ML20154A671

Tech Spec,Reflecting Changes Necessary to Support Cycle 12 Operation

ML20154A671
Person / Time
Site:	Fort Calhoun
Issue date:	09/02/1988
From:	OMAHA PUBLIC POWER DISTRICT
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ML20154A631	List: ML20154A628 ML20154A658 ML20154A671
References
NUDOCS 8809130048
Download: ML20154A671 (82)

Text

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ATTACHMENT A j

Technical Specifications Pace Numbera i

12 15 16 j

2 56 2-57 i

2 57a i

2 57c i

L Ficure Numbers 13 7

26 i

29 l

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l 8809130040 0S0902 DR ADOCK 0500 jD

1.0 SAFETY LIMITS AND' L!MIT!?iG SAFETY SYSTEli SETTINGS

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1.1 Safety Limits - Reactor Core (Continuec) would cause DNS at a particular core location to the actual heat flux at that location, is indicative of the margin Io Ot:8.

The minimum value of the DNBR during steady state opera-tien, normal operational transients, and anticipated tran-sients is limited to 1.18.

A DNBR of 1.18 corresponds to a l

95: probability at a 95% confidence level that CNB will not occur, which is considered an appropriate margin to ONS for all operating conditions.(1)

The curves of Figure 1-1 represent the loci of points of re-actor thermal power (either neutron flux instruments or AT in-struments), reactor coolant system pressure, and cold leg te ;erature for which the DNER is 1.18.

The area of safe opera-l tien is belcw these lines.

The reactor core safety limits are based on radial peaks limit-ed by the CEA insertion limits in Section 2-10 and axial shapes within the axial power distribution trip limits in l

Figure 1-2 and a total unrodded planar radial peak of 1.85.

tc The LESS in Figure 1-3 is based on the assumption that the un-redded integrated total radial peak (F ) is 1.80.

This peak-I ing factor is slightly higher (more ce servative) than the raximum predicted unrodded total radial peak during core life,

{

excluding measurement uncertainty.

Flow maldistributien effects for operation under less than full reacter coolant flow hav been evaluated via model tests.( )

The ficw model data established the maldistribution factors and het channel inlet temperature for the thermal analyses that were esed to establish the safe operating enve-lopes presenter. M Figure 1-1.

The reactor protective system is designed to prevent any anticipated ccmbination of tran-sient ecnditiens for reactor coolant system temperature, pres-sure, and ther al powe;' level that would result in a CNER of less than 1.18.(3) k Re f e rence s_

(:1) USAR, Section 3.6.7

()

USAR, Section 1.4.6 (3)

USAR, Section 3.6.2 I

A endment No. 1,32.A3.A7.

1-2 70.77,92

e 1.0 SAFETY LIMITS A?ID LIMITIdd SAFE"t SYSTD4 SEMI?iSS

((

1.2 Safety Limit. Remeter Coolant Fysten Pressure (Continued)

Peterences

A4, (1)

TSAR, Secticn b

t.c (2)

FOARi Secticn L.3 3

,u (3)

F3AR~,'Secticn L.3.k

,, m.

(k)

TGAR, Secticn L.3 9.5 u9%

(5)

TSAR, Secticn 7.k.5.1 s-t 1-5

b 1.0 SATETY LIMITS A!O LIMITING SAFETY SYSTD4 SEI' TINGS 13 Limiting Safety System Settings. Peneter Protective System Atmlic abili ty This specificatien applies to RPS Limiting Safety System settings and bypasses for instrument channels.

ObP etive To provide fer aute atic protection action in the event that the principal process variables approach a safety limit.

Specifientien The reactor prctective syste= trip setting limits and the remissible typasses fer the instru=ent channels shall be within the Li:dting Cafety Cyste Cetting as stated in Tsble 1-1.

Pssis The reacter prctective system consists of four instru=ent channels to 30nitor selected plant cenditions which vili cause a reactor trip if any of these conditions deviate frc= a preselected crerating range to the degree that a ta.ety limit =.ay be reached.

(1)

Hinh r var tevel - A reacter trip at high power level (neutren flux) is provided to prevent da= age to the fuel cladding result-ing frcn sc:e reactivity excursions too rapid to be detected by pre:sure and te=perature r. ear.urements (in addition. themal signals are provided to the high power level trip unit as a backup to the neutrcn flux signal).

During ner.al plant operation, with all reactor coolant pu.ps crerating, reactor trip is initiated when the reactor power l

level reaches 107.0% of indicated full power.

Adding to this l

i the pessible variatien in trip point dug to calibration and I

tessure ent errers, the taximum actual shhstate power at l

vhich a trip vould te actua}ej is 112%, which was used for the l

purpose of safety analysis.\\ll Provisions have teen r.ade to l

select different high-pover level trip points for varicus centinatiens of ret.ctor ecole.nt pu=p op telev under "Lov Feteter Ceolant Flov".(er)ation as described 2

Daring reacter cperatien at power levels tetvcen 19.1% and 100%

of rsted ;cver. the Variable High Fever Trip (VHPT) vill initiate a rea: tor trip in the event of a reactivity excursion that increases reactor pcVer by 10% or less of rated pover.

D.e high pcver trip set point can be set to t7re than 10% of I

f rated pcVer abcVe the indicated plant pcVer.

Operator acticn

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is required to increase the set point as plant power is increased.

The cet reint is aute:atically decreased as power decreases.

A enrent No. g, 32 14

2.0 LIMITI!!C COtIDITIONS FOR OPERATIOtt j

2.10 Reactor core (continued) 2.10.4 Power Distribution Limits Applicability Applios to power operation conditions.

i Objective To ensure that peck linear heat ra tee, D110 ma rgins, and radial peaking factors are maintained within acceptable limits during power operation.

Specification (1)

Linear lleat Rato The linear heat rate shall not exceed the limits shown on Figure 2-5 when the following factors are appropriato1y included 1.

Flux peaking augmentation factors are shown in Figure 2-8,

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2.

A measuremont-calculational uncertainty factor of 1.062, II?c lintAr.lat A4 n A hti(.h<.

3.

An engineering uncertainty factor of h1cni t:r e.l 123 Mt i.uC4rf.

1.03, d d Ltter Sqtte.n M 1",CCrdantC 4 '.

A linear heat rate uncertainty factor of l

b$th.yldfiCaNYJ 2.10 4 (l\\(Q t 1.002 due to axial fuel densification l

Of3.l0.4(%)I,b),C7NGsfAMtIy\\

and thermal expansion, and M d Skg L % % wif W S

^f*"M."***"**"*""**'"'Y**'

i H,e I;W h c4 R n 2 m

a)

.When the linear heat rate is continuously If&CgMR. W g1 sj hD fa4 monitored by the incore detectors, and the 2

7,,1(),y (thc) linear heat rate is exceeding its limits as

- indicated by four or more vclid coincident incoro detector alarms, either 4

(i)

Rostore the linear heat rate to with-1 in its limits within one hour, or i

(a) ao in at least hot standby within the n9xt 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

i 1

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Amendment tio. 7, ;c, ; A, 32, 4;, 47, 77 i

2-56

L..t.....

t,..,......

r-

,F........

u

....o

........c

.o

.u. n. 4 2.10 Eese::r cra (Continued)

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2.10.k Fever Distributien Linits (Centinued) s (b)

!! vhile c;erstir. under the previsi:n3 of fitt ( 1,'.

the ;11nt cen; uter in: Ore 'ietector s.arr.s tee:te in ;erable. 0;erati:n =sy to centinueh vithcut ca.clu-p. war,... Me $7seh erthe~ foll:ving

nditi:ns is satisfied:

7v,c 't o c.t q 3 p?m

& Af t

1. h1L M 'd'1 '

(i; A ccre ;;ver distributica vss ettair.ed y,,,; g s,.h,tre,an'J

t utili:ing in cre detectors within 7 day:

ri

r to the incere detectcr alarn cutage and the nessured peak linear hent rate vss no greater than 9C% of the vslue al-IcVed by (1) above.

(ii)

The Axisi Shape Index as =casured by ex-

re detectcrs rensins vithin 1.C5 of the value :ttained st the time of the isst essured in: re pcVer distributien.

(iii)

?:ver is net incressed ncr has it been in:resse; since the time of the last in-c re power distributicn.

When the lir. cst heat rate is c:ntinu:usly :: nit: red ty the ex::re detecters, vithdrsv the full length k-0"A's ter:nd the 1:nq ters insertien ituits cf 2;ecificati:n 2.10.2."*f If the linear heat rate m-is exceeding its 11:113 as deter =ined by the Axisi

_J.Jed'WOkdn b l

Iha;c Index, Y., teir.g cutside the limits cf Tifare 2-6, where 1C0'per:ent cf the til:vs'cle ;cver re-m s

n A ha U'ij {i.l '.Y e f t) {

i presents the ::xtnus ;cver alloved by the foll:ving b/fkn th",le'MN3 d i

ex;ressien:

1

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__is ure 2-4 a

)

15.22 x x I

F

.s vhere 1.

L is the maximum alicvable lir.ett heat rste as determined fr:n Figure 2-5 and is based en the c re sverste turnup at the ;L=e cf the latest ine:re pcVer :sp.

2.

M 10 the 2ximu 211:vstle frs: tion of rates ther:21 7:ver as deterninci tr -he T..- lin;

urve Of Ti?are 2-9 vnen ::nli:rir.; ty ex::re dete:::r:.

' = 1 vnen ::n;;: ring kv/ft Os;n; ir.-

l

re detect:rs.

F.est:re tr.c res:::r 7:ver :::..a;al 5h:;e

.....c..,..,....u........... ___..s

.,e..n.s

-6 vithin 2 h:urs. Or i

=

1 4

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2.0 LIMITING CONDITIONS FOR OPERATION _

2.10 Reactor Core (Continueo) 2.10.4 Power Distribution Limits (Continued)

(ii) 3e in at least hot stancby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

(2)

To'.31 Inteersted Radial Peskinn Fsetor The calculated value of Fk defined by Tk = FR (I*T ) shall g

be limited to < 1.80.

FR is determined from a power distribu.

tien map with no non-trippable CEA's inserted and with all I

full length CEA's at or above the Long Term Steady State Insertion Limit for the existing Reactor Coolant Pump The asirutnal tilt. T, is the ressured value of c -tinatien.

a Tq at the tire FR is detemined.

With F[ > 1.20 within 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />s:

(a) Redu:epcwertobringpowerandF3withinthelimits of Figure 2-9, withdraw the full length CEA's to or beyond the Long Tem Steady State Insertion Limits of Specification 2.10.2(7), and fully withdraw the NTCEA's, l

or (d) Be in at least hot standby.

(3) T9t31 Planar Radial Peskina Factor The calculated value of FxyT defined as Fxyi = Fxy (1+Tq) shall te limited to 1 1.85. Fxy shall be determTned ;' rom a pcwer distribution map with no non-trippable CEA's ime.'ted and with l

all full length CEA's at or above the Long Tem Steady State Insertion Limit for the existing Reactor Coolant Pump ccmdina.

tion.

This detemination shall be limited to core planes between 15t and 85% of full core heignt inclusive and shall exclude regions influenced by grid effects.

The sairuthal tilt.

T, is the ressured value of Tq st the time Fxy is detemined, g

'.30 With F,yT > h35 within 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />s:

(a)

Reduce power to bring pcwer and FxyT to within the limits I

of Figure 2 9, withdraw the full length CEA's to or beyond the f.ong Term Steady State Insertion Limits of Specification 2.10.2(7), and fully withdraw the NTCEA's, or l

l (b) Se in at least not standby.

1 A.' enc ent No. 32. A3,47,70,77,32,109 2-Ha

2,0 LIMITING CONO:1!0'45 FOR OPERATICN

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2,10 Reactor Core (Continued) 2.10.4 Power Distribution Limits (Continued)

(5) ONBR Margin During Pcwer Oceratien Above 15% of Rated Power (a) The following DNS related parameters shall be maintained within tne limits shewn:

(i)

Cold Leg Temperature 1 545*F*

(ii)

Pressurt:er Pressure 1 2075 psia *

(iii)

Reactor Ccolant Ficw 1197,000 gre" (iv)

Axial Shape Index, Y; 1 Figure 2-7"'

(b) With any of the above parameters exceeding the limit, restore the parameter to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce scwer to less than 15'; of rated power within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Basis Linear Heat R3te The limitation on linear heat rate ensures that in the event of a LOCA, the peak te m iture of the fuel cladding will not exceed 2200*F.

Either of ;he two core power distribution menitoring systems, the Excore Detec-(

ter M:nitoring System, or the Incore Detector Monitoring System, provide

(

adequate monitoring of the core power distribution and are capable of verifying that the linear neat rate does not exceed its limits, The Excere Oetector Monitoring System.:erfvres this function by coatinuously monitoring the axial shace index cith the operable quadrant symetric excore neutron flux detectors and verifyi g that the axial shape index is maintained within the allowable limits of Figure 2-6 as adjusted by Specification 2,l'),4(1)(c) for the allowed T of Figure 2-9.

linearheatrateofFigure2-5,RCPumpconfiguration,andF,Xdinestablishing In conjunction with the use of the excere monitoring system a the axial sha:e index limits, the following assumptions are made:

(1) the CEA insertion limits of Specification 2,10,2(6) and long term insertion limits of Specification 2.10.2.(7) are satisfied. (2) the flux peaking augmentatien factors are as shcan in Figure 2-8, and (3) the total planar radial peaking factor does not exceed the limits of Specification 2.10.4(3).

• Limit not applicable during either a thermal power ramp in excess of 5', of rated thereal power per minute or a thtrmal power step of greater than 10%

of rated thermal power, "This numoer is an actual limit and corresponds to an indicated flew rate of 202.500 gpm, All other values in this listing are indicated values and include an allowance for measurement uncertainty (e.g., S4{,F,, indic:sted, allcos for in actual T-of 547'F).

- e

'"Tre At! AL SKAPE lN EX,' Core pcwer shall te maintained within tne limits estaolisnea by the Cetter Axial Shape Selection System (BASS.5) for CEA insertions of tne leaa tank of < 65: when BASSS is opersole, er within the limits of Figure 2-7.

A.enr ent No. 32,A3,37,70,77,pg, 109 2-57c

590 l

I 580 6

\\-

N 570 N

\\

\\

\\

e'~60 N

N N

^

550 U

2250 psio N

=

'40 N

~

t:

N' N.2075esio o $30 u

U

~

O 520 u

' 1750 psio

$10

- + - -

500

-=

60 70 80 90 100 110 120 CORF, POWER (". OF RATED POWER)

P,g = 29.73PF(B)A1(Y)B + 18A4T

- 11350 g

PF(8)

= 1.0 81100F..

=

.0088 + 1.S 50T.< B < 100".

= 1.4 Bs50" A1(Y) = -0.35294Y; + 1.08S24 Y, s.25 0 57143Yi + 0.875 Y >.25

=

i Thermal Yorgin/ Low Pressure LSSS Omaha Public Power District figure 4 Pump Opercticn Fort Calhoun Station-Unit No.1 1-3

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i 100 i

90 E

(-0.06, 80)

(0.10, 80) i y

80 x

i g

/-

N i

(-0. 2, 75.0)

(0. 2,75.0) i S

70 E

a d

60 es

._e.

50 m

W 40 2

E 30 8

20 10 0

=-

-0.3

-0.2

-0.1 0.0 0.1 0.2 0,3 AXIAL SHAPE INDEX Y I LimitingConditionforOperationfor OmahaPublicP0werDistrict Tigure ExcoreMonitoringOfLS FortCalhounStation-UnitNo.1 2-6

i10 i

7 T

F LIMIT = Fxy LIMIT n

5 90 S

Ea d

80 b

l U.96. 75) s 5

l g

70 E

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l 60 i

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o 1.75 1.80 1.85 1.90 1.95 2.00 2.05 Fr ANDFj y

FlyFAandCorePower OmahaPubiicPowerDistrict Figure Limitations FortCalhounStation-UnitNo.1 2-9

FORT CA1J10UN STATION l

l UNIT No. 1 i

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CYC1.E 12 RELOAD EVAL.UATION i

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Fort Calhoun Cycle 12 License Application CONTDITS 1.

INTRODUCTION MiD SU!t%RY 2.

OPERATING HISTORY OF THE RETERENCE CYCLE 3.

CUiERAL DESCRIPTION l

4.

WEL SYSTEM DESIGN 5.

NUCLEAR DESIGN 6.

THERMAL.HYDRAbtIC DESIGN J

1 7.

TRANSIENT ANALYSIS 8.

ECCS PERFORNANCE MiALYSTS S.

STARWP TESTING 1

I 10.

RETERDiCES j

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1.0 INTRODt'CTION AND St NARY i

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j This report provides an evaluation of the design and performance for I

tne operation of Fort Calhoun Station Unit No. 1 during its twelfth d

fuel cycle at full rated power of 1500 WJt. All planned operating con.

ditions remain the same as those for Cycle 11.

I The core will consist of 89 presently operating J, K, L and M assem.

i blies and 44 fresh Batch N assemblies, f

The Cycle 12 analysis is based on a Cycle 11 ternination point between a

/

13,100 L'D/T and 14,100 KJD/T.

In performing analyses of design basis

}

)

events, determining limiting safety settings and establishing Ilmiting l

l!

conditions for operation, limiting values of key parameters were chosen l

to assure that expected Cycis 12 conditions vould be enveloped, provid.

l j

ed the Cycle 11 termination point falls within the above burnup range.

l In accordance with Reference 1, the fuel burnup limitations on Batch K

)

fuel further restrict the Cycle 11 upper bound to 13,860 KJD/MTV.

The i

j analysis presented herein will accommodate a Cycle 12 length of up to j

13,450 KJD/T, 1

The evaluation of the reload core characteristics have been conducted with respect to the Fort Calhoun Unit No. 1 Cycle 11 safety analysis described in the 1987 update of the USAR, hereafter referred to as the "reference cycle" in this report unicas noted otherwise.

l E

Specific core differences have been accounted for in the present anal.

ysis.

In all cases, it has been concluded that either the reference I

cycle analyses envelope the new conditions or the revised analyses pre.

[

sented herein continue to show acceptable results. Where dictated by I

j variations from the previous cycle, proposed modifications to the plant t

Technical Specifications have been provided.

(

)

The Cycle 12 core has been designed to reduce fluence to critical reac-tot pressure vessel welds to minimize the RTPTS shift of these velds, j

This will preclude the reactor vessel velds reaching the Pressurized Thermal Shock RTPTS screening criteria of the current 10 CI7,50.61 regulations and maximize the time to reaching the screening criteria if j

the Ree. Guida 1.99, Rev. 02, methods are used to revise 10 CTR 50.61.

l The u tysis presented in this report was performed utilizing the meth-i odology documented in the District's reload analysis nethodolo;f re.

l ports (References 1, 2, and 3).

These methodologies were previously transmitted in References 4, 5 and 6.

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2.0 OPERATING HISTORY OF THE PREVI0tf5 CYCLE i

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Fort Calhoun Station is presently operating in its eleventh fuel cycle

}

l utilizing Batch H,1 J, K, L and M fuel assembites.

Fort Calhoun

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Cycle 11 operation began on June 8,1987, and reached full power on l

f June 30, 1987. The reactor has operated up to the present time with l

j the core reactivity, power distributions and peaking factors having

[

closely followed the calculated predictions, 1

]

It is estimated that Cycle 11 vill be terminated on or about September 23, 1988.

The Cycle 11 termination point can vary berveen 13,100 MVD/T

]

and 14,100 MV0/T and still be within the assunptions the Cycle 12 1

analyses.

In accordance with Reference 1 the fuel bu. ap limitations on Batch K fuel further restrict the Cyete 11 upper bou,. to 13,860 j

MVD/MTV. As of July 24, 1988, the Cycle 11 burnup had reached 12,334 j

MVL/T.

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3.0 CENERAL DESCRIPTION The Cycle 12 core will consist of the r.aaber and type of assemblies and fuel batches shown in Table 1 1.

One H assembly, one I assembly, 6 J assemblies, 21 K assemblies and 15 L assemblies will be discharged this outage.

They will 'i replaced by 20 fresh unshimmed Patch N assemblies (3.T0 w/o enrichment) and 24 fresh shimmed Batch N assemblies (3.70 w/o. 0.020 gm B10/ inch).

Figure 3 1 shows the fuel management pattern to be employed in Cycle 12.

The primary chango to the core in Cycle 12 is the reduction of the initial enrichment by 0.1 w/o of U 233.

The locations of the poison pins within the lattice of shimmed assemblies and the fuel rod loca-tions in unshimmed assemblies are shown in Figure 3 2.

Figcre 3-3 shows the beginning of Cycle 12 assembly burnup distribution for a Cycle 11 terninstion burnup of 13,600 MWD /T. The fuel average discharge exposure at the end of Cycle 11 is projected to be 38,211 MWD /T.

The initial enrichment of the fuel assemblies is also shown in Figure 3 3.

Figure 3 4 shows the end of Cycle 12 assembly burnup dis-tribution.

The end of Cycle 12 core average exposure is approximately 29,224 MWD /T.

1

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Table 3-1 Fort Calhoun Cycle 12 Core Tewlirq Initial BoC IDC Poison Poison Asse:tly Number of Average Burrup (PMD/T)

Average Bunlup (PMD/T)

Rods per Tsw11rg Designation Assemblies

[BOC 11 = 13,600 PWD/T]

[BOC 12 = 13,450 PMD/T' A h ly ga Dio/ inch J*II) 8 34,858 39,819 0

0 K

8 39,188 43,823 0

0 L

21 25,196 39,129 0

0 I/

8 32,300 45,812 8

.01904 M

20 14,817 26,801 0

0 M/

24 17,792 33,361 8

.024 N

20 0

14,087 0

0 N/

24 0

17,502 8

.020

'IUTAL 133 (1)Assosablies Delivered for Cycle 8, But First Lewini Into Cycle 9

FIGURE 3-1 FORT CALHOUN STATION CYCLE 12 CORE LOADING PATTERN AA ASSEMBLY LOCATION 01 02 BB FUEL TYPE M

K

~

i 03 04 05 l06 07 Ja N

N N/

M/

08 09 10 11 12 13 J*

N N/

L M/

N/

14 15 16 17 18 19 N

N/

L/

M M/

L/

20 21 22 23 24 25 N

L M

M/

L N/

26 M

27 28 29 30 31 32 N/

M/

M/

L M

L 33 K

34 35 36 37 38 39 M/

N/

L/

N/

L L

FIGURE 3-2 FORT CALHOUN STATION CYCLE 12 ASSEMBLY FUEL AND POISON ROD LOCATIONS UNSHIMMED ASSEMBLY i

1 i

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r 1

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l 1

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I l

1 l

I i

ll I

t l

t

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1 I

l l1 I

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ll l

l_

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I 1

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L/, M/, N/ - 8 POISON RODS PER ASSEMBLY l

I I

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f i

X N

L l

l t

l l

l l

1 X

X 1

1 i

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i X

X l

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X; XI i

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l FUEL ROD LOCATION l

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POISON ROD LOCATION f

2 i

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FIGURE 3-3 FORT CALHOUN STATION CYCLE 12BOCkSSEMBLYAVERAGEEXPOSURE

/

AND INITIAL ENRICHMENT AA ASSEMBLY LOCATION 01 02 BB FUEL TYPE M

K C.CC ENRICHMENT (W/0 U-235) 3.80 3.50 D0,D00 ASSY AVG EXP (MWO/T) 15,523 39,644 03 04 05 06 07 J*

N N

N/

M/

3.50 3.70 3.70 3.70 3.80 34,843 0

0 0

19,293 08 09 10 11 12 13 J*

N N/

L M/

N/

l 3.50 3.70 3.70 3.80 3.80 3.70 34,874 0

0 28,353 18,505 0

l 14 15 16 17 18 19 N

N/

L/

M M/

L/

3.70 3.70 3.80 3.80 3.80 3.80 0

0 31,476 13,403 16,553 33,028 20 21 22 23 24 25 N

L M

M/

L N/

3.70 3.80 3.80 1.80 3.80 3.70 26 0

27,926 13,419 17,257 20,392 0

3.80 27 28 29 30 31 32 15,549 N/

M/

M/

L M

L 3 'a 3.80 3.80 3.80 3.80 3.80 33 0

18,532 16,569 20,741 16,189 26,067 K

3.50 34 35 36 37 38 39 38,732 M/

N/

L/

N/

L L

3.80 3.70 3.80 3.70 3.80 3.80 19,383 0

33,218 0

28,876 29.587 NOTE: E0C 11 CORE AVERAGE BURNUP = 13,600 MWD /T

o FIGURE 3-4 FORT CALHOUN STATION CYCLE 12 E0C ASSEMBLY AVERAGE EXPOSURE AA

- ASSEMBLY LOCATION 01 02 88

- FUEL TYPE M

K CC,CCC ASSY AVG EXP (MWD /T) 21,710 44,234 03 04 05 06 07 l

J*

N N

N/

4/

39,797 12,569 15,024 15,825

2,412 08 09 10 11 12 13 Ja N

N/

L M/

N/

39,841 15,213 18,019 41,879 34,344 18,792 14 15 16 17 18 19 N

N/

L/

M M/

L/

12,581 18,015 45,119 29,645 32,739 46,429 20 21 22 23 24 25 N

L M

M/

L N/

26 15,047 41,533 29,641 33,527 35,561 18,592 M

27 28 29 30 31 32 21,763 N/

'A/

M/

L M

L 33 15,863 34,376 32,710 35,774 31,245 39,232 K

34 35 36 37 38 39 43,411 M/

N/

L' N/

L L

32,526 18,790 46,5&.)

18,401 41,517 41,215

4.0 FUEL SYSTEMS DESIGN The mechanical design for the Batch N fuel is essentially the same as the Batch M fuel supplied by Combustion Engineering, Inc. in Cycle 11.

The Batch N fuel is similar in design to the Batch G fuel supplied by Combustion Engineering in Cycle 5 and is mechanically, thermally, and hydraulically compatible with the Advanced Nuclear Fuel (ANF) supplied fuel remaining in the core.

Reference 2 describes the Batch M fuel characteristics and design.

This report was previously transmitted in Reference 3.

References 4 and 5 remain valid for describing the design of the ANF-supplied fuel.

Thirty-six (36) total fuel pins from Batch K (20) and Batch L (16) are projected to exceed the burnup limits estab-lished in Cycle 11 for extended burnup of ANF fuel.

ANF has reanalyzed the subject fuel pins in accordance with References 5 and 6 and established that operation of the fuel pins in Cycle 12 will not violate any of the extended burnup criteria.

The maximum burn-up for these pins was increased from 49,000 MWD /MTU to 50,000 MWD /MTU for Batch K and from 51,600 MVD/MTU to 52,000 MWD /MTU for Batch L fuel.

O 5.0 NUCLEAR DESIGN 5.1 PHYSICAL CHARACTERISTICS 5.1.1 Fuel Manacement The Cycle 12 fuel management uses a low radial leakage design, with onec, twice, and thrice burned assemblies predominately loaded on the periphery of the core. This low radial leakage fuel pattern is utilized to minimize the flux to the pressure vessel welds and achieve the maximum in neutron economy. Use of this type of fuel management to achieve reduced pressure vessel flux over a standard cut-in-in pattern results in higher radial peaking factors.

The peaking factors for Cycle 12 are consistent with previous cycles in which low radial leakage patterns have been utilized.

As described in Section 3.0, the Cycle 12 loading pat-tern incorporated 44 fresh Batch N assemblies (24 shimmed N/ and 20 unshimmed N) with an enrichment of 3.70 w/o.

Eight thrice burned Batch J* ast+mblies, which were delivered for Cycle 8, but inicia11y loaded into the core for Cycle 9, are being returned to the core to be combined with 8 thrice burned K asseablies, 29 twice burned L assemblies, and 44 once burned M assemblies to produce a Cycle 12 pattern with a cycle energy of 13,450 500 MWD /T.

The Cycle 12 core char-acteristics have been examined for a Cycle 11 termina-tion between 13,100 MWD /T and 14,100 MWD /T and limiting values established for the safety analysis. The Cycle 12 loading pattern is valid for any Cycle 11 endpoint between these values.

Physics characteristics including reactivity coeffi-cients for Cycle 12 are listed in Table 5-1 along with the corresponding values from Cycle 11.

It should be noted that the values of paramaters actually employed in safety analyser are different from those displayed in Table 5 1 and are typically chosen to conservatively bound predicted values with accommodation for appropri.

ate uncertainties and allowances.

Table 5 2 presents a summary of CEA shutdown worths and reactivity allowances for the beginning of Cycle 12 Hot Zero Power Steam Line Break accident.

The BOC liZP SLB is the most limiting accidant of those used in the deter-mination of the required shutdown margin.

The Cycle 12 values, calculated for minimum scram worth, exceed the minimum value required Technical Specifications and thus provide an adequate shutdown margin.

5.1.2 Power Distribution Figures 5 1 through 5 3 illustrate the all rods out (ARO) planar radial power distributions at BOC12 MOC12

5.0 tTUCLEAR DESIGN (Continued) 5.1 PHYSICAL CHARACTERISTICS (Continued) 5.1.2 Power Distribution (Continued) and ECC12, respectively, and are characteristic of the high burnup end of the Cycle 11 shutdown window. These planar ra-dial power peaks are representative of the major portion of the active core length between abouc 20 and 80 percent of the fuel height.

The high burnup er,d of the Cycle 11 shutdown window tends to increase the power peakt.ng in this axial central region of the core for Cycle 12.

The planer radial power distributions for the above region, with Bank 4 fully inserted at beginning and end of Cycle 12 are shown in Figures 5 4 and 5 5, respectively.

The radial power distributions described in this section are calculated data without uncer:ainties or other allowances.

However, the single rod power peaking values do include the increased peaking that is characteristic of fuel rods adjoin-ing the water holes in the fuel assembly lattice.

For both DNB and kw/ft safety and setpoint analyses in.ither rodded or unrodded configurations, the power peaking values actually used are higher than those expected to occur at any time dur-ing Cycle 12.

These conservative values, which are used in Section 7 of this document, establish the allowable limits for power peaking to be observed during operation.

Figures 3 3 and 3-4 show the integrated assembly burnup values at 0 and 13,450 MVD/T, based on an EOC11 burnup of 13,600 MWD /T.

The range of allowable axial peaking is defined by the limit-ing conditions for operation covering the axial shape index (ASI). Within these ASI limits, the necessary DNBR and kw/ft margins are maintained for a wide range of possible axial shapes.

The maximum three dimensional or total peaking fac-tor anticipated in Cycle 12 during normal base load, all rods out operation at full power is 1.97, not including uncertain-ty allowancos.

5.1.3 Safety Related Data 5.1.3.1 Ejected CEA Data The maximum reactivity worth and planar power peaking factors associated with an ejected CEA event are shown in Table 5 3 for both beginning and end of Cycle 12.

These values encompass the worst conditionr. anticipated during Cycle 12 for any expected Cycle 11 termination point.

The values shown for Cycle 12 are calculated in accordance with Reference 7.

In addition, Table 5 3 lists those values used in the Reference Analysis (Cycle 11) for :omparison.

l l

5.0 fTUCLEAR DESIGN (Continued) 5.1 PHYSICAL CHARACTERISTICS (Continued) 5.1.3 Safety Related Data (Continued) 5.1.3.2 Dropped CEA Data The Cycle 12 safety related data for the dropped CEA analysis were calculated iden-tically with the methods used in Cycle 11.

5.2 ANALYTICAL INPUT TO INCORE MEASUREMENTS Incore detector measurement constants to be used in evaluating the reload cycle power distributions will br calculated in the same manner as for Cycle 11.

5.3 NUCLEAR DESIGN METHODOLOGY Analyses have been performed in the manner t.nd with the method-ologies documented in Referernes 8 and 9.

5.4 UNCERTAINTIES IN MEASURED POWER DISTRIBUTIONS The power distribution measurement ur. certainties which are ap-plied to Cycle 12 are the same as those presented in Reference 9.

I I

9 TABLE 5-1 FORT CALHOUN CYCLE 12 NOMINAL PHYSICS CHARACTERISTICS h

Cvele 11 Cvele 12 Critical Boron Concentration Hot Full Power, ARO, Equilibrium Xenon, BOC PPM 1081 1081 Inverse Boron Worth Hot Full Power, BOC PPM / top 113 113 Hot Full Power, EOC PPM / top 90 90 Reactivity Coefficients (CEAs Withdrawn)

Moderator Temperature coefficients Beginning of Cycle. HZP 10ap/*F

+0.23

+0.25 End of Cycle, HFP 10*'op/'F

-2,47

-2.49 Doooler Coefficient Hot Zero Power, BOC 10 53pj.F

-1.96

-1.97 Hot Full Power, BOC 10 53pf.F

-1.42 1.47 10 5,j.F 1.54

-1.57 Hot Full Power, EOC 3

Total Delaved Neutron Fraction. Reff BOC 0.00609 0.00607 EOC 0.00522 0.00521 Neutron Generation Time.

f*

BOC 10 6 sec 22.3 22.2 EOC 10 6 sec 28.0 28.0

TABLE 5-2 FORT CAIJIOUN UNIT 1 CYCLE 12 LIMITING VALUES OF REACTIVITY WORTHS AND ALLOWANCES FOR ll0T ZERO POWER STEAM LINE BREAK, top Cvele 11 (EOC)

Cvele 12 (EOC) 1.

Worth of all CEA's Inserted 10.07 8.70 2.

Stuck CEA Allowance 2.80 1,42 3.

Worth of all CEA's Less Worch of Most Reactive CEA Stuck Out 7.27 7.28 4.

Power Dependent Insertion Limit CEA Worth 1.35 1,41 5.

Calculated Scram Worth 5.92 5.87 6.

Physics Uncertainty plus Bias 0.59 0.59 7.

Net Available Scram Worth 5.33 5.28 8.

Technical Specification Shutdown Margin 4.00 4.00 9.

Margin in Excess of Technical Specification Shutdown Margin 1.33 1.28

TABLE 5-3 FORT CAIROUN UNIT 1 CYCLE 12 CEA EJECTION DATA BOC11 Value EOC11 Value BOC12 Value EOC12 Value Maximum Radial Power Peakinc Factor Full Power with Bank 4 inserted; worst CEA ejected 3.74 3.21 2.38 2.15 Zero power with Banks 4+3 inserted; worst CEA ejected 5.74 5.27 4.85 4.82 Maximum Ejected CEA Worth (SAo)

Full power with Bank 4 inserted; worst CEA ejected 0.39 0.38 0.39 0.29 Zero Power with Banks 4+3 inserted; worst CEA ejected 0.65 0.66 0.56 0.62

a e

FIGURE 5-1, FORT CALHOUN STATION CYCLE 12 ASSEMBLY RELATIVE POWER DENSITY 0 MWD /T, HOT FULL POWER, EQ. XENON AA ASSEMBLY LOCATION 01 02 B.B888

- ASSEMBLY RPO'S

.4057

.2778 C.CCC MAXIMUM 1-PIN PEAK ASSY 03 04 05 06 07

.3389

.9425 1.1267 1.0903

.9019 08 09 10 11 12 13

.3402 1.1497 1.3332 1.0026 1.1812 1.3379 1.670 14 15 16 17 18 19

.9447 1.3338 1.0213 1.2845 1.2570

.9847 20 21 22 23 24 25 1.1305 1.0111 1.2831 1.3089 1.1807 1.4149 26

.4083 27 28 29 30 31 32 1.0949 1.1831 1.2532 1.1682 1.1973 1.0094 33

.2839 34 35 36 37 38 39

.9052 1.3392

.9830 1.3966

.9638

.8824 MAXIMUM 1-PIN PEAK AT 30% CORE liEIGHT

FIGURE 5-2, FORT CALHOUN STATION CYCLE 12 ASSEMBLY RELATIVE POWER DENSITY 7000 MWO/T, HOT FULL POWER, EQ. XENON AA ASSEM8LY LOCATION 01 02 B.BB88 ASSEMBLY RPD'S

.4691

.3423 C.CCC

- MAXIMUM 1-PIN PEAK ASSY 03 04 05 06 07

.3655

.9401 1.1230 1.2012

.9999 08 09 10 11 12 13

.3664 1.1383 1.3587

.9932 1.1933 1.4252 1.658 14 15 16 17 18 19

.9407 1.3582

.9990 1.2033 1.2079

.9826 20 21 22 23 24 25 1.1242

.9987 1.2017 1.2034 1.1000 1.3875 26

.4710 27 28 29 30 31 32 1.2037 1.1932 1.2043 1.0903 1.1098

.9543 33

.3488 34 35 36 37 38 39 1.0014 1.4245

.9794 1.3733

.9169

.8422 MAXIMUM 1-PIN PEAK AT 30% CORE HEIGHT

l FIGURE 5-3, FORT CALHOUN STATION CYCLE 12 ASSEMBLY RELATIVE POWER DENSITY 13,450 MWD /T, HOT FULL POWER, EQ. XENON

~

AA ASSEMBLY LOCATION 01 02 B.BB88 ASSEMBLY RPO'S

.5289

.4021 C.CCC MAXIMUM 1-PIN PEAK ASSY 03 04 05 06 07

.3975

.9507 1.1275 1.2681 1.0504 03 09 10 11 12 13

.3982 1.1394 1.3610

.9867 1.1765 1.4293 1.691 14 15 16 17 18 19

.9510 1.3605

.9868 1.1487 1.1567

.9694 20 21 22 23 24 25 1.1283

.9913 1.1478 1.1340 1.0521 1.3520 26

.5307 27 28 29 30 31 32 1.2702 1.1768 1.1550 1.0452 1.0699

.9369 33

.4091 34 35 36 37 38 39 1.0520 1.4294

.9676 1.3439

.9054

.8414 MAXIMUM 1-PIN PEAK AT 22% CORE HEIGHT

FIGURE 5-4, FORT CALHOUN STATION CYCLE 12 ASSY RPO'S WITH BANK 4 INSERTED 0 MWO/T, HOT FULL POWER, EQ. XENON AA ASSEM8LY LOCATION 01 02 8.B888 ASSEMBLY RPO'S

.4391

.3104 C.CCC

- MAXIMUM 1-PIN PEAK ASSY XXXXXX CEA BANK 4 LOCATION 03 04 05 06 07

.2086

.7909 1.1327 1.1949 1.0154 08 09 10 11 12 13

.2096

.4724 1.0708 1.0101 1.3006 1.5092 XXXXXX 14 15 16 17 18 19

.7934 1.0718

.9434 1.3474 1.3917 1.1074 20 21 22 23 24 25 1.1373 1.0191 1.3462 1.4'25 1.2998 1.5573 1.797 26

.4422 27 28 29 30 31 32 1.2005 1.3030 1.3877 1.2863 1.2637 1.0030 33

.3174 34 35 36 37 38 39 1.0196 1.5112 1.1059 1.5378

.9575

.5292 f

XXXXXX MAXIMON 1-PIN PEAK AT 50% CORE HEIGHT

FIGURE 5-5, FORT CALHOUN STATION CYCLE 12 ASSY RPD'S WITH BANK 4 INSERTED 13,600 MWD /T, HOT FULL POWER, EQ. XENON AA ASSEMBLY LOCATION 01 02 B.BB88 ASSEMBLY RPO'S

.5873

.4595 C.CCC MAXIMUM 1-PIN PEAK ASSY XXXXXX CEA BANK 4 LOCATION 03 04 05 06 07

.2401

.7905 1.1493 1.4172 1.2036 08 09 10 11 12 13

.2406

.4500 1.0763

.9925 1.3088 1.6326 1.937 XXXXXX 14 15 16 17 18 19

.7911 1.0762

.8924 1.1940 1.2828 1.0948 20 21 22 23 24 25 1.1510

.9975 1.1935 1.2241 1.1525 1.4820 26

.5898 27 28 29 30 31 32 1.4206 1.3098 1.2814 1.1452 1.1134

.9099 33

.4679 34 35 36 37 38 39 1.2062 1.6336 1.0933 1.4741

.8792

.4721 XXXXXX MAXIMUM 1-PIN PEAK AT 22% CORE HEIGHT

i 4

6.0 THERMAL-HYDRAULIC DESIGN 6.1 DNBR Analysis Steady state DNBR analyses of Cycle 12 at the rated power of 1500 MWt have been performed using the TORC computer code described in Reference 1, the CE-1 critical heat flux correlation described in Reference 2, and the CETOP D computer code described in Reference 3.

This combination was used in the Cycle 8 through 11 Fort Cal.

houn reload analyses (References 4 through 7) and the reload meth-odology can be found in Reference 8.

Table 6-1 contains a list of pertinent thermal-hydraulic parame-ters used in both safety analyses and for generating reactor pro-tective system setpoint information.

The calculational factors (engineering heat flux factor, engineering factor on hot channel heat input, rod pitch and clad diameter factor) listed in Table 6 1 have been combined statistically with other uncertainty fac-tors at the 95/95 confidence / probability level (Reference 9) to define the design limit on CE 1 minimum DNBR.

6.2 FUEL ROD B0VING The fuel rod b1w penalty accounts for the adverse impact on MDNBR of random variations in spacing between fuel rods. The penalty at 45,000 KVD/MTU burnup is 0.5% in HDNBR.

This penalty was applied to the MDNBR design limit of 1.18 (References 6 and 10) in the statistical combination of uncertainties (Reference 9),

i

I 4 i

TABLE 6 1 f

Fort Calhoun Unit 1 Thermal Hydraulic Parameters at Ful1~ Power Enig cvele 12*

Total Heat Output (Core Only)

MWt 1500 6

10 BTU /hr 5119 Fraction of Heat Generated in Fuel Rod

.975 Primary System Pressure Nominal psia 2100 Minimum In Steady State psia 2075 i

Maximum In Steady State psia 2150 Inlet Temperature

'F 545 Total Reactor Coolant Flow gpm 202,500 6

(Steady State) 10 lba/hr 76.49 (Through the Core) 10 lbm/hr 73.08 Hydraulic Diameter-(Nominal Channel) ft

.044 6

2 Average Mass Velocity 10 lbm/hr.ft 2.24 l

l Core Average Heat Flux I

2 (Accounts for Heat Generated BTU /hr.ft 181,189 i

in Fuel Rod) l 1

2 Total Heat Transfer Surface Area ft 28,255**

[

t Average Core Enthalpy Rise BTU /lba 70.5 Average Linear Heat Rate kv/ft 6.1**

Engineering Heat Flux Factor 1.03***

i i

Engineering Factor on Hot Channel i

Heat Input 1.03***

l i

Rod Pitch and Bow 1.065***

Fuel Densification Factor (Axial) 1.01***

I

• Design inlet temperature and nominal primary system pressure

[

vere used to calculate these parameters.

l

• • Based on Cycle 12 specific value of 448 shims.

f

• • • These factors were combined statistically (Reference 8) with other uncertainty factors at 95/95 confidence / probability i

level to define a design limit on CE.1 minimum DNBR.

[

i I

f

- - ~ -. -

7.0 IE/WSIENT ANALYSIS This section presents the results of the Omaha Public Power District Fort Calhoun Station Unit 1, Cycle 12 Non-LOCA safety analysis at 1500 MWt.

The Design Bases Events (DBEs) considered in the safety analysis are listed in Table 7 1.

These events were categorizeo in the following groups:

1.

Anticipated Operational Occurrences (A00s) for which the inter-vention of the Reactor Protection System (RPS) is necessary to prevent exceeding acceptable limits.

2.

A00s for which the intervention of the RPS trips and/or initial steady state thermal margin, maintained by Limiting Conditions for Operation (LCO), are necessary to prevent exceeding accept-able limits.

3.

Postulated Accidents The Design Basis Events (DBEs) considered in the Cycle 12 safety anal-yses are listed in Table 7 1.

Core parameters input to the safety anal-yses for eva*-uating approaches to DNB and centerline temperature to melt fuel design limits are presented in Table 7 2.

As indicated in Table 7 1, no reanalysis was pe ormed for the DBEs for which key transient input parameters are within the bounds (i.e., conse-rvative with respect to) of the reference cycle values (Fort Calhoun Updated Safety Analysis Report including Cycle 11 analyses, Reference 1).

For these DBEs the results and conclusions quoted in the reference cycle analysis remain valid for Cycle 12.

For those analyses indicated as reviewed, calculations were performed in accordance with Reference 6 until a 10 CFR 50.59 determination could be made that Cycle 12 results would be bounded by Cycle 11.

All events were evaluated for up to a total of 6% steam generator tube plugging in Cycle 11.

Fort Calhoun Station currently has 1.084 steam generator tubes plugged, thus; no additional analysis is required.

For the events reanalyzed. Table 7 3 shows the reason for the remnal-ysis, the acceptance criterion to be used in judging the results and a sanaary of the results obtained.

Detailed presentations of the results of the reanalyses are provided in Sections 7.1 through 7.3.

4 TABLE 7-1 FORT CAU10UN UNIT 1, CYCLE 12 DESIGN BASIS EVENTS CONSIDERED IN THE NON-LOCA SAFETY ANALYSIT Analysis Statug 7.1 Anticipated Operational Occurrences for which inter /ention of the RPS is necessary to prevent exceeding acceptable limits:

5 7.1.1 Boron Dilution Reviewed 7.1.2 Excess Load Reviewed 7.1.3 Reactor Coolant System Depressurization Reviewed 7.1.4 Loss of Load Not Reanalyzed 7.1.5 Loss of Feedwater Flow Not Reanalyzed 7.1.6 Excess Heat Removal due to Feedwater Halfunction Not Reanalyzed 7.1.7 Startup of an Inactive Reactor Coolant l

Pump Not Reanalyzed 7.2 Anticipated Operational Occurrences for which RPS trips and/or sufficient initial steady state thermal margin, maintained by the LCOs, are necessary to prevent exceeding the acceptable limits:

2 7.2.1 Sequential CEA Croup Withdrawal Reanalyzed 3

7.2.2 Loss of Coolant Flow Reanalyzed,5 7.2.3 CEA Drop Reviewed 7.2.4 Transients Resulting from the 0

Malfunction of One Steam Generator Not Reanalyzed 7.3 Postulated Accidents 5

7.3.1 CEA Ejection Reviewed 5

7.3.2 Steam Line Break Reviewed 7.3.3 Steam Generator Tube Rupture NotReanalgzgd 7.3.4 Seized Rotor Reviewed '

NOTE:

All events evaluated or reanalyzed for the effect of increased steam generator tube plugging to 64/SC.

1Technical Specifications preclude this event during operation.

2Requires liigh Power and Variable High Power Trip.

3Requires Low Flow Trip.

0Requires trip on high differential steam generator pressure.

4-Event bounded by reference cycle analysis. A negative 10 CFR 50.59 deter-mination was made for this event.

TABLE 7-2 FORT CA!JIOUN UNIT 1, CYCLE 12 CORE PARAMETERS INPUT TO SAFETY ANALYSES FOR DNB AND CTM (CENTERLINE TO MELT) DESIGN LIMITS 4

l Physics Parameters

]lD111 Cvele 11 Values Cvele 12 Values i

Radial Peaking Factors l

For DNB Margin Analyses y

(F T) l r

R L

J Unrodded Region 1.80*

1.80*

l Bank 4 Inserted 1.98*

1.90*

I

)

For Planar Radial Component I

(F T) of 3 D Peak xy (CTM Limit Analyses) l I

Unrodded Region 1.85*

1.85*

Bank 4 Inserted 2.04*

1.94*

I i

Maximum Augmentation f

Factor 1.000 1.000 i

3 l

l J

Moderator Temperature l

Coefficient 10*0Ap/* F 2.7 to +0.5 2.7 to +0.5

[

l I

Shutdown Margin (Value l

}

Assumed in Limiting j

EOC Zero Power SLB) 4.0 4.0 i

r I

{

l

• For the Loss of Coolant Flow and CEA Diop Events, the effects of uncertain.

ties on these parameters were accounted for statistically in the DNBR and CTH j

i calculations.

The DNBR analysis utilized the methods discussed in Section

(

6.1 of this report. The procedures used in the Statistical Cembination of j

Uncerr.ainties (SCU) as they pertain to DNB and CTM limits are detailed in i

j References 2 5,

}

I i

)

4 1

(

l I

i I

r i

I i

t I

i l

I I

I

TABLE 7-2 (Continued)

Safety Parameters EDill Cvele 11 Values Cycle 12 Values Power Level MVt 1530*

1530*

Maximum Steady State Temperature

'F 547*

547*

Hinimum Steady State Pressurizer Pressure psia 2053*

2053*

Reactor Coolant Flow gpm 202,500*

202,500*

Negative Axial Shape LCO Extreme Assumed at Full Power (Ex Cores)

Ip 0.18 0.18 Maximum CEA Insertion

% Insertion at Full Power of Bank 4 25 25 Maximum Initial Linear Heat Rate for Transient Other than LOCA KV/ft 15.22 15.22 Steady State Linear Heat Rate for Fuel CTM Assumed in the safety Analysis KV/ft 22.0 22.0 CEA Drap Time to 1004 Including Holding coil Delay sec 3.1 3.1 Minimum DNBR (CE 1) 1.18*

1.18*

• For the Loss of Coolant Flow and CEA Drop Events, the effects of uncertainties on these parameters were accounted for statistically in the DNBR and CTM calcula-tions. The DNBR analysis utilized the methods discussed in Section 6.1 of this report.

The procedures used in the Statistical Combination of Uncertainties (SCU) as they pertain to DNB and CTM limits are detailed in References 2 5.

TABIE 7-3 IISICN BASIS EVDrr ICANALYZED FOR FORP CAUCCN CYCIE 12 Baa m for Acceptance S amary Event Ibanalysis Criterion of Results SognM CEA Grup Withdrawal Change in rod worth Minin

' "i greater IG5t = 6.99 (HZP) nanoonservative with t:w 1 -,.2 talac ~E-1 POER = 1.28 (HFP) lower reactivity in-tw adcr. Onnsient PDER < 22 kW/ft.

sortion rate.

w/ft.

Im : of (bolant Flow Change in rod worth a '

-w. -v.! c Mini== DER = 1.43 nanoonservative with 1, :.

L5-F lower reactivity in-L.

A sertion rate.

l 1

7.0 TRANSIENT ANALYSl1 l

i 7.1 ANTICIPATED OPEL.TIONAL OCCURRENCES 7.1.1 Boron Dilution Event The Boron Dilution event was reviewed for Cycle 12 to verify that sufficient time is available for an operator to identify the cause and to terminate an approach to criticality for all suberitical modes of operation.

Table 7.1.1 1 compares the values of the key transient parameters assumed in each mode of operation for Cycle 12 and the reference cycle (Cycle 11),

t.s noted in this table, the critical boron concentration for Cycle 12 is less than the corresponding Cycir: 11 values for all operating modes.

Therefore, the time to lose critical shutdown margin will increase from Cycle j

11 results due to the inverse relationship between response time and critical boron concentration.

Since all criteria were met in the Cycle 11 analysis, it is concluded that the criterion for minimum time to lose prescribed shutdown margin will be met for Cycle 12.

TABLE 7.1.1 1_

FORT CAU10UN CYCLE 12 KEY PARAMETERS ASSUMED IN THE BORON DILUTION ANALYSIS i

Parameter Cvele 11 Cvele 12 I

Critical Boron Concentration. PPM (All Rods Out. Zero Xenon) i 39de Hot Standby 1580 1560 i

Hot Shutdown 1580 1560 l

1 Cold Shutdown Normal RCS Volume 1480 1430

[

Cold Shutdown Minimum RCS Volume

• 1290 1250 l

Refueling 1400 1350 Inverse Boron k' orth. PPM /noo 09dt i

i Hot Standby 90 90 Hot Shutdown 55 55 Cold Shutdown Normal RCS Volume 55

$5 l

Cold Shutdown Minimum RCS Volume 55 55 Refueling 55

$5 Minimum Shutdown Marcin Assured. 480 t

3.2de i

Hot Standby 4.0 4.0 Hot Shutdown 4.0 4.0 Cold Shutdown Normal RCS Volume 3.0 3.0 l

Cold Shutdown Minimum RCS Volume

• 3,0 3.0 i

Refueling 1800 1800

• Shutdown Cru, A and B out, all Regulating Groups inserted except 1

cost reactive rod stuck out.

i i

. - - ~ - - -.--------~!

7.0 TPANSIENT ANALYS71 (Continued) 7.1 ANTICIPATED OPERATIONAL OCCURRENCES (Continued) 7.1.2 Excess Load Event the F.xcess Load avent was reviewed for Cycle 12 to deter-mine the pressure bias term for the TM/LP trip setpoint.

The Excess Load event is one of the DBEs analyzed to determine the maximum pressure bias term input to the TM/LP trip.

The methodology used for Cycle 12 is de-scribed in References 6 and 7.

The pressure bias term accounts for margin degradation attributable to measure-ment and trip system processing delay times.

Changes in core power, inlet temperature and RCS pressure during the transient are monitored by the TM/LP trip directly.

Consequently, with TM/LP trip setpoints and the bias term determined in this analysis, adequate protection will be providte for the Excess Load event to prevent the acceptable DNLR design limit frem being exceeded.

l l

The analysis of this event shows that a pressure bias i

term of 58.4 psia is required compared to the f.1.3 psia value in Cycle 11.

This in !* -r.t than that input from the RCS Depressu*i;acibh event, the other event for which 4 pressure bias term is calculated.

However, the current pressure bias term from the TM/LP P,[culated equa-y tion is 65 psia which bounds the 58.4 psia ca for Cycle 12.

This yields a negative 10 CFR 50.59 re-sult for this event, l'

l l

t r

l t

i l

l i

i

7.0 TRAN;'ENT ANALYSIS (Continued) 7.1 ANTICIPATED OPERA *GQNAL OCCURRENCES (Continued) 7.1.3 RCS Deoressurization Event The RCS rmpressurization event was reviewed for Cycle 12 to d ne.ruine the pressure bias term for the TM/LP sec.

point.

The RCS Depressurization event is one of the DBEs anal-yzed to determine the maximum pressure bias term input to the TM/LP trip. The methodology used for Cycle 12 is the same as that used for Cycle 11 and is described in References 6 and 7.

The evaluation of this event shows that a pressure bias term of 25.8 psia is required.

This is less than that input (to s the Excess Load event, the other event for which a pressure bias term is calculated.

Hence, the use of the Excess Load pressure bi s term in conjunction with the TM/LP trip, will provide.dequate DNBR margin for this and other A00's which require TM/LP trip protec-

tion, a

k

r 0

n 7.0 TRANSIENT ANALYSIS (Continued) 7.2 ANTICIPATED OPERATIONAL OCCURRENCES 7.2.1 CEA Withdrawal Event The CEA Withdrawal event was i unalyzed for Cycle 12 to determine the initial targins te.,t must be maintained by the LCOs such that the DNBR and fuel centerline to melt (CTM) design limits will not be exceeded in conjunction with the RPS (Variable High Power, High Pressuri:er Pres-sure, or Axial Power Distribution Trips).

The methodology contained in Reference 6 was employed in analyzing the CEA Withdrawal event.

This event is class-ified as ont for which the acceptable DNBR and center-line to melt limits are not violated by virtue of main-tenance of sufficient initial steady state thermal mar-gin provided by the DNBR and Linear Heat Rate (LHR) re-laced Limiting Conditions for Operations (LCOs).

For the HFP CEAW DNBR analysis, an MTC identical to that utilized in Reference 8 and the gap thermal conductivity consistent with the assumption of Reference 6 were used in conjunction with a variable reactivity insertion rate.

Th9 range of reactivity insertion rates examined is given in Table 7.2.1 1.

The HFP case for Cycle 12 is considered tn meet the 10 CFR 50.59 criteria since the results show that the required overpower margin is less than the available overpower margin required by the Technical Specifica-tions for DNB and PLHGR LCO's.

The zero power case was analyzed to demonstrate that acceptable DNBR and centerline melt limits are not ex-ceeded.

For the zero power case, a reactor trip, ini-tiated by the Variable High Power Trip at 29.14 (19.1%

plus 106 uncertainty) of rated thermal power, was as-sumed in the analysis.

The 10 CFR 50.59 criteria is satisfied for the HZP event if the minimum DNBR is greater than that reported in the reference cycle.

The zero power case initiated at the limiting conditions of operation results in a minimum CE 1 DNBR of 6.99 which is less than the Cycle 11 value of 7.35.

The anal-ysis shows that the fuel centerline temperatures are well below those corresponding to the acceptable fuel centerline melt limit.

The sequence of events for the zero power case is presented in Table 7.2.1-2.

Figures 7.2.1-1 to 7.2.1-4 present the transient behavior of core power, core averege heat flux, RCS coolant tempe-ratures, and the RCS pressure for the zero power case.

o 7.0 TRANSIENT ANALYSIS (Continued) 7,2 ANTICIPATED OPERATIONAL OCCURRENCES (Continued) 7,2,1 CEA Withdrawal Event (Continued)

It may be concluded that the CEA withdrawal event when initiated from the Tech. Spec. LCOs (in conjunction with the Variable High Power Trip if required) will not lead to a DNBR or fuel temperature which exceed the DNBR and centerline to melt design limits.

l I

I l

l j

}

l 1

1 i

TABLE 7.2.1 1 FORT CAIJIOUN CYCLE 12 KEY PARAMETERS ASSUMED IN THE CEA WITHDRAWAL ANALYSIS Parameter h

liZE liEE Initial Core Power Level MWt 1

102% of 1500*

Core Inlet Coolant Temperature

'F 532*

547*

Pressurizer Pressure psia 2053*

2053*

Moderator Temperature l

Coefficient x10ap/see

+0.5

+0.5**

Doppler Coefficient Multiplier 0.85 0.85 j

CEA Worth at Trip 10'2ap 5.28 6.33 Reactivity Insertion I

Rate Range x10ap/sec 0 to 1.0 0 to 1.0 CEA Croup Withdrawal Rate in/ min 46 46 Holding coil Delay Time sec 0.5 0.5

• The effects of uncertainties on these parameters were accounted for deterministically and the DNBR calculations used the methods discussed in Section 6.1 of this document and detailed in References 2-5.

• • DNBR analysis assumes MTC consistent with Reference 8.

TABLE 7.2.1-2 FORT Call 10UN CYCLE 12 SEQUENCE OF EVENTS FOR CEA WITHDRAWAL FROM ZERO POWER Time Litti Event Setooint or value 0.0 CEA Withdrawal Causes Uncontrolled Reactivity Insertion s

t 33.7 Variable High Power Trip Signal 29.1% of 1500 MWt Generated 34.1 Reactor Trip Breakers Open 34.6 CEAs Begin to Drop Into Core 35.05 Maximum Core Power 41.6% of 1500 MWt 35.92 Maximum Heat Flux 28.1% of 1500 MWt 38.98 Minimum CE 1 DNBR 6.99 40.2 Maximum RCS Pressute, psia 2230

100 i

90 1

80 y

70 8

S 60 e

a 50 5

40 a.

E g

30 20 10 0

0 10 20 30 40 50 60

TIME, SECONDS CEA Withdrawal (Zero Power)

OmahaPublicP0',erDistrict Figure CorePowervs. Time Fort Calhoun Station-Unit No. 17.2.1-1

100 i

i i

i i

90 80 8

S 70 e

60

>l 5

m 50 Qw I

40 ep 30 ua 20 10 0

0 10 20 30 40 50 60 TIME SECONDS CEA WithdrGWal IZero Power)

OmahaPublicPowerDistrict Figure CureAverageHeatFluxvs. Tite FortCalhounStation-UnitNo.i 7.2.1-2

570 i

i i

I L

560 u.

Sa g'

550 T

E H

O 5

T HE e

540 N

W

$c Tc 530 I

520 0

10 20 30 40 50 60

TIME, SECONDS i

i CEA 'n'ithdrawal (Zero Power)

DaahaPublicPowerDistrict Figure RCSTe:ceraturasvs. Tite FortCalhounStation-UnitNo.i

'/.2.1-3 l

2350 i

i i

i 2300 -

2250 2200 5

E 2150 i

5 2100 O

a:

[2050 8

2000 1950 1900 1850 0

10 20 30 40 50 60

TIME, SECONDS OsahaPublicPowerDistrict Figure CEA Withdrawal (Zero Power)

Fort Calhoun Station-Unit No i 7.2.1-4 RCSPressurevs. Time

7..

4 120 i

a

[

110 -

100 l

I 90 E

80 j

c 1

g

~

i 70 Ei f

w 60 h

50

)

E 40 J

w 8"

30 L

1 l

20 l

10 i

0 0

10 20 30 40 I

TIME, SECONDS i

i l

i i

Omah3PublicPowerDistrict figure CEA Withdrawal (Full Power)

Fort Calhoun Station-Unit No i7.2.1-5 CorePowervs. Time

120 i

I 110 -

y 100 -

S 90 E

s 80 a

70 5

d 60 e

y 50 AO 5

E 30 E

g 20 10 0

10 20 30 40 0

TIME, SECONDS OmahaPublicPowerDistrict Figure CEA Withdrawal (Full Power)

Fort Calhoun Station-Unit No. i 7.2.1-6 CoreAverageHeatFluxvs. Time

620 i

610 T'H 600 u.

8 590 a

T yg 560 A

e 4

5 570 560 m

e

[C 550 540 1

530 0

10 20 30 40

TIME, SECONDS l

CEA Withdrawal (full Power)

DaahaPublicPowerDistrict l Figure RCSTe.coeraturesvs. Tite FortCalhounStation-UnitNo.il7.2.1-7

4 2250 i

i i

2200 2150 2100 2050 E'

5 2000 Wc

[1950 e

1900 1850 1800 1750 0

4.0 20 30 40 TIME, SECONDS CEA Withdrawal (Full Power)

OsahaPublicPowerDistrict Figure RCSPressurevs. Tite Fort Calhoun Station-Unit No. i 7.2.1-8

7.0 TRANSIENT ANALYSIS (Continued) 7.2 ANTICIPATED OPERATIONAL OCCURRENCES (Continued) 7.2.2 Loss of Coolant Flow Event The Loss of Coolant flow event was reanalyzed for Cycle 12 to determine the minimum initial margin that must be maintained by the Limiting Conditions for Operations (LCOs) such that in conjunction with the RPS low flow trip, the DNBR limit will not be exceeded.

The event was analyzed parametrically in initial axial shape and rod configuration using the methods described in Reference 6 (which utilizes the statistical combina.

tion of uncertainties in the DNBR analysis as described in Appendix C of References 4 and 5).

The 4. Pump Loss of Coolant Flow produces a rapid ap.

proach to the DNBR limit due to the rapid decrease in the core coolant flow.

Protection against exceeding the DNBR limit for this transient is provided by the initial steady state thermal margin which is maincained by ad.

hering to the Technical Specifications' LCOs on DNTR margin and by the response of the RPS which provides an automatic reactor trip on low reactor coolant flow as measured by the steam generator differential pressure transmitters.

The flow coastdown is generated by CESEC.III (References 9 and 10) which utilizes implicit modeling of the reac.

tor coolant pumps.

This coastdown is shown in Figure 7.2.2 1.

Table 7.2.2 1 lists the key transient para.

meters used in the Cycle 12 analysis and compares them to the reference cycle (Cycle 11) values.

The low flow trip setpoint is reached at 2.80 seconds and the scram rods start dropping into the core 1.15 seconds later.

A minimum CE.1 DNBR of 1,43 is reached at 4.56 seconds.

Figures 7.2.2 2 to 7.2.2 5 present the core power, heat flux, core coolant temperatures, and RCS pressure as a function of time.

It may be concluded that for Cycle 12 the Loss of Flow event when initiated from the Tech. Spec. LCOs in con.

junction with the Low Flow Trip, will not exceed the minimum DNBR design limit.

TABLE 7.2.2 1 FORT Call 10UN CYCLE 12 KEY PARAMETERS ASSUMED IN THE LOSS OF COOLANT FLOW ANALYSIS Parameter h

Cvele 11 Cvele 12 Ir.: tial Core Power Level MWe 1500*

1500*

Initial Core Inlet Coolant Temperature

'F 545*

545*

Initial RCS Flow Rate gpm 208,280*

208,280*

Pressurizer Pressure psia 2075*

2075*

10 4,j.F

+0.5

+0.5 Moderator Temperature Coefficient 3

Doppler Coefficient Multiplier 0.85 0.85 LFT Analysis Setpoint t of initial flow 93 93 LFT Response Time sec 0.65 0.65 CEA Holding coil Delay sec 0.5 0.5 CEA Time to 1004 Insertion sec 3.1 3.1 (Including Holding Coil Delay)

CEA Worth at Trip (all rods out) top 6.85 6.50 Total Unrodded Radial Peaking 1.80 1.80 Factor (F T) g

• The uncertainties on these parameters were combined statistically rather than deterninistically.

The values listed represent the bounds included in the statistical combination.

s TABLE 7 ' M FORT CA11tOUN CYCLE 12 SEQUENCE OF EVENTS FOR LOSS OF FLOW Tlee ( S e e ),

h Setroint or Value 1.0 Loss of Power to all Tour Reactor Coolant Pumps 2.80 Low Flow Trip Signal Generated 936 of 4. Pump Flow i

3.35 Trip Breakers open 4.0 Shutdown, CEAs Begin to Drop into Core i

4.56 Minimum CE 1 DNBR 1,43 l

l 6.4 Maximum RCS Pressure, psia 2113 l

I l

i I

i e

i l

I 1

i I

l

1.0 i

0.9 0.8 0. 7,-

5 I

x

{

0.6 e

u.

0.5 x

3 u.

g 0.4 8

0.3 0.2 0.1 I

I 0

0 5

10 15 20 25 30 TIME SECONDS NOTE' CYCLE 12 LossofCoolantFlow OmahaPublicPowerDistrict Figure CoreFlowFractionvs. Tite FortCalhounStation-UnitNo.i 7.2.2 1

F 1,

110 i

i i

i i

t 100 90 i

x" 80 I

c oa 70 i

to e

I L

60 a

.4 4

50 eu 3

Oa 40

+

tu 6

cr i

I 8

30 l

20 I

1

)

10 i

i l

i 0

0 5

10 15 20 25 30 l

L TIME SECONDS i

I I

i i

(

I t

N0iE:

C'.C!.E 12 1

I i.

LossofCoolantFlow 0:ahaPublicPowerDistrict Figure 200POWerV3.Ii:0

,FCrtC31hCunSi3tiCD-Unit 50. !;7.2.2-2 ;

i t

6

--__,m_._._

4 110 1

i i

i I

100 l

a E

90 O

Om 80 u.a

,0 a

J 5

60 i

d I

Q 50 J

w I

i w

40 e

4 Cw 4

30 w

1 20 a

u i

j 10 0

0 5

10 15 20 25 30 TIME SECONDS 1

l NOTE:

CYCLE 12 i

I I

l 1

LossofCoolantFlow 0.T.anaPublicPowerDistrict Figure {

l CCTEAVEr690HeatI'UXVs.Ii:E i I0rt Calh0Cn $tati:n-Unit.10.1 l 7.2.2-3 3

610 600 590 s

u.

580 H

a y

570

\\

q a E N

-t 5

560 550 m

o l

u 540 l

530 '

520 0

5 10 15 20 25 30 TIME SECONDS NOTE:

CYCLE 12 LossofCoolantFlow C>anaPublicP0',serDistrict Figure RCSTe:Deraturesvs.!!:.e iFortCalhounStation-unitNo,ii7.2.2-4 i

=

j 2200 i

2150-i i

i 2100 l'

t-4 2 2050 w

C l

?n 2000 J

mw C

C.

1950 moC 1900 l

1850 u

1800 0

5 10 15 20 25 30 l

TIME SECONDS 1

1 l

l l

NOTE:

CYCLE 12 i

LossofCaolantFlow 0:anaPublicPowerDistrict Figure EC3Fr055"re',3.Ii:0 F0rt C31hCun $t3tiO3-Dit M.1 ! 7.2.2-5 L

f 7.0 TPANSIENT ANALYSIS (Continued)

I 7.2 ANTICIPATED OPERATIONAL OCCURRENCES (Continued) 7.2.3 Full teneth CEA Oroo Event The Full Length CEA Drop event was reviewed for Cycle 12 to p

determine the initial thermal margins that must be maintained r

by the Limiting Conditions for Operation (LCOs) such that the DNBR and fuel centerline to melt design limits will not be exceeded.

l This event was analy=ed parametrically in initial axial shape and rod configuration using methods described in Reference 6.

l The transient was conservatively analy=ed at full power with an ASI of 0.182, which is outside of ene LCo limit of 0.06.

i This results in a minimum CE-1 DNBR of 1.45.

A maximum allow.

able initial linear heat generation rate of 18.5 W/f t could l

exist as an initial condition without exceeding the accept.

I able fuel centerline to melt limit of 22 W/ft during this transient. This amount of margin is assured by setting the l

J Linear Heat Rate related LCOs based on the more limiting i

allowable linear heat rate for LOCA.

i i

i The CEA drop incident was reviewed for Cycle 12 and found to

(

be bounded by Cycle 11.

Since a negative 10 CFR 50.59 deter.

j mination was made for Cycle 12, the conclusions from Cycle 11

[

remain valid and applicable to Cycle 12.

1 7.3 POSTULAT*D ACCIDENTS l

l 7.3.1 CEA Ejection i

The CEA Ejection event was reviewed for Cycle 12. A sumnary containing the results of the analysis was submitted in Refer.

l enee 11 for Cycle 11 and has been validated for use in Cycle l

3 12.

I L

Sinca a negative 10 CFR 50.59 determination was made for the l

Cycle 12 CEA Ejection event, no reanalysis was performed.

6 i

7.3.2 Ste n Line Break Accident j

This accident was evaluated for Cycle 12 using the methodol-ogy discussed in References 6 and 12.

The Steam Line Break j

accident was previously analyzed in the Fort Calhoun FSAR and a

]

satisfactory results were reported therein.

The Steam Line Break accidents at both HZP and HTP vere examined in the re-(

l j

ference cycle (Cycle 8) safety evaluation with acceptable re-l suits obtained.

Both tbs FSAR and reference cycle evalua.

i tions are reported in the 1986 update of the Fort Calhoun

{

Station Unit No. 1 USAR I

The Cycle 12 Full Power Steam Line Break accident vap evalu-(

3

)

sted for a more neg@tive ef fective MTC of 2.7 x 10**ar/* F l

1 than the 2.5 x 10'"ap/'F value that was used in the Cycle j

8 analysis. However, the cooldown curve for Cycle 12 is l

l) f

7.0 TRANSIENT ANALYSIS (Continued) 7.3 POSTULATED ACCIDENTS (Continued) 7.3.2 Steam Line Break Accident bounded by that of Cycle 8 (as shown in Figure 7.3.21).

Thisfigureshowsthatthereactivityipsertionforthe cycle 12 core with an MTC of 2.7 x 10' ap/'F due to a Steam Line Break accident at full power is substan.

tially less than the value used in the Cycle 8 analysis.

(This smaller reactivity insertion is due to the use of the 01T cross sections which are valid for a range of moderator temperatures from room temperature to 600'K while the analyses prior to Cycle 9 vere performed with cooldown curves derived by conservatively extrapolating CEPAK cross section values to low temperatures.) The fuel temperature coefficient used in the Cycle 8 anal-ysis is conservative with respect to the fuel tempera-ture coefficient calculated for the Cycle 12 core includ-ing uncertainties.

The Cycle 12 minimum available shut-down worth is 6.53 tap compared to a Cycle 8 value of 6.68 tap.

The reduction of 0.15 tap in scram worth from Cycle 8 to Cycle 12 is offset by the 0.986 ap reduction in moderator cooldown reactivity. Tha net gain assures that the overall reactivity insertion for a Cycle 12 Steam Line Break is less than that of the i

reference cycle analysis.

Therefore, the return to power is less than that of the reference cycle and Cycle 6

1 FSAR analyses, i

A similar evaluation was performed for the Zero Power SteamLineBreakaccideng,p/*Fshowsasubstantial-Again the Cycle 12 cooldown for an MTC of 2.7 x 10' o ly smaller reactivity insertion than was used in the I

Cycle 8 analysis (as seen in Figure 7.3.2-1).

Since the I

minimum available shutdown margin for Cycle 12 remains l

unchanged from the reference cycle value (4 top), the t

overall reactivity insertion for the Cycle 12 Steam Line l

Break accident vill be substantially less than that of the reference cycle.

Therefore, the consequences of a zero power Steam Line Break accident for Cycle 12 will be less severe than that reported for the reference cycle and the FSAR (Cycle 1) cases.

Based on the evaluation presented above, it is concluded that the consequences of a Steam Line Brr,ak accident ini-tiated at either zero or full power are less severe than the reference cycle and FSAR (Cycle 1) cases.

Since a negative 10 CFR 50.59 aetermination was made for the Cycle 12 Steam Line Break Accident, no reanalysis was performed.

l l

8 i

5 E

6

\\

\\

FULL POWER d

h CYCLE 8 CYCLE 1 4

H s

CYCLE 8 e

CYCLE 1 p

2 CYCLE 12 N.,

s,

>-r Nsy E

\\

b ZERO POWER

\\

0 ec:

l g

200 300 400 500 600 700 CORE AVERAGE MODERATOR TEMPERATURE F

SteenLineEreakIncident 0:SnaPublicPaerDistrict figure :

Reactivityvs.ModeratorTe:erature Fort Calhoun Staticn-Unit No. i 7.3.2-! ]

O o

7.0 TRANSIENT ANALYSIS (Continued) 7.3 POSTULATED ACCIDENTS (Continued) 7.3.4 Seized Rotor Event The Seized Rotor event was evaluated for Cycle 12 to dem.

onstrate that only a small fraction of fuel pins are pre.

dicted to fail during this event.

Cycle 12 is bounded by the reference cycle (Cycle 9) analysis because an F T of 1.85 was assumed in the Cycle 9 analysis and R

the Cycle 12 Technical Specification of 1,80 remains conservative with respect to the F T value used in the R

Cycle 9 2nalysis.

Therefore, the total number of pins predicted to fail will continue to be less than 14 of all of the fuel pins in the core.

Based on this result, the resultant site boundary dose would be ws11 within the limits of 10 CTR 100.

Since a negative 10 CFR 50.59 determination was made for the Cycle 12 Seized Rotor Event, no reanalysis was per-formed.

i I

i 7

8,0 ECCS PFRF0PXANCE ANALYSIS I

Both Cycle 11 Lart,e and Small Break Loss of Coolant accident analyses I

were performed using the methodology discussed in Reference 1.

A su:.

mary containing the results of the an.tlyses was subrsitted in Reference j

2.

The Cycle 11 revised ECCS analysis was verified to be valid for use in Cycle 12 given the bounding input assumptions.

I since a negative 10 CTR 50,59 dete-mination was made for the Cycle 12 CCCS analysir., no reanalysis was performed.

l l

f h

l i

I h

I

(

t t

l L

i I

f

4 l

9.0 STARTUP TESTING t

The startup testing program proposed for Cycle 12 is identical to that used in Cycle 11.

It is also the same as the program outlined in the Cycle 6 Reload Application, with two exceptions.

First, a CEA exchange technique (Reference 1) for zero power rod worth measurements will be performed in accordance with Reference 2. replacing the boration/ dilution method. Also, low power CECOR flux maps and psuedo. ejection rod measurements will be substituted for the full core symmetry checks, The CEA exchange technique is a method for measuring rod worths which is both faster and produces less vaste than the typical boration/ dilution L

method.

Cycle 11 startup testing exclusively used the CEA exchange tech.

I nique.

Resui.ts from the CEA exchange technique were within the acceptance and review criteria for low power physics parameters.

The combination of the pseudo. ejection technique at zero power and low power CECOR maps pro.

l vides for a less time consuming but equally valid technique for detectitig l

azimuthat power tilts during reload core physics testing.

The psuedo.ejec.

tion rod measurement involves the dilution of a bank into the core, borating a C EA ou t, and then exchanging (rod swap) the CEA against other symmetric CEA's within the bank to measure rod worths.

The acceptance and review criteria for these tests are:

lttg Acceptance Criteria Review Criteria CEA Croup Vorths i 15% or predicted i 15% of predicted Pseudo ejection None The greater of: 2.5C l

rod vorth mea.

deviation from group suremett avarage or 15% deviation from group average, j

Low Pos er CECOR Technical Specifica.

Azimuthat tilt less than i

naps tion limits on F T,

204, l

R F

T xy and Tq OPPD has reviewed these tests and has concluded that no unreviewed safety question exists for implementation of these procedures.

i I

l t

l l

4 10,0 RErERENCES

(

i References (Chante rs.1%

t 1.

Letter A. Bournia (NRC) to R. L. Andrews (OPPD). datid March 11,

1988, i

2.

"Omaha Public Power District Batch M Reload Fuel Design Report."

CEN.347(0).P. November 1986, I

l 3.

Letter LIC.86 677. R. L. Andrews (OPPD) to A. C. Thadani (NRC),

i dated December 15, 1986, i

l 4

"Ceneric Mechanical Cesign Reptrt for Exxon Nuclear Fort Calhoun 14 x 14 Reload Tuel Assembly," M.NT.79 70.P, September 1979.

5.

"Extended Burnup Report for Fort Calhoun !teloads XN.4 and XN.5 l

(Batches K and L) " ANT.87 139(P), October, 1987.

[

6 "Qualification of Exxon Nuclear Tuel for Extended Burnup,' XN.NP.

82 06(P)(A) & Supplements 2, 4 6 5. Revision 1. October 1986, I

l 7.

"Omaha Public Power District Reload Core Analysis Methodology -

f Transient and Accident Methods and Verification," OPPD.NA.8303.P.

Revision 02, April 1984.

l t

8.

"Omaha Public Power District Reload Core Analysis Methodology

[

l Overview " OPPD.NA 8301.P. Revision 03 April 1988.

(

1 9

"Omaha Public Power District Reload Core Analysis Methodology.

Neutronics Design Methods and Verification," OPPD.NA.8302.P.

Revision 02 April 1988.

L

g s

10.0 EEFERENCES (Continued) ggf,gyances (Chanter 6) 1.

"*0RC Code, A Computer Code for Determining the Thermal Margin of a

'e. tor Core," 'ENPD *. ', July 1975.

2.

al Heat Flux Correlation For CE Fuel Assemblies with

• Spacer Grids, Part 1 Uniform Axial Power nistribution,"

12 PA Apr!1 1975.

3.

. D Code Structure and Modeling Methods for Calvert Cliffs I

.cs 1 and 2," CEN-191(B).P, December 1981.

4.

Safety Evaluation by the Office of Nuclear Reactor Regulation Sup-porting Amondment No. 70 to Facility Operating License No. DPR-40 for the Omaha Pubiic Power District, Fort Calhoun Station, Unit No. 1, Docket No. 50 285, March 15, 1983.

5.

Safety Evaluation by the Office of Nuclear Reactor Regulation Sup-porting Amendment No. 77 to Facility Operating License No. DPR 40 for the Caaha Public Power District, Fort Calheun Station, Unit No. 1, Docket No. 50 285, April 26, 1984 6.

Safety Evaluation by the Office of Nuclear Reactor Regulation Sup-porting Amendment No. 92 to Facility Operating License No. DIR 40 for the Omaha Public Power District, Fort Calhoun Section, Unit No. 1, Decket No. 50 285, November 29, 1985.

7.

Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment No. 109 to Facility Operating License No.

DPR 40 for Omaha Public Power District, Fort Calhoun Station, Unit No. 1 Docket No. 50 285, May 4, 1987.

8.

"Omaha Public Power District Reload Core Analysis Methodology Overview," OPPD NA 8301 P, Revision 02, November 1986.

9.

"Statistical Combination of Uncertainties, Part 2," Supplement l P, CEN 257(0).P. August 1985, 10.

Safety F. valuation P.eport on CENPD 207-P A, "CE Critical Heat Flux:

Part 2 Non Uniform Axial Power Distribution," letter, Cecil Thomas (NRC) to A. E. Scherer (Combust!" Engineering),

November 2, 1984.

10.0 REFERENCES

(Continued)

References (Chanter 7) 1.

"Amendment to Operating License DPR 40, Cycle 11 License Applica-tion", Docket No. 50 285, May 4, 1987.

2.

"Scatistical Combination of Uncertainties Methodology, Part 1:

Axial Power Distribution and Thermal Margin / Low Pressure LSSS for Fort Calhoun", CEN 257(0) P, November 1983.

t 3.

"Statistical Combination of Uncertainties Methodology, Part 2:

Combination of System Parameter Uncirtainties in Thermal Margin Analysis for fort Calhoun Unit 1", CEN 257(0).P. November 1983.

4 "Statistical Combination of Uncertaintics Methodology, Part 3:

Departure from Nucleate Boiling and Linear Heat Rate Limiting Conditions for Operation for Fort Calhoun", CEN 257(0).P.

November, 1983.

5.

"Statistical Combination of Uncertainties Methodology for Fort Calhoun," CEN 257(0)-P Supplement 1 P, August 1985.

6.

"Omaha Public Power District Reload Core Analysis Methodology -

Transient and Accident Methods and Verification", OPPD NA 8303-P, Revision 02, April 1988.

7.

"CE Setpoint Methodology", CENPD-199 P, Rev. 1-P, March 1982.

8.

"CEA Withdrawal Methodology", CEN 121(B)-P, November 11,9.

9.

"CESEC, Digital Simulation of a Combustion Engineering Nuclear Steam Supply System", Enclosura 1 P to LD 82-C01, January 6, 1982.

10.

"Response to Questions on CESEC", Louisiana Power and Light Company, Waterford Unit 3, Docket 50-382, CEN-234(C).P December 1982.

11.

Letter LIC 86 675, R. L. Andrews (OPPD) to A. C. Thadani (NRC),

dated *anuary 16, 1987.

12.

"Omaha Public Power District Reload Core Analysis Methodology -

Neutronics Design Methods and Verification", OPPD NA 8302 P, Revision 02, April 1988.

s

10.0 REFERENCES

(Continued)

References (Chanter 8) 1.

"Omaha Public Power District Reload Core Analysis Methodology -

Transient and Accident Methods and Verification", OPPD-NA 8303-P, Rev. 02, April 1988.

2.

Letter LIC 86 675, R. L. Andrews (OPPD) to A. C. Thadani (NRC),

dated January 16, 1987.

References (Chanter 9) 1.

"Control Rod Croup Exchange Technique." CEN-319, November 1985.

2.

"Acceptance for Referencing of Licensing Topical Report CEN 319 -

Control Rod Croup Exchange Technique," letter, Dennis M.

Crutchfield (NRC) to Rik W. Wells (Chairman CE Owners Croup),

April 16, 1986.

ATTACHMENT B Justification, Discussion, and Significant Hazards Considerations for Cycle 12 Reload l

l l

t I

I l

i

The Fort Calhoun Technical Specifications are being amended to reflect changes which are a result of the Cycle 12 core reload.

Table B-1 presents a summary of i

the Technical Specification changes and the explanation for the changes. Justi-fication for the changes is contained in the attached Fort Calhoun Cycle 12 Core Reload Evaluation.

TABLE B-1 Exclanation for Cycle 12 Technical Soecification Chances Tech. Soec. No.

Chance Reasons 1)

Page 1-2 Change total unrodded planar Revised value is conserva-radial peak from 1.85 to 1.80.

tive with respect to previ-ous value of 1.85.

The re-duced value will providt ad-ditional operating margir.

2)

Figure 1-3 Replace Figure 1-3 with en-The Cold leg Temperature closed Figure 1-3.

limit has been changed from 545'F to 543'F. This has resulted in a change to the Alpha, Beta and Gamma term of the TM/LP equation.

3)

Page 1-5 Change the reference from FSAR Reflect updated reference.

to USAR.

4)

Page 1-6 "strady" to "steady" Corrected typo.

5)

Figure 2-6 Replace Figure 2-6 with en-The LHR-LCO has been re-closed Figure 2-6.

vised to reflect the use of the more limiting LOCA R0PM limit in the analysis vs.

the transient analysis R0PM. (see Letter LIC 620, K. J. Morris (OPPD) to NRC Document Control Cesk, dated July 25,1988).

6) Figure 2-9 Replace Figure 2-9 with en-The F T and FR closed Figure 2-9.

limit!yas a function of l

power have been revised to I

maintain consistency with changes to items 1 and 5, above.

7)

Page 2 56 Add... "The linear heat rate This change clarifies how shall be monitored by the in-the linear heat rate should core detector system in accor-be monitored and what para-dance with Specifications meters apply to bound the 2.10.4(1)(a) or 2.10.4(1)(b),

limits, or maintain the Axial Shape Index Y Figure 2g, with the limits of 6 in accordance with Specification 2.10.4(1)(c)."

after2.10.4(1)5.

TABLE B-1 (Continued)

Tech. Spec. No.

Chance Reasons 8)

Page 2-57 Add "... for seven days from Clarify the point at which the date of the last valid the ex-core LHR-LC0 is power distribution..." to entered during operation.

Specification 2.10.4(1)(b).

9)

Page 2-57 Add "... and maintain the Axial This is to clarify the re-Shape index, Y, within the quirements of maintaining i

limits of Figure 2-6..." to the the Axial Shape Index first sentence of Specification within the requirements of 2.10.4(1)(c).

Figure 2-6.

T 10)

Page 2-57a Change F 1 1.80.

Revised value is conserva-xy tive with respect to pre-vious value of 1.85.

The reduced value will provide additional operating mar-gin.

11)

Page 2-57c Change "1545'F" to s "543'F" The Cold Leg Temperature is being changed to more accurately reflect actual operating conditions and to gain additional margin.

12)

Page 2-57c Change "545'F" to "543'F" The Cold Leg Temperature Footnote **

and "547'F" to "545'F" is being changed to more accurately reflect actual operating conditions and to gain additional margin.

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o DesgriptionofAmendmentRequeststoReducethePlanarRadialPeakingFactor F

t 1.80:

xy The proposed Technicc1 Specification changes in Table B-1 corresponding to Items 1, 5, 6 and 10 for Technical Specifications Section 1.1, Figure 2-6 Figure 2 9 9

and Section 2.10.4(3) on Page 1-2 and 2-57 concern reducing the F value xy from 1.85 to 1.80.

An error in the Cycle 11 setpoint evaluation has reduced the excore MR LCO tent from 90% power to 80% power during operation with the excore detectors.

By re-ducing F T,

dditiona{operatingmarginisgainedintheMR-LCOoperating TNEF and F limits as a function of power in Technical Spec-tent.

R y

ifications Figure 2-9 have been revised to maintain consistency with the change to Figure 2 6.

l Basis for No Sienificant Hazards Determination:

1 This proposed amendment does not involve a significant hazards consideration because the operation of Fort Calhoun Station in accordance with this amendment would not:

1)

Involve a significant increase in the probability or consequences of an f

accident previously evaluated.

Thischangemerelyallowsutilizagionof the additional margin available with the reduction of maximum F value with no changes in administrative specifications. Onthe5asisof x

technical safety evaluation, operating with gain in margin for Cycle 12 MR LCO would be no more limiting than operating with the Cycle 11 MR-LCO.

Therefore, this change does not increase the probability or conse-quences of a previously evaluated accident.

2)

Create the possibility of a new or different kind of accident from any accident previously evaluated.

It has been determined that a new or different type of accident is not created because no new or different modes of operation are proposed for the plant. The continued use of the same Technical Specification adeinistration controls prevents the possi-bility of a new or different kind of accident.

3)

Involve a significant reduction in a margin of safety. Administrative specifications involving the MR LCO engure that operating with the extra margin gafand from the reduction of F conforms to current plant xy conditions and, therefore, preserves the margin of safety.

Reducing the MR LCO tent does not affect the available margin and, therefore, will not reduce the margin of safety.

Based on the above considerations, OPPD does not believe that this amendment in-vc,1vea a significant hazards consideration.

x

Description of Amendment Requests Reducing Cold Leg Temperatures to 543*F:

The proposed Technical Specification changes in Table B-1 corresponding to Items 2,11 and 12 for Technical Specifications Figure 1-3,2.10.4(5) on page 2-57c and Footnote ** on Page 2-57c concern lowering the current cold leg temperature (T ) from 545'F to 543*F.

c The operation of the unit with a reduced cold leg T all provide additional c

margin for the TM/LP P equation.

The Alpha, Beta ar.d Gamma terms of the var TH/LP P trip 9quation were optimized given the rWuced allowable T and theuncEaranged FR operating parameters.

c All of the safety analyses and Cycle 12 design analyses were calculated at 545'F for conservative reasons; this bounds the use of a 543'F inlet temperature dur-ing Cycle 12 operations.

Basis for No Sianificant Hazards Determination This proposed amendment does not involve a significant hazards consideration because the operation of the Fort Calhoun Station in accordance with this amendment would not:

1)

Involve a significant increase in the probability or consequences of an accident previously evaluated.

This change allows the reduction of T to 543*F.

Thetemperaturechangeisboundedbytheprevioustechnicai safety analysis which addressed the 545'F inlet temperature.

Therefore, this change does not increase the probability or consequences of a previ-ously evaluated accident.

2)

Create the possibility of a new or different _ kind of accident from any accident previously evaluated.

It has been determined that a new or different kind of accident is not created because no new or different modes of operation are proposed for the plant.

The continued use of the same Technical Specification administrative controls prevents the possi-bility of a new or different kind of accident.

3)

Involve a significant reduction in a margin of safety. Administrative specifications involving T ensure that operating at a T of 543'F c

c j

conforms to current plant conditions and, therefore, preserves the margin of safety.

The temperature change is bounded by previous technical safe-ty analysis which addressed the 545'F inlet temperature and, therefore, will not reduce the margin of safety.

1 Based on the above considerations, OPPD does not believe that this amendment in-l volves a significant hazards consideration.

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Description of Amendment Requests for Changing References from FSAR to USAR:

The proposed Technical Specification changes in Table B-1 corresponding to Item 3 for Technical Specification 1.2 on page 1-5 concern changing all references mentioning "FSAR" to the correct reference "USAR."

One of the numerous post-THI related changes was to require that all licensed commercial nuclear power plants perform an annual revision to the FSAR.

This updated FSAR became officially recognized as the USAR (Updated Safety Analysis Report) to avoid any confusion with the FSAR.

Needed reference changes in the Technical Specifications are generally made at the time when the related change is made.

Basis for No Sianificant Hazards Determination:

This proposed amendment does not involve a significant hazards consideration because the operation of Fort Calhoun Station in accordance with this amendment would not:

1)

Involve a significant increase in the probability or consequences of an accident previously evaluated.

This change merely allows the Technical Specifications to reference the proper updated document with no changes in administrative specifications. Therefore, this change does not in-crease the probability or consequences of a previously evaluated acci-dent.

2)

Create the possibility of a new or different kind of accident from any accident previously evaluated.

It has been determined that a new or different kind of accident is not created because no new or different modes of operation are proposed for the plant.

The continued use of the same Technical Specification administrative controls prevents the possi-bility of a new or different kind of accident, j

3)

Involve a significant reduction in a margin of safety.

Administrative specifications involving the referencing of the USAR will not reduce the margin of safety.

Based on the above considerations, OPPD does not believe that this amendment in-volves a significant hazards consideration.

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Description of Amendment Request for Correcting a Typographical Error:

The proposed Technical Specification changes in Table B-1 corresponding to Item 4 for Technical Specifications 1.31(1) on Page 1-6 concerns correcting a typo-graphical error by changing the word "strady" to "steady."

During the evaluation of Technical Specification changes for cycle 12, a mis-spelled word was discovered in Technical Specification 1.3(1).

The word in question is spelled "strady," however, the correct spelling of the word is "steady."

This error is obviously typographic in nature and, therefore, poses no significant hazards consideration.

Basis for No Sirnificant Hnzards Determination:

This proposed amendment does not involve a significant hazards consideration because the operation of Fort Calhoun Station in accordance with this amendment would not:

1)

Involve a significant increase in the probability or consequences of an accident previously evaluated.

This change merely allows for correct spelling of a word.

Therefore, this change does not increase the pro-bability or consequences of a previously evaluated accident.

2)

Create the possibility of a new or different kind of accident from any accident previously evaluated.

It has been determined that a new or different kind of accident is not created because no new or different modes of operation are proposed for the plant.

The continued use of the same Technical Specification administrative controls prevents the possi-bility of a new or different kind of accident.

3)

Involve a significant reduction in a margin of safety. Neither this typo-graphical error nor its correction will reduce the margin of safety.

Based on the above considerations, OPPD does not believe that this amendment in-volves a significant hazards consideration.

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4 Description of Amendment for Revising Section 2.10.4:

The proposed Technical Specification changes in Table B-1 corresponding to Items 7, 8, and 9 for Technical Specification Section 1.10.4 concern changes to in-structions for entering into the excore LHR-LCO.

A review ot' Technical Specification 2.10.4 with the NRC Senior Resident Inspec-tor indicated that the requirements for entering into the excore LRR LCO (Figure 2 6) were unclear.

The changes made herein more accurately define when the LHR LCO should be entered, to allow sufficient time fer a power reduction to the maximum power allowed by Technical Specification Figure 2 6, should the reactor be in excess of that power level at the time the LHR LCO was entered.

Basis for No Sienificant Hazards Determination:

This proposed amendment does not involve a significant hazards consideration because the operation of Fort Calhoun Station in accordance with this amendment would not:

j 1)

Involve a significant increase in the probability or consequences of an accident previously evaluated.

This change clarifies the point at which the LHR-LCO (Figure 2-6) must be entered and provides better guidance for plant operation. The basis for the technical safety evaluation would be no more limiting than operating with the Cycle 11 basis.

Therefore, this change does not increase the probability or consequences of a previously evaluated accident.

t 2)

Create the possibility of a new or different kind of accident from any accident previously evaluated.

It has been determined that a new or different type of accident is not created because no new or different modes of operation are proposed for the plant.

The continued used of the Technical Specification administrative controls prevents the possibility of a new or different kind of accident.

3)

Involve a significant reduction in a margin of safety. Administrative specifications involving the LHR LCO ensure that the operators enter the LCO with sufficient time to reduce power, if necessary, prior to e

utilizing the excore instruments to monitor core power.

The changes have been implemented through strict administrative procedures and, therefore, will not reduce the margin of safety.

Based on the above considerations, OPPD does not believe that this amendment involves a significant hazards consideration.