADRANT POWER TILT RATIO shall be determined to be within'the '

limit above 50% of RATED THERMAL POWER by:

a. Calculating the ratio at least once per 7 days when the alarm is t OPERABLE, and

b. Calculating the ratio at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during steady state operation when the alarm is inoperable.

-c. The provisions of Specification 4.0.4 are not applicable.

4.2.4.'2 The QUADRANT POWER TILT RATIO shall be determined to be within the limit when-above 75% of RATED THERMAL POWER with one Power Rango channel inoperable by usi.ng the novable incore detectors to confirm that the normalized symmetric power distribution,~obtained from two sets of four symmetric thimble locations or full-core flux map, is consistent wi h the indicated QUADRANT POVER TILT RATIO at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

CATAWBA-UNITj1& 3/4 2-14 4-endment-No . :.d (Un i t 1) 8-t43

_ __-a_ _ _ _ - _ _ - -

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- ~

POWER DISTRIBUTION LIMITS 3/4.2.5 DNB PARAMETERS LIMITING CONDITION FOR OPERATION _ , _ _

3.2.5 The following DNB related partmeters shall be maintained within the limits shown on Table 3.2-1:

a. Reactor Coolant System T,yg. e,J-

b. Pressurizer Pressure, m

APPLICASigT'(: M0dE 1. \c feac tor C e ls.i y s te "- B "I R" O c^t c .

ACTION: .

nide 10: ^O 0 I ^ t .L . ". wa b. dove 0.. With any-of the above-parameters ^ exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce TfiERMAL POWER to less than 5% of RATED THERMAL POWER within 1.he next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

[Ad& Tncer + Q), n +%cV ed )

w._ _ _ . _ . _ _ -

SURVEILLANCE REQUIREMENTS

,I 4.2.5^ Each of the parameters of Table 3.2-1 shall be verified to be within their limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

~ . . . _ _.

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CATAWBA-UNITj14{ 3/4 2-15 R-44

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,af With the combination of Reactor Coolant System total flow rate and R within

b. the region of restricted operation *within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months reduce the Power Range Neutron Flux-High Trip Setpoint to below the nominal setpoint by the same amount (% RTP) as the power reduction required byphe-f4tJttre-spectftetMit 4he-GOL& r,w cc 5. 2 -l .

> r at er. 4 E.' M t.

,b'. With the combination of Reactor Coolant System total flow rate and % within C. the region of prohibited operation specified,jtn-the-00tR:

,, c a n cm -/l 1, Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either:

a) Restore the combination of Reactor Coolant System total flow rate and g o p hjg,the region of permissible operation, or b Restore the combination of Reactor Coolant System total flow

-7/in.r,tt c)o,dRFKEM- to within the region of restricted operation and comply with. action a. above, or c) Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux - High Trip Setpoint to less than or equal to 55% of RATED THERMAL POWER within i the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . y pjg 3, y _ (

2. -Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of initially .being within the region of prohibited

.[

y j\ operation specifiedin-the COL *, verify -thnough4ncore-f4w-mappHtg+

anAReactor-Goolant-System-total-f-low-rate-comparisort that the com-

"W L

f'bination ofM and Reactor Coolant System total flow rate are restored to within the regions of restricted or permissible operation, or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

IMed @

4c2.34. Tne Reactor Coolant System total fiow rate indicators snall be subjected to a CHANNEL CALIBRATION at least once-per 18 months. The measurement instrumentation shall be calibrated _within 7 days prior to the performance of the calorimetric flow measurement.

'4.2 Id The Reactor Coolant System total flow rate shall be determined by precision heat balance measurement at least once per 18 months.

g--

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L/

8-45

l 1

s 4

g TABLE 3.2 1 DNB PARAMETE_R,,5 i

PARAMETER LIMIT $

Four Loops in Operation Average Temperature

' Meter Average - 4 channels: < 592*F

- 3 channels: {592'F Computer Average - 4 channels: < 593'F

- 3 channels: [593'T -

Pressurizer = Pressure Meter Average - 4 channels: > 2227 pi,ig*

- 3 channels: -[2230psig* i Computer Average - 4 channelst > 2222 psig*

- 3 channels: [2224psig*

R, cf or Ceel.4 ' Sdim 74l Fl.w Eale F;7tc. 3.7 1

't b

i I'

l:

p

Limit' not applicable during either a THERMAL POWER ramp in excess of 5% of RATED THERMAL POWER per minute or a THERMAL POWER step in excess of 10% of

~._

\ . SATED THERMAL POWER.

t CATAWBA-UNIT $1_$$~ .3/4 2-16 -Amendment-HoH4-(Unit it 8 t. 6 **M =nt "r ::.f Urit 25-

-. . . . . . . . - . , . . - . . - . - , . . . - . , , . - . . - . . - . . - , - - . , ~ , - - . , . , - . .. - ,, ,-.. .-a

1 1

I Tigure 3.2 1, Reactor Coolait System Total rio.' Rate Versus f@ Ratw; Therital Power Tour Loops in Operation

! l 388850 --

A Pehb!!y of 0.1% for undetected feee.ater venturl touting and a Permissible ,

rnessurernent uncertaint, et p.1% for Operation oo .re ineiud.e in w . n oure. Region l 383000 . .. . . . . . . . . . . . . . . .. . . . . . . . . .. . . . .. . . . .. . . . . . . . . . . . . . .. J S 83.e.s _

0).

Restrleted Operation (08881160)

E 381150 j ,

Region

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Operation

.$N (90,369600) Region 4 y369600-

'365750 361900 , , , , , ,

86 88 90 92 94 96 98 100 102 !

[

-k Fraction of Rated Thermal Power CATAk'BA UNIT 1 8-47

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

i l .

EMERGENCY CORE 000 LING SYSTNS l

l EURVE!!.U,NCE RE0VIREVENTS (Continuee)

h. By performing a flow balance test, during shutdown, foll: wing com-pletion of modifications to the ECCS subsystems tnat alter the subsystem flow charseteristics 4.nd verifying that:

1) For cent?ifugal charging pump lines, with a single pump running:

a) The sum of the injection line flow rates, exclueing the highest flew rate, is greater than or equal to -ES-SU gp.u, and b) The total pump flow rate is less than or equal to 565. gpm.

2) ist Safety Injection pump lines, w th a single pump running a) The sum of the injection line flow rates, excluding the

highest flow rate, is greater than or equal to 46 YM gpm, ana c) The total pump flow rate is less than or equal to 660 gpm.

3) For residual heat removat pump lines, with a single pump running, the sum of the injection line flow rates is greater than or equal to 3648 gpm.

(

4 l

l CATAWBA - UNIT $ 1 & 2 3/4 5-8 8-48

i 344.2 POWER DISTRIBUTION LIMITS n

[V \

BASES __

The specifications of this section provide assurance of fuel integrity during Condition 1 (Normal Operation) and 11 (Incidents of Moderate Frequency) events by: (1) maintaining the calculated DNBR in the core greater than ce equal to design limit DNBR during normal operation and in short-term transients, and (2) limiting the fission gas release, fuel pellet temperature, and cladding mechanical properties to within assumed design criteria. In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria'34mR-of-BB00EF-4+- not exceeded.

Q%

The definitions of certain hot channel and peaking f actors as used in these specifications are as follows vece #e w h $ $

F (Z) Heat Flux Hot Channel Factor, is defined as the phlocal heat 0- flux on the surf ace of a fuel rod at coreleie*M4en-Z divided by the everage fuel rod heat flux, allowing for manufacturing tolerances on

{L.- Q- y fuel pellets and rods; s

F g, Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of si the integral of linear power along thefrodv:!th the H$.en tntsontad.

(Q Ay)f' .powan to the average rod power. (J7W7 ora TYe n L T)

(V / ~KN is e ned o tk nelp 6 W U"ot b <- f u co' c Mc[

3/4.2.1 AXIAL FLUX DIFFERENCE g,,.g p4 (xg)

_ The limits on AXIAL FLUX OIFFERENCE (AFD))'-nwee3thattheF(ZN"EE** 0 bound-enalopelef-the F RTP 1M ecified in S CORE OPERATING LIMITS REPORT]

((COLR)feme the normali:cd =i01 ):: king f actc&4+ not exceeded during either normal operation or in the event of xenon redistribution following power changes.V - - ---

( Target flux difference is determined at equilibrium xenon conditions.

Ths full-length rods may be positioned within the core in accordance with their respective insertion limits and should be inserted near their normal position for steady state operation at high power levels. The value of the target flux difference obtained under these conditions divided by the fraction

. of RATED THERMAL POWER is the target flux difference at RATED THERMAL POWER 4 for the associated core burnup conditions. Target flux differences for other

{THERMALPOWERlevelsareobtainedbymultiplyingtheRATEDTHERMALPOWERvalue by the appropriate fractional THERMAL POWER level. The periodic updating of 4

{ithetargetfluxdifferencevalueisnecessarytoreflectcoreburnup

\ponsiderations - - -

Dele t.e w

ud Me Fu (A Y ) liW ts q h AFD ence.lys. geclOcd i., Mc CoLR, hvo been oMj W he ~~

rn en:, v n .m nt. unu -Lont.y. _f CATAWBA-UNITj1$( B 3/4 2-1 -Amendment-Ho (Sni t-it 8-49 Waendment Ne, OG (Unit 2)

I P_0WER DISTRIBUTION LIMITS i

o BASES At power levels below APL"0 , the limits on AFD are defined in the COLR, l 1.e. , that defined by the RAOC operating procedure and liinits. These limits were calculated in a manner such that expected operational transients, e.g.,

load follow operations, would not result in the AFD deviating outside of those limits. However, in the event such a deviation occurs, the short period of time allowed outside of the limits at reduced power levels will not result in signia ficant xenon redistribution such that the envelope of peaking f actors would ND change sufficiently to prevent operation in the vicinity of the APL power level.

At power levels greater than APL ND , two modes of operation are permis-sible; 1) RAOC, the AFD limits of which are defined in the COLR, and 2) Base i Load operation, which is defined as the maintenance of the AFD within a COLR specified band about a target value. The RAOC operating procedure above ND is the same as that defined for operation below APL HO However, it is APL ,

possible when following extended load following maneuvers that the AFD limits may result in restrictions in the maximum allowed power or AFD in order to guarantee operation with F (z) less than its limiting value. To allow operation 9

at the maximum permissible value, the Base Load operating procedure restricts

.Z>e/ete CATAWBA-UNITk1hk B 3/4 2-2 Amendment-Ho. 7' (Utt4-1+-

8-50 Amendment-k. SB @n%

-__ '~~i_____________

( -

1 i

[

L POWER DISTRIBUTION LIMIT $

9

BASES ,

l

l 8

3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR, and REACTOR COOLANT SYSTEM-FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Cor % d)--

the indicated AFD to relatively small target band and power swings (AFD target i

A'O band as specified in the COLR, APLNU < power < APLOL or 100% Rated Thermal Power, l whichever is lower). For Base Load operation! it is expected that the Units will i operate within the target band. Operation outside of the target band for the short time period allowed will not result in significant xenon redistribution such that the envelope of peaking f actors would change suf ficiently to prohibit continued operation in the power region ~ defined above. To assure there is no

-residual xenon redistribution impact from past operation on the Base Load

. operation, a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months waiting-period at a power level above APL ND and allowed l by RAOC is necessary. During this time period load changes and rod motion are restricted to that allowed by the Base Load procedure. After the waiting i period extended Base Load operation is permissible.

The computer determines the one minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for at least 2 of 4 or 2 of 3-0PERABLE excore channels are: 1) outside the allowed al power operating space (for RAOC operation), or 2) outside the-allowed al target band (for Base Load operation). These alarms are active when power is greater than: 1) 50% of RATED THERMAL POWER (for RAOC operation), or

2) APLND (for Base Load operation). Penalty deviation minutes for. Base Load j

, operation are not accumulated based on the short period of time during which j

, gration outside of the target band is allowedy m., The limits on heat flux hot channel factor e-coolent '1 e rate,.and nuclear not,

<x""

g enthalpy rise hot channel- factor ensure that: . (1) the-design limits on peak--

local power density and minimum DNBR are not exceeded and (2) in the event of-a LOCA _the peak fuct c44d-temperature d11 r.ct-ex:::d the 2000*F ECCS acceptance

- criterial44m4_ heeeMimits are specified in the CORE OPERATING LIMITS REPORT (cola)

l. per Specification 6.9.1.9. {7ae pattog -

k & D h of th::: i: measurd.1& out will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3. This periodic i- surveillance is suf ficient to_ insure that the limits are maintained provided:

a. Control rods in a-single group move together with no individual rod -

insertion differing by more than i 12 steps, indicated, from the L group demand position; .

b. Control rod groups are sequenced with overlapping groups as described in Speci fication 3.1.3.6; TS hat %. he channi (hto r- cmd nudeu-en M td y nWe. h 4 chuw/ futw w e- each l

i CATAWBA-UNIT (1{} B 3/4 2-2a -Amendment "c. 74 (Unit 1) 8-51 -Amendment-No-68-(4)n+t-0-b

.. .-. _~ . _ _ . - - - . - -- -

POWER DISTRIBUTION LIMITS

\

b BASES HEAT FLUX HOT CHANNEL FACTOR, and REACTOR COOLANT SYSTEM FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued)

c. The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained; and "

l in am d*se P ".5

d. l The axial power OlFFERENCE, is maintaineddistribution, expressed within the limits. in terms Qo.5 of AX1AL j

F ill be maintained within its limits provided Conditions a through d. ]

abovearemaintained.[AsnotedonthefigurespecifiedintheCOREOPERATING

N LIMITS REPORT (COLR), Reactor Coolant System flow rate and F g may be " traded off" against one another (i. ., a low measured Reactor Coolant System flow rate

isacceptableifthemeasuredFfH is also low) to ensure that the calculated f DNBR will not be below the design DNBR value. The relaxation of F as a H

function of THERMAL POWER allows changes in the radial power shape for all permissible rod insertion limits.

^

! j R as calculated in Specification 3.2.3 and used in the figure specifiec 6/ in the COLR, accounts for Ffg less than or equal to the F limit specified i the COLR. This value is used in the various accident analyses where F

3g influences parameters other than DNBR, e.g. , peak clad temperature, and thus is the maximum "as measured" value allowed. The rod bow penalty as a function of burnup applied for F H is :alculated with the methods described in WCAP-8691, Revision 1, " Fuel Rod Bow Lvaluation," July 1979, and the maximum rod bow penalty is 2.7% DNBR. Since the safety analysis is performed with plant-specific safety l

@DNBRlimitscomparedtothedesignDNBRlimits,thereissufficientthermal rgin available to offset the rod bow penalty of 2.7% DNBR.

( The hot channel factor FM (z) is measured periodically and increased by a

' 9 cycle and height dependent power factor appropriate to either RAOC or Base load operation, W(z) or W(z)gt, to provide assurance that the limit on the hot channel factor, F (z), is met. W(z) accounts for the of fects of normal oper-9 fationtransientsandwasdeterminedfromexpectedpowercontrolmaneuversover the full range of burnup conditions in the core. W(z)gg accounts for the more i restrictive operating limits allowed by Base Load operation which result in l

less severe transient values. The W(z) function for normal operation and the W(Z)BL function for Base Load Operation are specified in the CORE OPERATING

_j (O(LIMITSREPORTperSpecification6.9.1.9.

cp>la c e with A ttuu, men t. ,Q CATAWBA-UNITj1pq B 3/4 2-4 Amendment-No-4(4Jnit-1-)-

-Amendment-Ho--68-(4Jtrit-et-S-52

for loser Distribation Lirlts Pases Attachment 1:

7he limits on the nuclear enthalpy rise hot channel f actor, Ttit(X,Y), are epo:ified in the CDLR as !bximam A110 sable Padial reaking lirits, obtairr.d by dividing the !b>: imam Allosable Total leaking (!%P) lirlt by the axial peak (AXIAL (X,Y)) for location (X,Y) . Lef dofinition, the luximam A11cuablo Padial Peaking limits will, for !hrk-EU fuel, result in a Dilm for the lirlting transient that is cqaivalent to the u!IR calculatcd with a design r tt(X,Y) e value of 1.S5 and a limiting referenac axial roscr chape. The lurk-EM !%P limits my be applicd to OTA fuel, pre /idcd an appropriato adjust cnt factor is applicd to pre /ido cqaivalence to a 1.49 design Tati(X,Y) for the OTA.

This is reflected in the 1%P lirits spo:ificd in the ODIR. The relaxation of P![(X,Y) as a function of 'ntDWsL IOitR allcus changea in the radial rosor for all termissible control tank insertion limits. This relaxation is t

implementtd by the application of the folloaing factorm:

k = (1 + (1/FJUI) (1 - P))

store k = roser factor multiplier appl. icd to the !%P lirits P = 7HLTMsL I%TR / }&Tl:D 'IMD?%L 10@

1001 is given in the CDLR S-53

l for Power Distribution Limits !sases l

Attachment 2:

r"(X,Y,Z) n and TAllPM (X,Y) are measured periodically, and comparisons to the allowable limit are made to provide reasonable assurance that the limiting criteria vill not be exceeded for operation within the Technical Specification limits of Sections 2.2 (Limiting Safety Systems Settings), 3.1.3 (Movable control Assemblies), 3.2.1 (Axial Flux Difference), and 3.2.4 (Quadrant Power Tilt Ratio). A peaking margin calculation is performed to provide a basis for decreasing the width of the AFD limits, for reducing the K3 value from OTAT, and f or reducing THEPJiAL P0kTR.

k' hen an Tn M (X,Y,Z) measurement is obtained in accordance with the surveillance requirements of Specification 4.2.2, no uncertainties are applied to the measured peak; the required uncertainties are included in the peaking limit. k'he n TnM (X,Y,Z) is measured for reasons other than meeting the requirements of Specification 4.2.2, the measured peak is increased by the radial local peaking factor to convert it to a local peak. Allowances of 54 for measurement uncertainty and 36 for manufacturing tolerances are then applied to the measured peak, k'he n an TAllR"(X ,Y) measurement is obtained, regardless of the reason, no uncertainties are applied to the measured peak; the required uncertainties are included in the peaking limit.

l 8-54

} .

POWER DISTRIBUTION t1MITS t w p oted W

$3a c t, f u .- spec3#hte 5 BASES -

HEAT FLUX HOT CHANNEL FACTOL and REACTOR COOLANT SYSTEM FLOW RATE AND NUCLEAR INTHALPY RISE HDT CHANNEL FAcil5R (continueo)

WhenReactorCoolantSystemflowrateandFharemeasured,noadditional }

allowances are necessary prior to comparison with the limits of the figure

$pecified in the COLR. Measurement errors of 2.1% for Reactor Coolant System total flow rate and 4% for Fh have been allowed for in determination of the design DNBR value.

The measurement error for Reactor Coolant System total flow rate is based upon performing a precision heat balance and using the result to calibrate the Reactor Coolant System flow rate indicators. Potential fouling of the feedwater venturi which might not be detected could bias the result from the precision heat balance in a nonconservative manner. Therefore, a penalty of 0.1% for undetected fouling of the feedwater venturi is included in the figure specified in the COLR. Any fouling which might bias the Reactor Coolant System flow rate measurement greater than 0.1% can be detected by monitoring and trending vcrious plant performance parameters. If detected, action shall be tnken before per-forming subsequent precision heat balance measurements, i.e., either the effect of the fouling shall be quantified and compensated for in the Reactor Coolant System flow rate measurement or the venturi shall be cleaned to eliminate the fouling.

The 12-hour periodic surveillance of indicated Reactor Coolant System flow is sufficient to detect only flow degradation which could lead to opera-tion outside the acceptable region of operation specified on the figure spec- '

Q fied in the COLR. ,

3/4.2.4 QUADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that the radial power distribu-tion satisfies the desi0n values used in the power capability analysis.

Raiisi power distribution measurements are made during STARTUP testing and per Mdically during power operation.

h3f" The ibit of 1.02, at which corrective action is required, provides ONB Atb w Bind linear _ heat oeneration rate _ protection with x-v Diale power tiltL f A P imTt of 1.02 was selected to provide an allowance for the uncertainty associated ith the indicated p_ower tilte The 2-hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.C9 is provided to allow identification and correction of a dropped or misaligned control rod. In the event such action does not correct the tilt, the margin for uncertainty on s reinstated by reducino ,

the maximum allowed power by 3% for each percent o tilt in excess of @

For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the movable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO. The incore detector monitoring is done with a full incore CATAWBA - UNITS 1h{ B 3/4 2-5 -Amendment-HO. M (i.mit-1-)-

b55 -Amendment-No-Hunitt)-

i l

t i

for kwer Distribution Limits Ibscs Attacitw.nt 31 i

i A peaking increase that reflects a QUAT %VNr 1%TR TILT PJd20 of 102 is j

includod in the generation of tho /ED limits.

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(3 POWER 01STRIBUT10N LIMITS }

(v) BASES 1

OVADRANT POWER f!LT RATIO (Continued) flux map or two sets of four symetric thimbles. . The tw0.JLttrLgLfour symtttric de/M thimbles is a unig !

C;87F5N! 11~~R3,vj_s.e1 H-13. L-5, of eight._ detector locationi[The L-if,~F87ATte7nate locations are normal

~

availablelocations

< cr if any of_the normal _, locations are unavailable.f 3,,44.2.5 DNB PARAMETERS, (Re vice Lf at L d . c a b el) l The limits on the ONB-related parameters assure that each of the parameters are maintained within the normal steadestate envelope of operation assumed in the transient and accident analyses. The limits are consistent with the A nd Lw + l .

initial FSAR assumptions and have been analytically demonstrated adtguate to/ {

maintain a design limit DNBR throughout each analyzed transient.(UThe indicated value and the indicated pressurizer pressure value correspond to analytical T,yftsof5948'Fand22053psigrespectively lim . . , with allowance for measurement uncertainty. @ Add InscM r_ j Ah4J The 12-hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their

[m\ limits following load changes and other expected transient operation. Indica-

\'j tion instrumentation measurement uncertainties are accounted for in the limits provided in Table 3.2 1.

C b 51rt !

N

-F Wwi1-1-be-maintained-within-its-limits- we. 1,2 -rovided-Condi t-ions-a,-through-d,--

-abovea re-maintainedr- As noted on be-figure-specified-in-t CQREgRATjg p

L-1MI-TS-REPORT-(COLRh Reactor Coolant System flow rate and {gmay be " traded off" against one another (i.e. a low measured Reactor Coolant System flow rate TMJ ML Fwl !

is acceptable if the measured N is also low) to ensure that the calculated N

t DNBR will not be below the design DNBR value. -The-relaxation-of T g ,,_, _

-function-of-THERMAL-POWER-allows-changes-in-the-radial power-shape-for-all-

- pe rm i s s i bl e-ro d --i n s e rti o n -1 i m i ts , -- ~7h re. /n &nship drE,scd onyF'M i

rema,n: e lel u, Ic, . , ci ; He l : <. 4 c p lac e ct on Mc noch n Cn ki py r ; L e ho + ils a n n e l lh ed er, I~l,in Cyce[-: 4,7373 i arc m c,s L . n e cl .

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8-57

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[ncerr 2 fr,s e r t of @ u. p rc cet elinj D o3t .

When Reactor Coolant System flow rate-end-MIe- AH no additional 3,t.

measured,gw,c 5 : n,, 2 allowances are necessary_ prior to comparison with the limits of the-f4 9 eet-specM4ed-tn-the-CMR,-EMeasurement error ( of 2.1% for Reactor Coolant System tot al flow rate -and-45-for-f g 2fv'ebeenallowedforindeterminationofthe design DNBR value.

The measurement error for Reactor Coolant System total flow rate is based upon performing a precision beat balance and using the result to calibrate the

Reactor Coolant System flow rate indicators. Potential fouling of the feedwater ,

venturi which might not be detected could bias the result from the precision, ,

heat balance in a nonconservative manner. Therefore, a penalty of 0.1% for s undetected fouling of the feedwater venturi is included e in the-f4gure-spee+fia,vre f4fd:'- :

--in-the-00LRr- Any fouling which might bias the Reactor Coolant System flow rate !

measurement greater than 0.1% can be detected by monitoring and trending various plant performance parameters. If detected, action shall be taken before per-forming subsequent precision heat balance measurements,- i.e. , either the ef fect !

of the fouling shall be quantified and compensated for in the Reactor Coolant

$ystem flow rate measurement or the venturi shall be cleaned to eliminate'the

-/

O fouling.

<_) -

i i

1 l'

i 8-58 a

/ ADMINISTRATIVE CONTROLS t

\ SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT (Continued)

The Radioactive Effluent Release Reports shall include a list and description of unplanned releases from the site to UNRESTRICTED AREAS of radio-active materials in gaseous and liquid effluents made during the reporting period.

The Radioactive Effluent Release Reports shall include any changes made during the reporting period to the PROCESS CONTROL PROGRAM (PCP) and to the OFFSITE DOSE CALCULATION MANUAL (00CM), as well as a listing of new locations for dose calculations and/or environmental monitoring identified by the land use census pursuant to Specification 3.12.2.

MONTHLY OPERATING REPORTS 6.9.1.8 Routine reports of operating statistics and shutdown experience, in-cluding documentation of all challenges to the PORVs er safety valves, shall be submitted on a monthly basis to the U.S. Nuclear Regulatory Commission.

Attn: Document Control Desk, Washington, D.C. J0555, with a copy to the NRC Regional Vffice, no later than the 15th of each month following the calendar month covered by the report.

CORE OPERATING LIMITS REPORT f

6.9.1.9 Core operating limits shall be established and documented in the CORE

OPERATING LIMITS REPORT before each reload cycle or any remaining part of a

( reload cycle for the following:

1. Moderator Temperature Coefficient BOL and EUL limits and 300 ppm surveillan e limit for Specification 3/4.1.1.3,

2. Shutdown Bank Insertion Limit for Specification 3/4.1.3.5,

3. Control Bank Insertion Limits for Specification 3/4.1.3.6,

4. 'AxialFluxDifferenceLimits,targetbandfandAPLNDior Specification 3/4.2.1,

5. Heat Flux Hot Channel Factor, F RTP K(Z),W(Z)$# PLN T nd W(Z)BL forSpecification3/4.2.3,a,nd gaggpyf

6. NuclearEnthalpyRiseHotChannelFactor,F[H , nd Power Factor Multiplier, _ MF[g*,Nimits for Specification 3/4.2.3.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by NRC in:

1. WCAP-9272-P-A, " WESTINGHOUSE RELOAD' SAFETY EVALUATION METHODOLOGY,"

July 19M (W Proprietary).

j

/G (Methodology for ? pecifications 3.1.1.3 - Moderator Temperature N -

} Coef ficient, 3.1.3.5 - Shutdown Bank Insertion Limit.

[6aeif

(,pakof /

3.1.3.6 - Control Bank Insertion Limits, 3.2.1 - Axial Flux y

CATAWBA - UNITS 1 & 2 6-19 Amendment-Her?4-(-Un44-1-)-

8-59 -Amendment-He-684Urt44-2-)-

l, -i i

for Specification 6.9.1.9 t

l[ Attachment 1:

?

Reference 5 is not applicable to target band and APL*,

References 5 and 6 are not applicable : V(Z). APL*, and W(Z)g.

Reference 1 is not applicable'to TAHRL. l Reference 5 is not applicable to lif and MPa. j 4

t k

i s s .

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8-60

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

ADMINISTRATIVE CONTRCLS O __

a' CORE OPERATING LIMITS REPORT (Continued)

Difference, 3.2.2 Heat Flux Hot Channel Factor, and 3.2.3 -

Nuclear Enthalpy Rise Hot Channel Factor.)

2. WCAP-10216-P-A, " RELAXATION OF CONSTANT AXIAL OFFSET CONTROL FQ

> SURVEILLANCE TECHNICAL SPECIFICATION," June 1983 (W Proprietary).

(Methodology for Specifications 3.2.1 - Axial Flux Difference (Relaxed Axial Offset Control) and 3.2.2 Heat Flux Hot Channel Factor (W(2) surveillance requirements for F gMethodology.)

3. WCAP-10266-P-A Rev. 2, "THE 1981 VERSION OF WESTINGHOUSE EVALUATION MODEL USING BASH CODE," March 1987, (W Proprietary).

(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)

T (he e.g.core , fueloperating thermal-mechanical limits shall limits, be determined so that all applicable core thermal-hydraulic limits, ECCS limits f lits, nuclear limits such as shutdown margin, and transient and accident

.ialysis limits) of the safety analysis are met.

,L ,

/ The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or

( supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and 7 Resident Inspector.

\

\fs e. r ?" A r'/; red ri.YW [ E )

_ . . ~

l l

V) l CATAWBA - UNITS 1 & 2 6-19a Amencrent-Ho 2A-(Uni t 1) l 8-61 Amencment-No 6kunit-E-)-

1

l s >

j for Specification 6.9.1.9 I

Attachment 2:

-4. BAW 101$2. A, NOODLE A Multi Dimensional Two Group Reactor Simulator," ;

June 1985.

(Hethodology for Specification 3.1.1.3 . Moderator Temperature Coefficient.) !

$. BAW 10163P.A, " Core Operating Limit Methodology for Vestinghouse. I Designed PWR's.* June 1989.

(Methodology for Specifications 3.1.3.5 Shutdown Rod Insertion a Limits, 3.1.3.6 Control Bank Insertion Limits, 3.2.1 Axial 4 Flux Difference, 3.2.2 Heat Flux Hot Channel Factor, and 3.2.3 - '

LNucicar Enthalpy Rise Hot Channel-Factor.) .

6. BAW.10168P.- Rev.1, "B&W Loss of. Coolant Accident Evaluation Model for l'

Recirculating Steam Cenerator Plants," September, 1989. j

-(Mothodology for Specification 3.2.2

~

Heat Flux Hot Channel Factor.) il I

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l PRE XNARY Cat %ta Unit 1 Coro orcratirn Limits Report 2.2 Shutdwn Red Inc,grtion Limit (Sprification 3/4.1.3.5) l 2.2.1 7he chutdwn rtds shall to withdrawn to at 1 cast 222 steps.

2.3 fontrol Red Igpat. ion Limits (Spo ification 3/4.1.3.0) 2.3.1 7ho control red tanks r. hall to limitcd in physical incertion as r-hwn

in rigure 2.

I i 2.4 M:ial Flyy DiffqrgIgg (Specification 3/4.2.1) 4 j 2.4.1 7ho M:IAL TLUX CIITTRDiCC (MD) Limits are providcd in Figuro 3.

negade Am IM frm Fi@ro 3.

(MD Li.mit)n 9 tivolo cMuumit>mjk,athemuuve mu 1we fr r1,re 3.

l l

8-64

...$ 1Y. % ,, R '

O Catasta 1. Cycle 6 Core Operatin:: Limitt Report (Fully withdrawn) 220 (281,222) (7M.222) 200 -

BANK B 180 -

- (0%,162) (1001,1N 160 y BANK C

$ 140 -

5 is E 120 -

O!,e 100 a

} BANK 0 IC 80 -

E E -

cr 60 (01,47) 40 l

20 (30%,0) 0 i i e i 0 20 40 60 80 100 (Fully inserted) Relative Power (Percent)

Figure 2 Control Rod Bank Insertion Limits vs. percent RATE 0 THERMAL POWER i

8-65 1.-

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-) , ...

.4 : .} .g(e L :. )

> % , . t o ,,.t .}. e .~

<a L. ~

04. 4xbc 1 Cycle 6 Core Operating Lit.its Report

c. ;i 5

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(-12.97,100) 100 . s I: (10,87,100) l'NACCEPTABLE UNACCEPTABLE OP ERA'II ON OPERATION BD ACC EPT ABLE OPERATION 60 -

50

(-21.87,50) (22.37,50) 40 -

20 -

0 I l i 1 1 1 l I

-50 -40 ~30 -20 -10 0 10 20 30 40 50 FIGURE 3 Axial flux Difference Setroints As A runction of RATED THERMAL P0h'ER B-66

a f, '

i P.REUMNARY :

Catasto Unit 1 Core Operating Limits Report (Continuod) 2.5 Heat Flux Hot ChAGpol Factor (Specification 3/4.2.2) 2.5.1 Fn" = 2.32 2.5.2 K(Z) is provided in Figuro 4 for Park-IM fuel.

l 2.5.3 K(Z) is providcd in Figurc 5 for OTA fuel, t;

The follwing }nrareters are required for core tonitoring por the Surveilhnce Roquirer nts of Specification 3/4.2.21 2.5.4 (F[(X,Y,Z))* is ptwidod in Table 1.

l khcro (F[(X,Y,Z))* = raximum allwablo design peaking factor khich 1 cnsures that the Fn(X,Y,Z) limit will bo "

l proscrved for operation within the ICO limits,

' including al.1 wancca for calculational and reasurcmont unecrtaintics.

j. Notot (F/(X,Y,Z))* la the pararctor identified as B;0ES in DAW-10163P-A.

l l 2.5.5 (F[(X,Y,Z))*" is providcd in Table 2.

where (F[(X,Y,Z))"' = mu:imum allowablo design peaking factor which ensures that the centerlino fuel relt limit will be preservcd for operation within the LCD

. limits, including allwances for calculational l and reasurcrent tincertaintics.

l

. Notet (F[(X,Y,Z))" is the Inrameter identificd as DCDES in BAW- -

!. 101631'A.

l IO .

8-67 L.__....______.._.___ ._ _ _ _ _ . _ . - . _ . ._ _ -

PRL ENARY natw;'oa Unit 1 Core Cycratirg Limits Report (Continued) 2.5.6 11SI Dl E = 1. 0' 3

whero llSIDPC; = ikgative AFD limit adjustment Itquired to cogensate for each 1% that Fc(X,Y,Z) excoods its limit.

2.5.7 IslDIE = 1.0' 3

where ISIDIT = 3 Positive AfD limit adjustront requirtd to cmpensato for each 1% that rn (X,Y,3) exooeds its limit.

2.5.8 VSIDIE a .045' where }GIDPE = Adjustncnt to the K ivalue from Crrt>T required to coqcnnate for each 1% that ro(X,Y,Z) exceeds its limit.

' typical value; actual values will be supplied when nonitority inputs are camputed.

8-68

i i

,1 J .

K PRELIl\CNARY J

Catawba 1 Cycle 6 Core Operating Limits Report 1.2 4 . ,

4 1.0

i 4 . 4 s , , .

0.8 - -

b 0.6 -' - - - - +-

x . ,

i. - .

FA LEUlOSI E(Z) - - -

0.4 -- 00 ' 00 - - . - - - - - - -

8.0 1.00 ~" ' ' ' " ' ' ' ' '

10,8 0.94 i

'~

12.0 0.65 0.2 -~ - , .

..a.. , 4 4

' ' ' i I 0.0 - '

O 2 4 6 8 10 12 CORE HEIGHT (FT)

Figure 4 h \-

K(Z) - Normalized F q(X,Y,2) as a Function of Core Height for Mark-BW 8-69

1

(,~ PRELINARY Catawba 1 Cycle 6 Core Operating Limits Report 1.2 6 4 1.0 -

t 0.8 .

+ ~- -

i .

, @ 0.6 '- - - + . -

- r N :e i__. . . , . . ... r ,

y con.smI Km ... . . . . . . . .

C'* **

0'4 -

~ - ' - ~< +~ -

6.0 1.00 .._.2.. .' . L .y._ . . .

0.94 10,.8 3 ,o ey . .. -- . . . . . _ . .,

- . . - . . +

0.2 -*--- - + - - - - . e- +

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

4a . . ,

. .. . - ..r : . . . .

+. ;.

0 2 4 6 8 10 12 CORE HEIGHT (FT)

Figure 5 (N,- K(Z) - Normalized Fq(X,Y,Z) as a Function of i

l Core Height for OFA 1

S-70 t

I t

W , ..

Catakta Unit 1 Core Operatirg Limits Report (Continued)

G )Agkear L'nthalw P.ise Hot Channel Pactor (Specification 3/4.2.3) 2.6 Fj'i = F%P(X,Y,Z)/ AXIAL (X,Y) khere AXIAL (X,Y) is the axial peak frun the normlized axial pwer shape.

2.6.1 FRP(X,Y,Z) is prwidad in Table 3.

We follwim parameters are required for core mnitoring par the surveillance Requiremants of Specification 3/4.2.3 2.6.2 FMUE(X,Y) is prwided in Tabic 4.

Wocre FAHFf(X,Y) =mximum allevablo design radial peakirg factor which ensures that th9 Fa(X,Y) limit will be preservcd for operation within the I4:0 limits, including allcr.mnces for calculational and measuramant uncertaintics.

Note: FMUf(X,Y) is the paramotor identificd as HIDES in PM-10163P-A.

2.6.3 Fal#(X,Y) = F5(X,Y)/MA#/ AXIAL (X,Y) p khere Fj(X,Y) is the measured radial peak at location X,Y W# is the value of MAP (X,Y,Z) obtained fram Tabic 3 for the measured peak.

2.6.4 PRH = 3. 34' khen 0. 8 < P $ 1. 0 RRH = 1.67' khen P $ 0.8 khere RRH = 2erm1 Iwer reduction requircd to omrxttrate for each 1% that Fw(X,Y) cxceeds its limit.

p, hermal IWer Ratcd term 1 Iwer 2.6.5 TRH = 0.01*

khore TRH = Rcduction in CfroT K setpoint mquired to compensate for 3

i N

cach 1% that Fa(X,Y) exceeds its limit.

typical value; actual values will be supplicd when monitoring inputs are

. computed.

NctrE: Tablas 1, 2, ard 4 will be supplicd when mnitoring inputs are camputed.

)

i

\

v 8-71

=

1 I l i

- <4 .. .

l-a <- ..-.4<, J ... Y .,;,,, <

I

~

i.

l i

1:

i 7able 1. (F[(X,Y, Z) )*

!- (later) i i.

i n Table _2. (F[(X,Y, Z) )*

-(Iater)

I 1hi f

f 4

l I

e I

d 8-72 E _. ; _ . _z..- . . _ _ _ _ . . . _ ~ . _ _ _ . _ _ ~ - - . _ . . __ .. _.___._-._-.. _ .~..____.. . . . . _ _ . _ . . . _ _ _ _ _ - . , . , - . . _ _ _ _ . _ . _ _ . . _ , . . . _ _ _ . , - ~ _ _ . . _

). . w

. . A

. . . V J. A .

at

/ Tablo 3. Catata 1 Cycle 6 Cparatirg Limit thxiru:n A11cuablo 'Ibtai Peaks V

Peak __

FAP(X,Y,Z) 1%P(X, Y, Z)

Elc(Atipnit. Axial (OPM flurk-I,HL 2 1.1 1.747 1.818 4 1.743 1.814 6 1.737 1.808 8 1.725 1.796 10 1.701 1.771 2 1.2 1.948 2.028 4 1.939 2.018 6 1.924 2.003 8 1.902 1.900 10 1.048 1.923 2 1.3 2.158 2.246 4 2.141 2.228 6 2.115 2.201 8 2.072 2.157 i 10 1.962 2.042

,- s 2 1.4 2.333 2.428

\ 4 2.327 2.422 k'\' ) 6 2.294 2.388 8 2.185 2.274 10 2.058 2.142 2 1.5 2.498 2.600 4 2.496 2.598 -

6 2.399 2.497 8 2.278 2.371 10 2.149 2.237 2 1.7 2.824 2.939 .

4 2.710 2.820 6 2.574 2.679 8 2.443 2.543 ,

10 2.313 2.407 4 2 1.9 2.964 3.085 4 2.854 2.970 6 2.723 2.834 8 2.591 2.697 10 2.462 2.562

.O

\v) 8-73

n PRELIMNARY

. Table 4. FAllif(X,Y)

(later) u l

t i

s r

1-l 8-74

.. . , . . . - . . - . . . . , . . . . . . . . . , , - , . . . . . ~ - - . .

._ . . . . .- .m.._.__._. _. _ . . _ _ . . _ . _ _ . _ . - . . . . - . _ . . . . _ . . . . - . . _ _ _ . _ _ . . _ _ _ _ _ - . . . _ _ _ _ _ . _ . . . . . . _ _ . . . . . _ _ _ . .

(>

1 i-Cnarges to Final Safety Analysis Report i-l i

e 1:

'i 4

s-75 B&W FuelCompany

ATTACHMENT 1 CNS Feedwater System malfunction causing an increase in feeda ter flow O

t

'v) 15.1.2 15.1.3 Excessive increase in secondary steam flow 15.1.5 Steam system piping failure 15.4.2 Uncontrolled rod cluster control assembly bank withdrawal at power a 15.4.4 Startup of an inactive loop 15.5.1 Inadvertent operation of Emergency Core Cooling System during power operation 15.6.1 Inadvertent opening of a pressurizer safety or relief valve.

Loss of main feedwater flow is a Condition II occurrence by itself and is analyzed in Section 15.2.7. There is no credible reason for any of the Condition II events listed above to cause a loss of feedwater flow. There-fore, a loss of feeowater is not. considered coincidently with those occurances listed above which are Condition II.

For the steamline break c ansient it is conservative to assume main feeowater is available. This maximizes the amount of steam generator inventory avail-able to be blowdown and prolongs the transient.

,m The response times and discharge rates for some important plant valves 'and

/

( w) pumps are listed in Table 15.0.8-3.

" 15.0.9 FISSION PRODUCT INVENTORIES 15.0.9.1 Inventory in the Core The fission product radiation sources considered to be released from the fuel to the containment following a maximum credible accident are based on the assumptions stated in TID-14844 (Reference 1), namely 100 percent of the noble gases, 50 percent of the halogens and a core power level of 3565 MWt.

The time-dependent fission product inventories in the reactor core are calculate ty the OMCEN Ocde-(Reference 2) using a data library based on ENDF/B-IV Reference 3). .The core inventories are shown in Table 15.0.9-1.

4-, wkm.a,- pu by W Ppa-XO Lt=

The Equilibrium Appearance Rate of Iodines in the RCS due to conservative and realistic fuel defects ce shown in Table 15.0.9-2.

15.0.9.2 Inventory in the Fuel Pellet Clad Gao The radiation sources associated with accidents which may cause more than 1 percent failed fuel (loss of coolant accident, rod cluster control assembly ejection, and fuel handing accidents) are based on the assumption that the fission products in the gap between the fuel pellets and the cladding of the dameged fuel. rods are released as a result of cladding failure.

b g j The gap activities were determined using the model suggested in Regulatory v Guide 1.25. Specifically,10 percent of the iodine and noble gas activity 15.0-12 8-76

CNS O REFERENCES FOR SECTION 15.0 1, DiNunno, J. J., et al., " Calculation for Distance Factors for Pou r and Test Reactor Sites"," TID-14844, M&rch 1962,

_ cc-cn$-48,

P90-X 9 , S

2. -&,4L-4520, "0RICEM Yie%qrske, -end Crc;; \%%.- hct4en; - % citer Tren3mstetica-ind--.

" Geeey-Ott+--F+em-E*D F/0 - I V" , nedk t i o n S h k i dh -h t f o ma t i o n C e n t e r , ta k --

Ridne Ne t-iona l Laborat+cyApt " """' .-

t.u 3, RSIC-DLC-38, "0RIGEN Yields and Cross Sections - Nuclear Transmutation and Decay Data from ENOF/B-IV", Radiation Shielding Information Center, m

Oak Ridge National Laboratory, September 1975.

4. Chelemer, H. , Boman, L. J. , Sharp, D. R. , " Improved Thermal Design Procedures", VCAP-8567, July, 1975.

5, " Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors", 10CFR50.46 and Appendix K of 10CFR50, Federal Register, Volume 39, Number 3, January 4,1974.

6. Bordelon, F. M. , et al, " SATAN-VI Program: Comprehensive Space - Time Dependent Analysis of Loss of Coolant", WCAP-8301 (Proprietary) and WCAP-8306 (Non-Proprietary), June 1974. .

7. Bordelon, F. M. , et al, "LOCTA-IV Program: Loss of Coolant Transient Analysis", WCAP-8301 (Proprietary) and WCAP-8305 (Hon-Proprietary), June W

9 i

8, 1974.

Hargrove, H G., "FACTRAN - A Fortran IV Code for Thermal Transients In A 00 Fuel Rod", WCAP-7908, June, 1972.

9. Burnett, T. W. T. , McIntyre, C. J. , Buker, J C. , Rose, R. P. , "LOFTRAN Code Description", WCAP-7907, June, 1972,

10. Ri sher, D. H. , Jr. , Barry, R. F. , "TVINKLE - A Multi-Dimensional Neutron Kinetics Computer Code", VCAP-7979-P-A, WCAP-8028-A, January,1975,

11. " Westinghouse Nuclear Energy Systems Division Quality Assurance Plan",

WCAP-8370-A, August, 1984.

12, S. S. Kilborn, Westinghouse Letter of 1/27/88, DCP-88-508, l O

15.0-16 11/89 Update 8-77

Table 15.0.9-1 9 Iodina and Noble Gas Inventory in Reactor Core and Fuel Rod Gaps" Core Inventory Fraction of Inventory Inventor g Huclide (Curies) in Gao** (%) Gap (Curies) y I-131 1. 0EOS S.S E o1 ' .10 1. 0E07= S 5 80'

- I-132 1. EE08- 1 3 E oS .10 1.5E07= l.2Eo1

. I-133 0.1E00 \.sE06 .10 2.1E07= l SEo1 I-134 - 2. 3E00 2.1 G*O .10 - 2. 3E07 = 2.\ d 07 I-135 -3r0E08 l.SEc8 .10 2.0E07 1

B 607 Xe-131m 1. 4E0E= MEo' .10 1.4E09 1.aso5 Xe-133m 0.;E00= 5 6606 .10 6-SE05- 5,4 so5 Xe-133 -er1EOB- 1.iE oS .10 2.1E07 = t.S eo1 Xe-135m -3,0E07 dmo 7 .10 -3. 0EOS = 3.VE o' Xe-135 -2,250b t.9 sos .10 -2,2[07 1.se ol Xe-138 -1. 0E00 1. 4 E as .10 1.0E07= n. 6 E o'l Kr-83m 1. 35G7-- Ec .10 1. 3E00 1 \ E o' Kr-85m 2. 0E07-- % . .10 2. 0 E00 - 1 5 80' Kr-80 fE05- WE e5 .30 2.2E05= l.4E65 9 Kr-87 Kr-88 Kr-89 4r45G7- 't.~t Eo f

7. 4 E07 4.5 eoT

0. 4E07 9.se67

.10

-E.4E0S= % M o'

-h4E06- 4. 5 ac6

-0.4E05 B.0 Go6 4

Based on an equilibrum cycle core at end of life. The seven-region core operates at a power level of 3636 Wt and an average. cycle burnup of 10,500 MWD /MTU.

NRC assumption in Regulatory Guide 1.25 O

1 l

8-78

T ,. ewe .

v K ./

_ Table 15.0.'12-1 (Page 1)

Offsite Doses-(Rem):

FSAR' Exclusion Area Boundary Low Population Zone Accident Section Whole Body Thyroid Whole Body Thyroid 15.1.5 Main. Steam Line Break Case 1 (No lodine spike) '8.6E-2_ ' 7. 6 ' 4;4E-3 2.6E-1 t Case 2 (Pre-spike) 1.03-2 4.22 9.38-4 3.23-1 5.77-1 Case 3 (Coincident spike) 1.26 3.32 2.29-3 Loss of Power .z 15.2.6 Case 1 (No todine spike) 4.5E-3 7.0E-2 5.9E-4 6.5E-3 Case 2 (Pre-spike) 4.5E-3 7.3E-2 5.9E-4 7.6E-3 8.2E-3 Case 3 (Coincident spike) 4.5E-3 7.2E-2 5.9E-4

?[

d Ejection Accident Primary Side-Release Secondary Side Release 15.4.8 fE-2--].s E- 1 -f3- 5. 9 3.3E-2 2. 2. L2-- 1 8 E+l

+-1E-2 5 2 E - Z-ME--3 I 0 E- I 4 G 7 6 - 1 l 3.8E-2 6.o

~

Instrument Line Break 15.6.2 Case 1 (No iodine spike) 1.6E-1 3.2E-1 5.1E-3 1.0E-2 Case 2 (Pre-spike) 1.8E-1 1.9E+1 6.0E 6.3E-1 Case 3 (Coincident spike) 1.8E-1 5.2 6.0E-3 1.7E-1 i

Steam Generator Tube Rupture 15.6.3 ,

Case 1 (No ic- :ne spike) 6.4E-1 1.5 2.1E-2 8.8E-2 Case 2.(Pre-spike) 7.1E-1 4.4E+1 2.4E-2 1.5 Case 3 (Coincident spike) 7.0E-1 1.2E+1 2.3E-2 4.6E-1 Loss of Coolant Accident 15.6.5 Case 1 (With ECCS leakage) M 9.1 -1.5E+2 f.3 e 2. 9e. /. I m 3.zE+l Case 2 (Without ECCS leakage) %3- 9./ .L4E+2 1 2 E+2 9, 1f,/ 2 or+if.3E+1

{

Waste Gas Decay Tank 15.7.1 Rupture 5.0E-1 -

1.6E-2 -

~

, c.e e ou .

4. 9 E - , 3. z E-z.

, . 2_

% g (pu Spub} - - - -

37 g,

. _ . - . . . . . .. . ~ .- -. . . . . .

Table 15.3.3-2 (Page 1) t j Parameters for Postulated Locked Rotor Analysis 1.. Dati and and assumptions used to- conservative estimate radioactive source from postulated accident i

a. Power Level (HWt) 3555

b. Percent of fuel defected 1 c.. Total staam generator- 1 gpm tube leak rate during accident and initial 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months

.40 %

d. Activity released to reactor :-Wr of gap inventory a coolant from failed fuel i

e. Offsite Power Not'available

f. Reactor coolant activity Primary and -Secondary *

- prior to accident Activity During Normal Operations (Table 11;1.1-4)

2. Data and' assumptions used to- -

-estimate activity released A ,

. a. -Iodine partition factor 0.01

^1 (f .

b. Initial steam release from 515,247 lb (0-2 hr) 4 steam generators: 1.040,910 lb (2-8 hr)

- c .= . Duration of plant:cooldown '

8 by , secondary system 'after accident,-(hrs)

3. Dispersion-data !

~

a.. Distance to exclusion area boundary -(m) 762 :

b. Distance to low population zone-(m) -6096

[

c. 1/Q at exclusion ~ area boundary (sec/m3) -5.5E-04 d .~ x/Q :at low population:rone '(sec/m3) 1.8E-05 4.= Dose data- l a.- Method of' dose calhulations Regulatory Guide 1.4

'b.- Dose conversion assumptions Regulatory Guides 1.4 & 1.109

['

N'\ f

-- +

l L 8-80 I

4 - --

Table 15.3.3-2 (Page 2)

.(] Parameters for Postulated Locked Rotor Analysis Conservative

c. Doses.(Rem)

Case l'(No iodine spike)

Exclusion area boundary Whole body -4.50-03 4NE-ol Thyroid -7.OC-4 2- 3 6 Low population zone Whole body 5,^E o.9 g . g Thyroid -0. 03 /. '2-Case 2 (With pre-existing iodine spike)

Exclusion area boundary Whole body -4,5E-03 4 4 E - /

Thyroid -7.3E-02--5 7 Low population zone -

Whole body 5.0[-04-3 8 C.g Thyroid -7.0E-03 = /

k l

%' 3 (With coincident iodine spike) ~

Ex F area boundary Thyroid 7.2E-02 I

, 73 y Low populat pio Whqle40Ty 5.5E-04 k

M yroid 8,2E-03 L __

O 8-81

l

\

15.4.3.3 Environmental Consecuences OThemostlimitingrodclustercontrolassemblymisoperation,accidentalwith-drawal of a single RCCA, is predicted to result in less than JQuel clad 5%

damage. Thesubsequentreactorandturbinetripwouldresultinatmospneric stkam dump, assuming the condenser was not available for use. The radiolog-ical consequences from this event would be no greater than the e r -M ....

" event, analyzed 15,3,3 , n , g, t *Y I 1k'9 % 3 GMYE ^

~ TEA Une st t) t.ontoJn Section W 15.4.3.4 Conclusions t5 %+% san lysts Q w y[,l[#'7 $ ^/M [/f Mg 1 .

Y ,'l'$rgdtH*4 *!

Pom nvsl u t peurootspg e b,;), yW rer A h nm l era s,9 , o For cases of dropped RCCAs or dropped banks, for which the reactof 16 T,rippo_o,, /evit d by the power range negative neutron flux rate trip, there is no reduction in in fr. ped the margin to core thersal limits, and consequently the DNB design basis is o s o r., ,, ,,'

met. It is shown for all cases which do not result in reactor trip that the DNBR remains greater than the limit value and, therefore, the ONB design is met.

Fcr all cases of any RCCA fully inserted, or bank C inserted to its rod insertion limits with any single RCCA in that bank fully withdrawn (static misalignment), the DNBR remains greater than the limit value.

For the case of the accidental withdrawal of a single RCCA, with the reactor in the automatic or manual control mode and initially operating at full power with bank 0 et the insertion limit, an upper bound of the number of fuel rods experiencing DNB is 5 percent of the total fuel rods in the core.

STARTUP OF AN IMACTIVE REACTOR COOLANT PUMP AT AN INCORRECT O15.4.4 TEMPERATURE 15.4.4.1 Identification of Causes and Accident Description If tne plant is operating with one pump out of service, there is reverse flow through the inactive loop due to the pressure difference across the reactor vessel. The cold leg temperature in an inactive loop is identical to tha cold leg temperature of the active loops (the -reactor core inlet temperature). If the reactor is operated at power, and assuming the secondary side of the steam generator in the inactive loop is not isolated, there is a temperature drop across the steam generator in the inactive loop and, with the reverse flow, the hot leg temperature of the inactive loop is lower than the reactor core inlet temperature.

Administrative procedures require that the unit be brought to a load of less than 25 percent of full power prior to starting the pump in an inactive loop in order to bring the inactive loop hot leg temperature closer to the core inlet temperature. Starting of an idle reactor coolant pump without bringing the inactive loop hot leg temperature close to the core inlet temperature would result in the injection of cold water into the core, which would cause a reac-tivity insertion and subsequent power increase.

Should the startup of an inactive reactor coolan't pump accident occur, the transient will be terminated automatically by a reactor trip on los coolant a- '

15.4-15 8-82

CNS

2. Beginning of Cycle, Zero Power Control bank 0 was assumed to be fully inserted and banks B and C were at their insertion limits. The worst ejected rod is located in control bank D and has a worth of 0.784 ak/k and a hot channel f actor of 11.0.

The peak clad average temperature reached 2656'F, the fuel center temperature was 4140'F,

3. End of Cycle, Full Power Control bank D was assumed to be inserted to its insertion limit. The w0rst ejected rod worth and hot channel factors were conservatively calculated to be 0.25% ak/k and 5,90 respectiveiy, This resulted in a peak clad temperature of 2276'F. The peak hot spot fuel temperature reached melting, conservatively assumed at 48000F, However, melting was restricted to less than 10% of the pellet.

4. End of Cycle, Zero Power Control bank 0 was-assumed to be fully inserted and bank C was at its insertion limit. The ejected rod has a worth of 0.90% ok/k and a hot channel factor of 19,0 respectively, The peak clad and fuel center temperatures were 2586 and 3845 F respectively. The Doppler weighting f actor for this case is significantly higher than for the other cases, due to the very large transient hot channel factor.

A summary of the cases presented above is given in Table 15.4.8-1. The nuclear power and hot spot fuel and clad temperature transients for the worst cases are 9 presented in Figures 15,4,8-1 through 15.4,8-4, (Beginning of life fulI power and beginning of life zero power). I The calculated sequence of events for the worst case rod ejection accidents, as '

shown in Figures 15.4.6-1 through 15.4.8-4, is presented in Table 15.4.1-1.

For ali cases, reactor trip occurs very early in the transient, after which the nuclear power excursion is terminated. As discussed previously in Section 15,4,8,2,2, the reactor.will remain suberitical following reactor trip.

The ejection of an RCCA constitutes a break in the Reactor Coolant System, lo-cated in the reactor pressure vessel head. The effects and cor, sequences of loss of coolant accidents are discussed in Section 15.6,5. Following the RCCA ejection, the operator would follow the same emergency instructions as for any other loss of coolant accident to recover from the event, Fission Product Release o

It is assumed that fission products are released fr he gaps of all rods entering DNB, In all cases considered, less than ercent of the rods en-tered DNB.t n ed er : detH lad three d;in mional TilINC :n lysis (Deference 10).

Pressure Surge A detailed calculation of the pressure surge for an ejection worth of one dollar at beginning of life, hot full power, indicates that the peak pressure does not exceed that which would cause stress to exceed the faulted condition 15.4-33 1988 Update 8-83

. . . - ~ - . - - . - - . . _ - . .- - -. - . - - - . _.

CNS stress limits (Reference 10). Since the severity of the present analysis does not exceed the " worst case" analysis,-the accident for this plant will not re-sult in an excessive pressure rise or further damage to the Reactor Coolant System.

Lattice Deformations A large temperature gradient will exist in the region of the hot spot. Sin:e the fuel rods are free to move in the vertical direction, differential expan-sion between separate rods cannot produce distortion. However, the tempera-ture gradients across individual rods may produce a differential expansion tending to bow the midpoint of the rods toward the hotter side of the rod, Calculations have indicated that this bowing would result in a negative re-activity effect at the hot spot since Westinghouse cores are undermoderated, and bowing will tend to increase the undermoderation at the hot spot. Since the 17 x 17 fuel design is also undermoderated, the same effect would be observed. In practice, no significant bowing is anticipated since the struc-tural rigidity of the core is more than sufficient to withstand the forces produced.- Boiling in the hot spot region would produce a net flow away from that region. However, the heat from the fuel is released to the water rela-tively slowly and it is considered inconceivable that cross flow will be sufficient to produce significant lattice forces. Even if massive and rapid boiling sufficient to distort the lattice is hypothetically postulated, the large void fraction in the hot spot region would produce a reduction in the total core moderator to fuel ratio and a large reduction in this ratio at the hot spot. The net effect would therefore be a negative feedback. It can be concluded that no conceivable mechanism exists for a net positive feedback re-g)-

sulting from lattice deformation. In fact, a small negative feedback may re-V sult. The effect is conservatively ignored in the analysis.

15.4.8.3 Environmental Consequences A conservative analysis for a postW ated rod ejection accident is performed to determine the.resulting radiological consequences, The analysis is based on a instantaneous fission product release to the reactor coolant of the gap activity from 10 percent of the fuel rods in the core.-ple the nuvily h w-

-an ::umed 0. 5 p m e n core alt, Prior to the postulated rod ejection accident, it is assumed that the plant is operating at equilibrium levels of radioactivity in the primary and secondary systems with 1 percent fuel defects and a steam generator tube leak rate of 1 gpm. Following the accident, two activity release paths contribute to the total radiological consequences. The first release path is via containment leakage resulting from release of activity from the primary coolant to the L, containment. The second path-is the contribution of contaminated steam in -

l 'the secondary system dumped through the relief valves, since offsite power is assumed to be lost.

The following conservative assumptions are used in the analysis of the release of radioactivity to the environment in the event of a postulated rod ejection accident. A summary of parameters used in the analysis is given in Table 15.4.8-2.

F%

O 1. -Tar pkrcent of the gap activity is released.tc the cortehment etmosphw+.

15.4-34 L

8-83a

CNS

? 2. 50 perceni ef-the iedtnes-e + 100 percent-of-the neble geses-fn-the

'd' -meMed40:1 :re r-e4+ased. r.M M%gs i "M%^ % ,

3. 50 pe dr- rb>Tefet efVowP

-the-iedtne100 N f.e,r

.n ,e.

id re+ eased ca er-see wboy de[esit)-d e5 a~ruw t eh5 in the Y'rc4. *.rt sca IU wsey i(~* cebse to 8 s 'tM u.ar.

(. r 4*iet- '

Annuipaa. ser see reaut mW su.it e t m re

4. us activity which is ex austed prior to the time at which the annulus i reaches a negative pressure of -0.25 in.w.g. is unf.iltered.

i etemi. \

5. 4GGS 5f*64 10:k:d:ie,3oceues-et-twice-the-mexfeveMe-5 fr< 4 o.M Wec.,<tal, AoS ferW retional

,y) Jaleekage.,w

<.)t o r y . icet -e_

c_ 3 C. CCCS leakage-begins et the ear +fest pessitrie time sump recirculation-cen-tegin.

4. 7. Bypass leakage is 7 percent.

l.,8. The-effective annulus volume is 50 percent of the actual volume.

BE The annulus filters become fouled at 900 seconds resulting in a 15 percent reduction in flow.

't16. : Elemental iodine removal by the ice condenser begins at 600 seconds and continues for 3328.3 seconds with a removal efficiency of 30 percent.

l*.R. . One of the containment air return fans is assumed to fail.

)t. H.- The containment leak rate is 50 percent of the Technical Specifications

[vl 12. 1 5.

limit after 1 day, Fed ke-partition facto sfer ECCS leak ~e 10 0.1 for the cour:e of the-accid:nt. hh r%.v.) crave by Gat"*WA sfrays is Me<< f'e he,4d ad

. p<ilc.M.. i o dW e. .

14. Ne credit is teken for the euxiliary building filters for CCC4 leakage.

15. The redundant-hydrogen re:0=b4ncr; and-igniter fai' Therefore, pur-ges-

-ar: ' requ4 red fer hydeegen-centeA-- '

(The-following assumptions apply to the secondary side analysis).

m M. :All the; activity released is mixed instantaneously with the entire reactor coolant volume, N d K The primary to secondary leak rate is 1 gal / min.

)5.1&. The iodine partition factor is 0.1. o.o t .

% d97 Tt: :ter rele:sc termin:ta:

sr. seary syw n in s kes,120 seconds. m ol-rN *F f*d l **\ M Y

n. 20'. :All. noble gases which leak to the secondary side are released.

IsE The primary and secondary coolant concentrations are at-the maximum allowed by technical specifications.

q-1*(~

Based on the foregoing model, the primary and secondary side releases may be calculated as well as the offsite doses. The doses, given in Table 15.4.8-2,

. 15.4-35 8-84

I Table 15.4.8-2 (Page 1)

Parameters for Postulated Rod Ejection Accident Analysis Co nervative Realistic

1. Data and assumptions used to estimate radioactive source from postulated ac-cidents

a. Power level (MWt) 3565. 3565.

b. Percent of fuel defected 1. . 0.12

c. Steam generator tube leak rate prior 1. 0.008 to and during steam dump (gpm)

d. Failed fuel

percent of same fuel rods in core

e. Activity released to reactor coolant from failed fuel and available for release 60 Noble gases 40 percent of same core gap inventory

's) Iodines percent of same core gap inventory 00

f. Melted fuel -O # r)ercent of core 0.

g. Activity released to reactor coolant from melted fuel and available for release to containment o.0 Noble gases erFr percent of 0.

core inventory 00 lodines 0.15 percent'of 0.

core inventory

h. Iodine Fractions (organic, Regulatory same elemental, and particulate) Guide 1.4

2. Data and assumptions used to estimato activity released

a. Containment Free volume (ft') 1.015E+06 same

b. Containment leak rate 0.3 percent of 0.05 percent I containment vol- of containment ume per day, volume per day, f's'>) 05ts24 hr 05ts24 hr 11/89 Update 8-85 .

Table-15.4.8-2_(Page2)

., I

? Parameters for Postulated Rod Ejection Accident Analysis Conservative Realistic 0.15 percent of 0.025 percent I containment vol- of containment ume per day, volume per day, i t>24 hr t>24 hr

c. Bypass leakage fraction '0.07 0.07 Iodine partition-factor for steam o.cl d.

. -release

e. Offsite power Lost --

plant emendown h .hrsd g,

f. 4t:= tmp 'see relgteconWg(b)

. e f--val ve s s_ 44500; -

-+ -- Deratien of dee+ irum relisi velves- 4&- -

i 46*C+

3. - ' Dispersion data

a. Distance to exclusion area boundary (m) 762. 762.

T b. -Distance to low. population zone (m) 6096.' 6096.

J (J

c. .x/Q at' exclusion area boundary (sec/m') 4 0-2 hrs 5.5E-04 1.3E-04

-d. ' x/Q at-low population zone (suc/m .) 8 0-8 hrs 1.8E-05 6.2E-06 8-24 hrs 1.2E-05 5.4E 1-4 days 4.3E-06 2.5E-06 4+ days 1.2E-06 9.7E-07 4; -Dose data

a. Method of dose calculation Regulatory same Guide 1.77 -

b. Dose conversion ~ assumptions Regulatory Guides same 1.4 and 1.109

c. Doses (Rem)

~ Primary side 3h*.,

Exclusive area boundary O Whole body 7. 2 C --2 c2 . TE - I Thyroid --5+ 59 11/89 Update 8-85a

1 i

9, ?

(]- Table 15.4.8-2 (Page 3) 'j Parameters for Postulated Rod Ejection Accident Analysis Conservative- Realistic

-Low population zone !

Whole body -1.1C-02. 8' ZE- L ,

Thyroid- M, g ,7 E - l l- j Secondary side ._

I Exclusion area boundary - '

Whole body - 3. 3 E =ce 2. 2. i Thyroid M j,yE+l

. Low population zone .

.Whole body. -hfe-03 l ' 0 0 ~ I Thyroid -- 3.SE-02 6 O i

i s.

i-5 i

(

i 8-86 11/89 Update

. . . - - - .~

CNS-Offsite Dose Consecuences gs

(' The:offsite radiological consequences of a LOCA are calculated based on the t .following assumptions and parameters:

1. 100 percent of the core noble gases and 25 percent.of the core iodines are released to the containment atmosphere.

2. 50 percent chu ne. is; of- the core iodines are deposited in the sump *. ti*h> d O'Og ,W'.

sp4u f raet.w b o.'ti ne=\; o os f r s.

t. L Annulus activity which is exhausted prior to the time at which the annulus reaches a negative pressure of -0.25 in, w.g. is unfiltered.

5. /. ECCS leakage begins at the earliest possible time sump recirculation can begin.

L 5. fr.CS EC leaka.Ee occu.rs at twic.ec.the maximum operatiolal p leakage. A,h.

us idM t 'is3.M*'f w

m

. Sc. N t.a So y m . fac W. .m. ,

,f t=um ww swt.m k w.

.,s .%.

w se

% toc 1e a, csqw,w ;e/

3. ww ,

7,f Bypass leakage is 7 percent.

0. 7, The effective annulus volume-is 50 percent of the actual volume.

'1. Jf. The annulus filters become fouled at 900 seconds resulting in a 15 percent reduction in flow.

-lo g. Elemental iodine removal by the ice condenser begins at 600 seconds and

. continues for-0520.0 seconds with a removal efficiency of 30 percent. g^

, S7) Htn t (jl.16.1 One of the containment air return fans is assumed to fail.

'lC M.-- The containment leak rate is fifty percent of the Technical Specific tion limit after 1 day. .i Ib.)f. Iodine partition factor for ECCS leakage-is 0.1 for the course of the accident.

K M. - No credit is -.taken for! the auxiliary building filters for ECCS leakage.

15.J4. Th: redund:nt hyrdeg:n ree biner: :nd igniter: f:!', Ther fere, purge:

.are requ! red fer hydregen _ eentre! %,.h reaw ) e.ea,h t by co4 4 ;g 97, _

.-;i %6 f,r. e.l 4,ctu ., 4 - tW ed x ee h .

The doses are presented in Table 15.6.5-9 and are within the 1-1;;;it: ef p:.t.t w )w ,d 10 CFR 100,-

i

/3 a (u/

15.6-18 1988 Update 8-87

CNS

! Control Room Operator Oose 7

kj The maximum postulated dose to a control room operator is determined based on TfihrDNses of a Design Basis Accident. In addition to the parameters and assumptions listed above, the following apply:

. 2,600

1. The control room pressurization rate is 4,000'efm; the filtered recircula-

- tion rate is 2,000 cfm. 1, Aeo J- t,aras %e c.atra ( area (c.%Me J l etesrecsA es.ame x..n roo m s ),

2. The unfiltered inleakage into the control room is 10 cfm.

3. Other assumptions are listed in Table 15.6.5-10.

.15.6.6 A NUMBER OF BWR TRANSIENTS Hot applicable to Catawba.

t k.a 15.6-19 11/89 Update 8-88

~ -.. .

2 1

Tabl'e~15.6.5-9 (Page 1)

/ Parameters =for LOCA Offsite Oose Analysis .c

. Conservative Realistic

1. D'ata-and'assumpiions used to estimate radioactive source.from postulated

-accidents ,

a.' -Power level 3565. 3565.

I bc Fail fuel- 100% of fuel 2% of fuel L rods in core- rods in core

- Activity released- to reactor coolant.

-c.

from failed-fuel-and available for release ,

' Noble gases- '100% of core 2% of core '

activi_ty: activity

lodines 50% of core 2% of core 4

activityi activity i TIodinel fraction's-(organic, elemental, Regulatory Guide same' n - d.

.and particulate) 1.4

),

(/ Data and assumptions used:to estimate 2.- -

activi_ty. released

. a.

Containment free volume a 3

dpper containment -volume (ft3)' 6.70+05 same-JLower containment, volume-(ft3); .

3.45E+06 -same

-Totaljcontainmentfreevolume(ft3)_

1.015E+06- same

.b., LIodine: activity" released to 25% 'same-containment"

, c .1 / Containment leak rate;- 0.3% of.contain- :0.05% of con- bl ment .vol ume 'per-- - tainment volume: .

Lper . day, :01ts24 hr' 1 day, 05t$ 24-hr l-

" 0.15% of contain- _0.025% of con-~

~

' ment volume per tainment volume 4

day, ts24.hr per day, t>24 hr- t n 0,07 0.07

.- d.; -Bypass leakage: fraction'

'e.- Annulus ventilation. iodine filter 99%

7;O efficiency 95% ,

l:' ' ( . )

%d 1988 Update 89

- ~ _. 2. -- . _ . . .-

i Table 15.6.5-9 (Page 3) _

Parameters for LOCA Offsite Dose Analysis Conservative Realistic-

c. Doses (Rem)--

Case 1-(with ECCS leakage)

Exclusion Area Boundary Whole Body -Er4- %)

. Thyroid- 1.;E;; g.3E4c.

Low Population Zone Whole Body -9.45 1.t !

Thyroid- 3. iE* 1: 3.2.E9 i -i

. Case 2 (Without ECCS leakage)

- Exclusion Area Boundary -

Whole Body -56 - 't.) i Thyroid. 1.4C:2 i . 2 et 2.

Low Population Zone Whole Body 0.?: 1 i.s . .

Thyroid 2. 0:: 1- i.s si i

~!

. 1

)

Y l

, i L

8-90 11/89 Update '

.:-._.__.._=-. _ . _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ . . , _ . ._ . . . _ -, _,. . - , , _ , , , . ,

_ . _ . . - ._ .. _ _ . _ _ . . . _ _ _ - _ _ _ _ . . . - . _ _ _ . . _ _ . _ . _ . _ - . _ _ . ~ . _ _ _ . . . ____

i j ..-

e-

-Table 15,6.5-10 (Page 2)

- :7-

\ Parameters for LOCA Control RooE Qtse Analysis QLnservative Realisti_g j

< 3. Dispersion data - Control room-intake eu. x/Q (sec/m')

0-8 hrs 9.9E-04 8-24 hrs 7.2E-04 '

1-4 days 5.1E-04 4+ days 2.8E-04

4. Dose data.

'a. Method of dose calculations Standard Review '

Plan 6.4

.b. . Dose conversion assumptions Regulatory Guides 1.4, 1.109

c. Doses (Rem)

Whole body 4Mr4 16 ,

Thyrotd- 2. 00 : ;- n.%st t

' Skin 1.7 ' 3.a s t i q '

r L

i 1

8-91

p

9. STMUUP HUSICS TEETfDC The staniuti scope of ruload startup physics testirg corducted at the Catawba Unit 1 is summrized below:

Zero Power Mwsics 'Ibstim (ZPPT)

All Ibds Out Critical Doron Coneontration (.@OCDC)

- Isothernal Tenperature Coefficient (IT)

- Control Rod Bank Worth Diffcrential Boron Worth (DBi)

Power Escalation 7tstim (PET)

- Flux Symetry Check (Iru P:ucr, 59, 30 %FP) 8 -

Core Ruer Distribution - CPD (Intermediate Power)

CFD (FullL Power)

AROCDC (Full ther)

BiFC has tuviewed the startup thysics testirg program (scope, test methods, and acceptance criteria) for Catawba. All aspects of the existirg program are acceptable with respect to inplementation or the BfFC licensirg analyses and a carpleto reload batch of Mark-Bi fuel assemblies. Therefore, operation with either a mi):ed Westirghouse and BiFC core or future cores with all BfEt fuel will not require any charges to the current Duke startup physics testirg program.

'~1 B&W Filel Company

f% '

lv)

-10.0 i<uudziCES

-1. Catawba Nuclear Station, Final Safety Aralysis Report, Eccket Nos. 50--

- 413 ard 50-414.

2. . BAW-10173P, Mark-N Reload Safety Analysis for Catawba ard MMre, PaWR & Wilcox, Lynchbug, Viqinia, Man::h 1989.

3. PAW-10172P, Mark-N Mechanical Design Report, MWM & W11oax, Lynchburg,. Virginia, July 1988.

4.' BAW-10084A, Rev. 2, Pnagram to Detennine In-Reactor of B&W Fuels-Cladding Creep Collapse, naWk & Wilcox, Lynchburg, Virginia, October 1978.

5.. BAW-10141P-A. Rev.1, TACO 2-Fuel Pin Perfonnarco Aralysis, Babcock.&

Wilcox, Lynchburg, Virginia, June 1983.

6. BAW-10162P-A, TACO 3,-Ibel Pin 'Jhermal Analysis Code, mWk & Wilcox, Lynchburg, Virginia,- November 1989. _,

/D;

) 7.- BAW-101!li26, NCODE - A Multi-Dimensional 'No-Group Reactor Simulator,

}V-mW4 & Wilcox,1Lynchburg, Virginia, June 1985.

S.- BAW-10117P-A, Babcock & Wilcox Version of PDQ User's Manual, mWk & - )

Wilcox, Lynchburg, Virginia l January 1977.

9. BAW-10124A, FIAME3 - A 'Ihree Dimensional Nodal Code for. Calculatiry Core Reactivity ard Power Distributions, naW2 & Wilcox, Lynchburg, Virginia, August 1976.

-10. BAW-10170P-A, Statistical Core Design for Mixing Vane Cores, MWk & .;

Wilcox, Lynchburg, Virginia, Wer 1988. -

11. BAW-10159P-A, BWQ4V Correlation of Critical Heat Flux in Mixing Vane Grid Fuel Assemblies, Babcock & Wilcox, Lynchburg, Virginia, July 1990.

1 1

12. BAW-10173P, Rev. -1,_ Mark-N Reload Safety Analysis for Catawba ard ItGuire, mWk ard Wilcox, Lynchburg, Virginia, October 1990i

- 13 . _= BAW-10173P, Rev. _2, Mark-N Reload Safety Analysis for Catawba ard j; McGuire, mWR & Wilcox, November 1990.

14. . BAW-10174, Mark- N Reload LOCA Analysis for the~ Catawba and M:Guire Units, Babcock & Wilcox, September.1989.

.A

~/

lj 10-1

\,,

B&W FuelCompany u

15. IVM-10168P, Pav.1, B&W Ims-of-Coolant Accident Evaluation Model For Racirculatirg Steam Generator Plants, Babcock & Wilcox, Lynchburg, Virginia, September 1989.

16. BAW-10174, Dev.1, Mark-IM Reload LOCA Analysis for the Catakta and itGuire Units, Baboa:A & Wilcox, November 1990.

17. EWW-10163P-A, Core Operatirg Limit Methodolcgy for Westirghousc-Dosigned IMPS, Bhk & Wilcox, Lynchburg, Virginia, June 1989.

\

10-2

)

1 B&W FuelCompany

ML20066E070

Text

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