Text

Amend 43 to Facil Oper Lic DPR-40,changing Tech Spec IAW Cycle 5 Reload Analysis & Deleting Tech Specs Permitting Partial Insertion of Control Element Assemblies
	ML20062F186
Person / Time
Site:	Fort Calhoun
Issue date:	12/05/1978
From:	Reid R; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20062F178	List: ML20062F172; ML20062F186; ML20062F192; ML20062F200;
References
	NUDOCS 7812150114
	Download: ML20062F186 (25)
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/'O % UNITED $TATES

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/4 NUCLEAR REGULATORY COMMISSION

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) WASHING 7oN. D. C. 20565 '

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gv.....f OMAHA PUBLIC POWER DISTRICT DOCKET NO. 50-285 FORT CALHOUN STATION, UNIT NO. 1 MENDMENT TO FACILITY OPERATING LICENSE Amenoment No. 43 License No. OPR-40

1. The Nuclear Regulatory Comission (the Comission) has found that:

l A. The applications for amendment by Omaha Public Power District (the licensee) notarized March 21 and August 3,1978, as revised by letters dated October 31, November 7 and 27,1978, l comply with the standards and requirements of the Atomic ,

Energy Act of 1954, as ame :ed (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I;

8. The facility will operate in confomity with the applications, '

the provisions of the Act, and the rules and regulations of r the Comission; '

C.

There is reasonable assurance (1) that the activities authorized by this amendthent can be conducted without endangering the health and safety af the public, and (ii) that such activities will be i conducted in compliance with the Comission's regulations; D.

The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable rtquirements have been satisfied.

7 81 '? 15 01 l'[i

o 2

2. Accordingly, the license is amended by changes to the Technical Speci'ications as indicated in the attachment to this license anendnent, and paragraph 3.B. of Facility Operatir.g License No. DPR-40 is hereDy amended to read as follows:

B. Technical Specifications

! The Technical Specifications contained in Appendices i

A and B, as revised through Amendment No. 43, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the l

Technical Specifications.

3. This license amendment is effective as c' the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION S"

/. '

Robert W. Reid, Chief Operating Reactors Branch #4

)

Division of Operating Reactors j

Attachment:

i Changes to the lechnical Specifications Date of Issuance: December 5, 1978

i ATTACHMENT TO LICENSE AMENDMENT NO. 43 FACILITY OPERATING LICENSE NO. DPR-40 DOCKET NO. 50-285 Replace the following pages and figures of the Appendix "A" Technical Specifications with the enclosed pages and figures. The revised pages are identified by Amendnent number and contain vertical lines indicating the area of change.

Remove Pages Insert Pages 11 11 111 fii 2 2 1-2 1-2 2-18 2-18 2-20 2-20 2-22 2-22 2-38 2-38 2-50 2-50 2-50f 2-50f 2-51 2-51 2-56 2-56 .

2-57 2-57 2-57a thru c 2-57a thru c 2-62 2-62 2-66 2-e6 44 4-4 6-4 -

Remove Figures Insert Ficures 2-6 2-6 2-8 3 2-9 2-8 & 2-9

t r

l TABLE OF CC'ITE:lTS (Continued)

Page l 3. 0 SURVIILUJICE RIC.UIPIC."!S . ....................................... 3-1 J

3.1 Instru=entation and Centrol ................................ 3-1 l 32 Equipment and Sa=pling Te st s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-17 3.3 Reactor Coolant System Surveillance ........................ 3-21 3.h Reactor Coolant System Integrity Te sting . . . . . . . . . . . . . . . . . . . 3-36 35 contain=ent Tes: ........................................... 3-37 ;

3.6 Sarety Injection and Contain=ent Ccoling Syste=3 Tests . . . . . 3-5h 3.7 E=crgency Pover Sys:c= Pericdie Tests ...................... 3-58 i

3. 8 Main Stes: Is olati on valve s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-61 39 Auxiliary Feedvater Syste: ................................. 3-62 3 10 Reactor Core Pars =eters..................................... 3-63 l 3.11 Environ = ental Radiolegical Monitoring . . . . . . . . . . . . . . . . . . . . . . 3-6h .

3.12 Radio ac t ive Mat e ri als . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-69 3.13 Radioactive Material Scurces Surveillan:e . . . . . . . . . . . . . . . . . . 3-76 3.1h Shock Sdppressors (Snubbers) ............................... 3-77 3.15 Fire Prc ectic: Syste: ..................................... 3-80 8 h.0 o rS .G.i :--

va. J s,.---

a .................................................... h-1 t

h.1 Site ..................................................l..... h-1 i

, h.2 Contain=ent Design Features ................................. h-1 'L h.2.1 'sntain=ent S:ructure .............................. k-1 t h.2.2 Penetrations ....................................... h-1 l h.2.3 Centain=en Stru: ure Cooling Systems .............. h-2 l h.3 Nuclear Stes: Supply Syste: (NSSS) .......................... h-3 l h.3.1 Reseter Cociant Syste: ............................. b-3 !

h. 3. 2 Reacter Core and C:ntr:1 ........................... k-3

k.3.3 Energency Core Cooling ............................. h-3 l h.k Fuel Sterage ................................................ h-h I h.h.1 New Tuel Sterage .......................%7.~......... h a j h.h.2 Spent Fuel Sterage ................................. h L i k.5 Seisni: resign rcr clas s I syste=s . . . . . . . . . . . . . . . . . . . . . . . . . . L-5 t

5.0 L .. C ..,S A r... .- ..< C . 3, J, ............................................ ,-i i

5.1 Responsibility ............................................. 5-1 52 organi:a ica ............................................... 5-1 53 Facility St art qualiri:stions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1 5.h Tra i n i n g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5- 3 5.5 Reviev and Audit ........................................... 5-3 551 Plan Review C ==ittee (FR ) ...................... 5-3 552 Sare:y Audi: and Feviev Cen=1::ee (SA2C) .......... 5-5 5.6 Reportable ce:urrence A::len ............................... 5-8 5.7 Sarety L1=i: violation ..................................... 5-9 5.8 Procedures ................................................. 5-9 i

I

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A=end: cat No. ' , 97, j , )d 3 l' 11 ;

TA2!.E CF C03Tl:33 (Continued)

DI?-

59 Peporting Pegairements ................................ 5-10 591 P o ut in e R e po rt s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-10 592 Rep.rt atie O cc -trer.c e s . . . . . . . . . . . . . . . . . . . . . . . . 5-12 5.9.3 3;ecisi Pe;crts ............................... 5-15 5.9.h Unique Pe;c rting Pequirement s . . . . . . . . . . . . . . . . . 5-15 5.10 P e co rd s Pe t e nt i on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-18 5 11 Pa41stion Protection Prograa .......................... 5-19 5 12 Peepir atory Prot ection Prograa . . . . . . . . . . . . . . . . . . . . . . . . 5 20 5.12.1 Peepirato ry Protection Prcgraa . . . . . . . . . . . . . . . . 5-20 5.12.2 Protection Prograa ............................ 5-20 1

S.O I3"! RIM SPE0! AL TEC'It!:AI. SPE0!T10AT:C55 . . . . . . . . . . . . . . . . . . . . . 6-1 6.1 timits on Peactor Coolant Pwp Operation . .. . .. . . . . . . . . 6-1 6.2 Use of a Spent Fuel Sht; pics Cast ............r......... 6-2 i11 uesue. u. ;,r, pt; ar. Xfj 43 1

9 DEFI!IITIO!IS .

REAC'"C3 OPEP.ATI!iG C0!iDIT!C:iS (Ccatinued)

Cold Shutdovn Conditien (Operating Mode h)

The reactor coolant Teold is less than 210 F and the reactor coolant is 0

at shutdown borca concentratics.

Refueline Shutdown Conditien (Operating Mode 5)

The reacter coolant is at refueling borca concentration and Teold is less than 2100?.

Refueling Coeratien Any operation involving the shuffling, renoval, or replacenent of nuclear fuel, CEA's, or startup sources.

The Refueling Beren Concentration A re, actor coolant boren concentration of it least 1700 pp=, which corres-pends to a shutdovn nargin of net less than 5% vith all CEA's withdravn.

Shutdown 3cron Cencentratien The boren cencentratica required to =ake the reactor suberitical by the a= cunt defined in paragraph 2.10.

?efuelin Outare or Fefuelin Shutdown A plant outage or shutdovn to perfern refueling operatic =s upon reaching the planned fuel depletien f:r a specific core.

Plant Crerating Cycle The tL=e period fren a Refueling Shutdevn tc the nex Refueling Shutdown.

Anendnent :ic. " , 2f,yl,C,- 2

1.0 3 AF"lY t, NIM AND !.MI M %C SYS Lv SI"'!"NOS 7

1.1 sersey Limite - Peeeter Cer. (Continue )

(033R). defined as the ratio of the heat flux that woult cause On ;

4 at a particular core location to the setual heat flux at that loca-ttan is indientive of the margin to On. The mini =uz value of the DNBR d'Lring steady ftste cperation, normal operational transients, and anticipated transients is 11aited to 1.30. A NB!t of 1.33 cor-responds to a 955 probattlity at a 955 confidence level that Osa vill not occur, which is cons djred an appropriate r.argin to 1N5 for all operating conditions. *'

The curves of Firaes 1-1,1-2. and 1-3 represent the loci of points of reactor thermal pcwer (either neutren flux instruments or t.T in-struments), reactor coolant sywtes pressurs and cold leg tarperature of various pump ccabination for whl=h the ABR is 1.30. ne area of safe operation is below these lines. For 3 and 2 pump operation.

the limiting condition is vcid rather than the ONBR. ne voii trac-tion li tits assure stable flow and saintenance of On3 greater than

1. 3 "he reactor core safety linits are based on radial peaks limited by

] the CIA insertion 11=its in Section 2-10 and axial shapes within the axial pcwer distribution trip limits in Firce 1 k and a total un-r attet planar radial peax of 1.62. he "Ess in Figures 1-5. 1-6 l and 1-7 are based on the ass =ption that the unrodSed integrated total radial pesX (FI) is 1.57. *his peaking factor is slightly l hisher (mcre conservEtive) than the maxinum predicted unrodded mal radial peaX during : ore life, emeluding =easurement u=cer-tainty.

Flow mallistribution effects for operation under less tta.g full re-actor coolant flow have been evaluatet via model tests.L33 The flow model data established the r.a1 distribution facters and hot channel inlet te=perature for the thermal ar.alyses that were used to esta-tlists the safe :perating envelopes presentet in Fitnes 1-1,1-2, and 1 3. These ficaes were established on the tesis that the therr.a1 margin for part-loop operatten should be equal to or greater than the thermal margin for nor al operation. The rea: tor pro.

tective system is sesigned to prevent a=y anticipate 1 cc=binatien 4

of transient coniittons for reactor coolant systes te=;erature.

4 ;ressureasithql*al less taan 1.30.

pcwer level that would result in a ON3R of Peter **ees (1) FSAR. Section 3.5.5 (2) FSAJt. Section 3.5.2 (3) TSAP. Section 1.4.6 (b) F2AR. Se:ti:n 3.5 3 s

Anessment n./. p, 43 1-2

i 2.0 LIMITING CONDITIONS FOR OPERATION I 2.2 Che=1 cal and Volu=e C0ctrol System (Continued)

a. One of the operable charging pu=ps =ay be re=cved frc= service provided two charging pu ps are oper- -

, able within 2h hours,

b. Both boric acid pumps =ay be cut of service for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

c. One concentrated boric acid tank may be out of ser-vice provided a mini ==m of 68 inches of 6-1/h per-cent to 12 percent by weight beric acid solution

,j at a te=perature of at least 20CF above saturation te=perature is centained in the operable tank and provided that the tank is restored to operable status i '

vithin 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

i'

d. Only cne flow path frc= the concentra.ted beric acid ta=ks to the reactor coolant syste= =ay be operable ,

provided that either the other flev path fro = the !

concentrated boric acid tanks to the reactor coolant syste= or the flow path frc= the SIEW tank to the charging pu=ps is restored to operable status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

e. One channel of heat tracing =ay be out of service provided it is restored to operable status within 2h hours.

1

f. One level instru=ent en each concentrated boric ,

acid tank =ay be out of service for 2h hours.

Basis The che=ical and volu=e control syste prevides ecutrol of the re- l aeter ecolant system bor:: inventory.(le This is ner= ally acec=- !

plished by using any cne of the three charging pu=ps in series with !

I cne of the two berie acid pu=ps. An alternate =ethed of boratica vill be to use the charging pu=ps directly frc= the SIEW stcrage t

> tank. A third method vill be to depressuri:e and use the safety !

- injection pu=ps. There are two sources of borated water available

for i=jection through three different paths.

i (1) The borie acid pu=ps can deliver the ec=centrated beric acid :

tank centents (6-1/h - 12 veight percent concentratics of borie

! acid) to the charging pu=ps. The tanks are located above the charging pu=ps so that the beric acid vill flev by gravity .

vithout being pu ped. ;

o l

l (2) The safety injection pu=ps can take su:ti:n fr = the SIEW ,

tanz (at least 1700 pp= bere: solution). l i

AmendmentNo.kJ' 2-13

= .

l 2.0 t,IM!""M Cac!T!"MS **P 7t'A"'!M 23 b arra ey Care Cootter ??stes 3

An11eability Applies to the operating status of the emerge =cy ore cooling system.

Objeettve

o assure operability of equipment required to remove decay heat from the core.

Speetftentions (1) M!nte pep 11r*=e-t e The reactor shall not be made critical *.utless all of the fol-

!a%ng conditions are mets

a. The S rt tank =entains not less stas 2S3.000 gallons of water with a boren concentraties of at least 1700 ;;m st a ta=perature not less thas k007. l

b. One means of te=perature indicati n (1ccal) of the 3Dtv task is operable.

c. All four safety icje: ties tasks are c;eratie and pres-surised to at least 2LO ;sig with a task liquid cf at least 116.2 inches (675) and a saxi=um level of 129.1 inches (7k1) with refuelics boros ceneet.tratics,

d. Cne level and one pressure instrument is eperable os each safety injection tant.

e. One low-pressure safety injection pump is c;erable en each bus,

f. Cne high-pressure safety injection puz; is operable en each bus.

g. 3cth shutdevn heat ex: hangers and three of four ecapo- ,

nent cooling heat exenangers are operstle,

h. Piping and valves shall be operable to ;rovide two flav paths from the 3:F'd tar.a to the reart:r coolast syste.

1. All valves, piping and interlocks associated with the above c m;ccents and required to f:.;n:ti n 1ur1:3 ac:1 dent ecediticas are operable. 30'i-291L, 293k, 297k, aal 2954 shall have power re:xsved fr:a t eir ::ter operators by locking open the circuit breakers in the Nver supply lines to the vs.lve ::x:t:r o;erators. F7I-326 shall be locket c;es, i i

1 Amea.*. ment .to. /, 32) N 2-20

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2.0 L;'4T?!iO OCE!?!?M FOP OPDA?!*'t

2. 3 Laerree v tore Oo %1nc fvstes (Continued) . >

(3) Vbenever the reactor coolant system coli leg tez,erature is below 2100F and the reactor vessel head is installed. at least two (2) EPS: pu=p control switches shall be placed in pull-stop.

Whenever the reactor coolant systas cold leg temperature is below 1100F and the reactor vessel head is installed, all three (3) EPSI pup control switches shall be placed in pall-4 stop.

In the event that co charging ;mps are operable, a single KPSI pump may to taken from pull-stop and utilised for boric acid injection to the core.

feele no normal procedure for starting the reactor is to first heat the reactor coolant to seat operating temperature by r*:.ning the reartor coolant pumps. De reacter is then aMe critical by withdrawing , ,

cf.A's ans diluting boron in the reactor coolant. Vith this mode' of start-up, the energy stcret in the reactor coolant during the a;proach to criticality is substantially equal to that 1etng ; ver '

c;eration an$ therefore all engineerei safety features and auxiliary cooling syste.ts are required to to fully c;erable. hring lov ;cver physics tests at lov temperatures, there is a negligibis amount of storet energy in the reacter coolants therefore. an s:ci$ent com- r parable in severity to the 1esign basis accident is att possible and the engineered safer 4sris systems are not required.

no 3Dr4 tank contains a sinimurs of(*l83.000 I This isgallens of usable suffleient tcronveter cen-containing at lesst 1700 ;;a boren. l i centration to provide a shuttevn margin of $5. including allovenees for un:ertainties, with " ccatrol rods withdrawn an$ a new core at a te=;erature of 6007 S e limits for the safety injection tant ;ress2re and volume assure <

the required amount of vster injection Suring an accident ar.1 are base $ on values used fer the accident analyses. Oe sini=us 116.2 inch level :orresponts to a valune c* 625 ft3 anA the maxiz:.:n 128.1 inch level correspenas to a volune of $93 5 ft3 Prior to the time the reactor is brought critical, the valving of the safety injecti:n syntes sust be che:tet for ccrrect align??.ent and apprcpriate valves lo:ket. Since the system is '4 set for shut ,

down ecoling, the valving vill be changet ana nunt te ;rcperly saignet

prior to start-up of the reacter.

"he operatie status of the various syntess and compene to is to be ietonstratei by perio11: tests. A large fra:tien Of these tests vill to ;erferzet vt11s the restter is c;erating in the pcver range.

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2. 0 lit *I*TN1 C7f0!*!**M MP CPE*A?!*N 2.0 Pef 3eline 0-erattu s (Continued)

(7) Direct comrunication between ;erson=el in the ccatrol reca and ;

at the refueling nacaine shall to available whenever changes <

in core geonetry are takteg place.

(8) ha irradiated fuel is being handled in the auxiliary tuli-ing, the exhaust vect11ation frem the spent fuel pool area vill be diverted through the charcoal filter. ;

()) Prior to initial core loading and prior to refueli:g c;erations, e complete check out, including a load test, shall be conducted .

on fuel handling cranes that vill te required during the re- [

fueling operation to handle spe=t fuel assentlies.

(10) A ninimum of 23 feet of water above the top of the core shall be maintained whenever irradiated fuel is being handled.

If any of the above c nditions are not met, all refueling operations snall cease i= mediately, verk shall te initiated to satisfy the re-guired conditians, and no operaticas that =ay char.ge the reac*ivity of the c:re shall be made. However, rePaeling cperations may ces-nonce and centinue with less than 5 centaic=ent at=es;here and plant ve:tilatien auct radiation moniters provided that gross, particulate and iodise monitors are =cr.itoring the stack efflue=t. These three plant ventilation duct radiation =enit=rs vill initiate elesure of i the containment pressare relief, air sa=ple and purge systes valves [

and shall e= ploy a one-out-of-three logic for the initiation of VIAS.

Irrailstoi fael movement shall not be initiated tefore the reacter core has decayed for a mini =um of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months if the reacter has been operates at power levels in excess of 21 rated power.

Basis

he equipment and general procedures to to utilizei during refueling cperations are discussed in the ySAR. Detailed instructicas, the above specifications, asi the design of the Pael hesiling equipment incorporating tuilt-in interlocks a 4 safety features provide assura:ce that no isc11ent couli occur during the refueli:q c;e aito:s that would result in a hazard to public health and safety.11J Whe ever changes are act teing =ade in ecre sec=etry One thz mecitor is suf- .. .

fielent. This per=its salutenance of the 1:str=entatics. Continzus

=cnitoring of radiation levels and neutron flux ;revides i=neilste i=.tiestien of an unsafe condition. The shuttovu cooling pu - p is ,

used to saintain a uniform beron concentratica.

he shut svu margin as indicatet vill'heep the core suberitical even if all O!.A's were withdram from the core. Durisc rePaeli=g c; era-tiens. the rea:ter refaeling envity is filled with a;;roximately 250,000 galices ef tcrated veter. "he bcron ::n=estratics of this water (at least 1700 ;;m beren) is suffi:1ent to maistain the reactor l suberitical by :ssre than 51,1::luiing allavas:e far uncertaisties, is the :214 c:ndition with all rods withirawn.lM Periotic : hecks of re.

Peeling vster bcren ee: centra:Lon ensare the pre;er shu*iews =argis.

Cormnicatien ret 2irements allow the ecstrel recs operator to ist:rs the refueling rachine cperator of a y imposting unsafe ::=iitica de-tected frxs the main c:strel beard inilcators tari:g fuel =cvement.

Amenizent3s.f,;/,25,O 2 35 t

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I i 2.0 !.!'C*!M C0!?!0*ts M CP!M*!CN 2.10 Peactor Cere (Continue $)

1 2.10.2 Pesetivity *m*rel 'vstems a-4 Sre Mvetes Partneters *,1r.its y Arlicabtlity -

Applies to operation of control element asser.nlies and actitering of selectet core parameters vnenever the reactor is in coli cr hot shatsown, hot standby, or power operation conditions. .

cM eettve To ensure (1) adequate shutdovn margin following a reactor trip, p (2) the !CO is within the limits of the safety esslysis, and (3) t I

control element assembly operatica is within the limits of the -

setpoint and safety analysis.

Sreetfteatten I

' (1) Phutdown Margin with Teng >210 07

'tenever

- the reacter is in hot shutdevn, het stanity er power operatica coniiticas, the shutdown zargin shall be 13 75 t.k/k. With the shutdown nargin <3.7% 'k/k. initiate i

and continas boratics until the requirei shutdsvn margin ?

t is achieved.

e (2) Shutdown ftargin with Teoit 12100F ,

1 '

1 Whenever the reactor is in cold shutievn conditions, the

) shutdown margin shall to 21 .C5 Ak/k. Vith the shutievn margin <2.C5 t.k/k, initiate and continue boratica until the required shutdown margin is achieved.

(3) Merater Te- er,ature Nf ftetent "to moderator tes;erature coefficient (PCO) stall be r l

s. 14ss positive than +0.2 x 10 k as/'T including uncer- f tainties for pover levels at er above SCS of rated pcver.

4

b. 14ss pcsitive than +0.5 x 10 4 as/'T including uncer-rainties for power levels below 301 of rated pcver.

c. More positive than -2.3 x 10 k to/*F including uncer- I tainties at rated power.

With the moierator tentgrature coefficient ccafic wd outside l

any one of the above limits. change reactivity centrol para-e,eters to bring the extrapolatei .5C0 value within the above limita vitain 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months or be in at least hot shutievn within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

i Ananiment :fo. /. 32j 2 50 I F 1

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L 2.0 1.191t!M COC!TTONU MP CeEPATICM 2.10 Peneter core (Continues) 2.10.2 Penettvity cen*rei syste-s a-1 cere PWsies ?srs:-eters '.i its (Continued)

1. The total available shutdown =argin =ay be reduced to 25 f.k/k during the measurement of the shuttavn CIA group rea:tivities, or t t

2. The total available shutdown margin may be redaced to the worth of the worst stuck CEA's I during the measurement of the stuck CEA re-4 activity. s

( (ii) If the shutdown margin specified in part (1) abcve [

is not availatie immeilately, initiate and conti-nue beration until the requirements of 2.10.2(1) {

v are met. L I

I (iii) "he shutdeva me.rgin speciflei in part (1) above shall be verified every 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months shi"*

?

. Moderater Te=; erat

e Oceffi: lent i (i) The moderator te=perature coefficien* (!CC) re- !

suspenies during quirements physics tests of at 2.10.2(3) less than may 10 be,1

' of rated ;over.

i (ii) If power exceeds 10-11 of rated pover, either

1. Redu:e p:ver to less than 10-15 of rated power within 15 minutes, or

2. Se in het shutdown in 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

?nste i

l Shuttev. Farrin ?

A sufficient shuti:vn mm.rgin ensures that (1) the reactor can to

=ade suberitical from all c;erating c:nditices. (2) the reactivity trustents associated with postulated accident ::ciiti:cs are :en.

trollatie within acceptable lir.its, ani (3) the rea: tor v11' be maintained sufficiently saberitical to pre.lude inadvertent criti-cality in the shutacvn conditi:n.

Shutdown margin reTaire .ents vary throughout : Ore life as a functi:n of fuel iepletion. RC3 bores ccacentratien, and R 3 Taft. The =ost restriettve eeniition occurs at E01. vith Tgyg at no 1:ai operating temperature, and is associated with a ;ostulated stens 1.ine break a:cident sad resulting un:catrollei R03 :ooldovu. In the analysis

' Of this a:cident, a sinim s shuti:vn =argin of 3.7% ik/k is ini-tialir sietuate to centrol the resetivity transient. A::crdingly .

l

i. ,

Amou:nent ilo. 32. 't A 2 5cf I

i i

l l !

2.0 LI:tITir!G CO::DITIO*ts Ton OPrRATIO t 2.10 Renetor Dire (Continued) 2.10.2 Henetivity Control Sycters and Cora Phynten Parn.-atere Limits (Continued) the shutdovn marcin requirement is based upon this limitinC condition and is consistent with FSAR safety analysis assumptions. With Teold

<2100F, the reactivity transients resultinC from any Postulated ae-cident are minimal and a 2% dk/k shutdown marcin provides adequate l protection.

Control Element Ascenblies

The specifications of this section ensure that (1) acceptable power distribution limits are maintained, (2) tho mininun shutdown margin is maintained, and (3) the potential effects of CEA misalignments are limited to acceptable levels.

The statements which permit limited variations from the basic re-quirenents are ac::npanied by additi:nal restri:tions which ensure that the or161nal design criteria are met.

The cpecifications applicable to one or nere CIA's that are deter-mined to be untrippable or stuck, and to one or more nisaligned CIA's that cannot be restored to within 12 inches of any other CIA in their group, require a pr: pt shutdown of the reactor since any of these conditions may be indicative of a possible loss of ne:hanical func-tional capability of the CIA system and in the avent of an untrip-pable CEA, the loss of shutdown margin.

For snall =1salignments ( <13 inches absolute) of the CIA's , there

is 1) a small degradation in the peaking factors rolstive to those assumed in generating LCO's and LOOS setpoints for 0: !23 and linear heat rate, 2) a small effret en the time dependent long tern power distributions relative to those used in genarating LOO's and LSSS setpoints for D:*B3 and linear heat rate, 3) a small effe:t on the available shutdown margin, and b) a s=all effe:t on the ejected CEA vorth used in the safety analysis. Therefore, the action statenent associatel with the small misalignrent of a CCA permits a one hour tine interval during which atte= pts may be made to re-store the CCA to vithin its alignment requirerents pri;r to ini-t1ating a reduction in power. The One hour tine is sufficient to

1) identify c2uses of a misaligned CIA, 2) take appropriate cor-rective acticn to realiCn the CEA's, and 3) =inimite the effects

, of xenon redistribution.

Overpower mar-in is provided to prctect the cere in the event of a lar,e nisaligncent (213 inches) cf a CCA. !!ovever , this mis-aliCnnent vould cause distortien of the core power distribution.

The ren tor protective system would not detect the derradation in radial peaking facters and since variations in oth-r sy ten p0ra-meters (e.c. , prescure :nd coolant. te perature) .ay not 5 sufficien Amendnent ;o. .,,J, 32 N 2-71

9 2.0 l!w! TINS carf!?!Sf3 F0p tPreA* TOM 2.10 Peseter C:re lContinuet) 2.10.i. pg er : stri Mtt w '!-!ts (Ccatinued)

(b) If while operating ander the provisions of part (a),the plant ecaputer incere detector alar =s become inoperable,

.cperation may be continued without reducing power pro.

vided each of the following conditicas is satisfied (1) A core power distribution was obtained utilisir.g incere detectors within 7 days prict to the in-core detector alars outage and the measured peak linear heat rate was no greater than 905 of the value allowed by (1) above.

(ii) The Axial Shape Indez as asasured by escore de-tsetors remains within 105 of the value obtaiced at the time of the last measured incore power distribution.

(iii) Power is not increased nor has it been increased since the time of the last incere power distribu-tion.

(c) W.en the linear heat rate is continuously scnitored by the excore detectors, withdraw the full length CIA's beyond the long ters insertien limits of Specification 2.10.2.7. If the linear heat rate is ex:eeding its limits as determined by the Axial Chase indez, T:, betes outside the limits of Figure 2-6, vnere 100 percent of the allowable power represents the maximua power allowed by the following erpressient L

1*.T * *'

v ,e,e

1. !. is the naziz m allovable limeat heat rate as deter =1ced frca y! care 2 5 and is based on the core averade burnup at the time of the latest in:cre power =ap.

2. M is the maxima:n allevable power level f r the trieting Pes: tor Coolant P=p co= tina-tion.

3 N is the maximun allowable fraction of rates ther'aal power as detersined ty the TgyI 11=1t eurve of Ficare 2-) vnen monitoring ty ex-ecte ietectors. 3 = 1 vsen menitcrits kv/ft using incore setectors.

(1) Postcre the rea:ter power and Axial $hape Intez.

Y ! , to within the limits of Fisare 2-6 vithin 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or (11) Se in at least hot stanity vithin the ner. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Am ,dient :to. jf, gd, 32 43 2-57

2.0 . I,Av. . ,i..

G Cr.i3o. v. ..s.m.a . 0,. 0 e. rn. .A . r0,.i..

2.10 Resetor Core (0:nti=ued) 2.in.k Power Distributi:n Lizits Applicability Applies to power operation ecnditions.

Objective To ensure that peak linear heat rates, :::i3 =argins , and radial peak-ing factors are raintained within acceptable li=its during power operatien.

Specifiestien, (1) linear Rest ?ste The linesr hest rste shall not exceed the li=its shev:. On Figure 2-5 when the following fset;,rs are appropristely in-c ., . . a.,. 4 ..

1. Flux peaking aug=entatien facters are sh0vn in Fig.re 2-8,

2. A =essureze: t-calculational un:ertainty facter

.* 1.0en,

u. ~

3 An engineering uncertainty f a:ter of 1.03,

!. . A linear heat rste uncertainty factor of 1.002 l due to axial fuel densification and therral ex-pa.sicn, and

5. A pcver =easurement uncertainty fa:ter of :. 02.

"he reasurement-:alculati:nal uncertainty in (1).2. sheve shall be in:ressed, as ne:essary, purzusnt t: Spe:ifica-ti:n 2.10.3(5)(a).

(s) 7nen the linear heat rste is centinu:usly renitored by the ine:re dete::crs , and the lines heat rste is ex-ceedin.; i* s li=its as indicated by four er =cre valii

incident ine:re dete:::r slar=s , either:

i t .4 .... s.... .. .. .. u..

. .,s..

. es.. s..

.. . .a..s, 4 ., s

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s.

(44 ) s, 4

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. . ..w.. . .x;

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6 hcurs.

Ameninent ic. [. % , "., 32 'd 2-56

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

i i

2.0 htvr f m marninw rre ce*eAMvs

2. m s< w ,.e core (runtinen) e

. i o. 4 e...- m ur s t...t hn TDite (f~ntinued )

(2) Total fntecretai Pait a: eeskier Paeter

he calculated value of F3 7 tefined by TpI = F3 (1+T ) shall be limited to 11.57. FA is determined free a power 3 istribu.

l tion map with no part length CIA's inserted and with all fdl length CIA's at or above the Long Ters Steady State Insertion Limit for the existing Reactor coolant Puerp combination. '"be asirathal tilt. Tg, is the measured value of Tg at the time Fn is determined.

With T37 *1.57 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months coerply with any oce of the foi- l

loving

6 -

(a) Peduce power to bring ;over and T3 7 within the limits of Picure 2-), wi',hdraw the full length CIA's to or to-yeni the Ler.g Ters Steety State Insertion Limits of 3;eci.

fication 2.10.2(7), and fally withirow the PLOZA's,or

! (b) Setace the allove t ;over of the axisi pver distributien J

333 monitoring erve of Tigare 2-7 by the esount of i 30x QF R7 /1.571-6 x 100 percent of rated pcVer, and l redace the values of 6563. 5122. and kP17 in the Pvar eyJations b r$/1 57)-1 of )';*cificatica x 100. withdraw 1.3(4)the by the amountCIA's ful ler.c.h of 20 to m l

t or beyond the ste'sdy state insertions limits of Speci-fiestion 2.10.2(7) and Tally withdraw the P 4ZA's. or (c) 3e in at least het stanity.

(3) ?7tal Plenar Weital Pea.Mf re ?stter Se calculated value of Fxy7 tefined as FryT

Fay (1*Tg ) shall to 11 mite 4 to *1.62. Fxy shall be seterminet frem a power distribution map with no part length CIA inserted and with l

all full length CIA's at or above the 1er.g *erm Steady State Insertion Limit for the existics Peactor Coolant Fusp com.

binatica. *his ieterzination shall be limited to enre planes between 15% and $55 of rull core heignt inclusive and shall excluse regions influenced by gril effects. The seizuthal tilt, T . gis the measured value of Tg at the time Try is Setermines.

With Tay7 *1.62 within 6 hears ecaply with any cce of the foi- l loving:

(a) Re tuce pcver to brise power and T I to withic the limits of figure 2-9. ana withirav the f31 lench CIA's to or besd the an, :er, stes 7 etate Inser.1:n u its of

, $;ecificatien 2.10.2!7) and fully withdraw the PLCIA's.

I sr d

Amentrent 23. 32 O i-57a i

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

t k

2.0 2.10 LINGT!NG CONDITICNS FOR OPERATION R? actor Core (Continued)

' ~

2.10.k Power Distributien Limits (Centinued) _

NOTE:

When power is reduced to co= ply with (a) above and' linear heat rate is being ronitored using excore , !

detectors, Figure 2-6 =ust be adjusted in accordance with Specification 2.10.k.(1) (c). ,

(b)

Reduce the allowed power level of the axial power dis-tribution linear heat rate =cnitoring (L=its of Figure 2-6 by the anount 2.22 x 8F3g/T/1.62)-L x 100 percent

'of rated power, and reduce the positive and negative l axial

. 01 xshape index trip limits of Figure 1 k by the a=ount

[(F xyT /1.62)-() x 100 ASI units, withdraw the full length CEA's to or beyond the Long Ter: Steady State l i Insertionthe withdraw Limits of Specification PLCEA's, or 2.10.2(7) and fully (c) 3e in at least hot standby, (k)

A.zimuthal Power Tilt (To)

When operating above 70% of rated power, the azi=uthal power tilt (T )qshall not exceed 0.03.

(a)

With the indicated azi=uthal pcver tilt determined to be >0.03 but <0.10, correct the power tilt vdthin two hours or deter =ine within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and at least once radialper peaking subsequent factor,8Fhours, 7 that the total integrated Specificatica 2.10.k(2) 3 , is within the li=it of T and that the total planar radial peaking factor Fxy , is within the limit of 2.10.k(3),

8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months of confirmingg Tor reduce power to less than 70% of rated

>0.03.

(b)

With the power indicated pcver tilt deter =ined to be >.10, egeratic: -

=ay proceed up to 2 hcurs provided Fa' andteFx or .:- do net exceed the pcVer li=its of Figure 2-9, at least hot standby within 6 hcurs. Subsequent cperatica for the purpose of =casure=ent to identify the cause of the tilt is allovable provided: "

(1)

The pcver level is restricted to 20% of the =axi-

=un allevable ther=al pover level fer the exist-ing reacter coolant pu=p cc=bination, and (ii) !

"'he axial power distributien s=d ther=al =argin trip setpoints and the peripheral axial shape

' index limits of Figure 2-6 and 2-7 are adj'ated in accordance with Specificaticas 2.10.k(2) and 2.10.k(3). ,

A=end=en No. 36 7 2-575 "

I 1.

! 2.0 l'X.TL'iG CO*TDITICNS ?CR C?r*r'!C'T 5 2.10 Reactor Core (Continued) ' '

2.10.k Power Distributica li=its (Continued)

(5) D*T3R Marri Durin: ?cver C e stien Abeve IST cf Rated Fever The folleving ::iB related pars =eters shall be =aintained (a) a within the 11=its shev: on Table 2-6. .

(i) Cold Leg Te=perature 1

l (ii) Pressurizer Pressure i

(iii) Placter Coolant Flov

!, (iv) Axial Shape Index, Y- -

j (b) Vith a 7 of tPs above para:eters ex:eeding its limit, t

restere the pars =eter to withis its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce pcver to less ths: 155 cf rated ;over 4

vithin the next 8 hcurs.

Basis linear Hest Rate

'"he li=itatics en linear hest rate ensures that in the event of a MCA, the pesi te=perature of the fuel claddi:g vill t exceed 22000F.

i i

Iither of the two : re power distributic: =ositoring syste=s, the Ixecre Oetecter Monitoring Syste=, or the 2:cre Oetector Mcniter-ing Syste=, provide adequate :::it: ring of the core pcver distri-3 tutics and are capable of veri.^/ing that the li:ese hest rate d:es

=ct exceed its 11=i* s. The Ix: Ore Detec cr M::itering Systes per-f:r:s this fu : tics by ecutinuously =enitorbg the axial shape in-dex vith the cpersble quadrs : sy==e:ri: execre neutres flux de-

, tectors and verifyi g that the axial shape index is =tistained with-

!' in *he slievable limits Of Fisce 2-6 as adjusted by Spe:1ft:stic:

2.10.L. (1). (c) f:r the all:ved linear hest rate of Figure 2-5, 20

! Pu.-p cc fitrstics, and Fxy! Of Firre 2-9 *: :::Ju=ctics vith the use f the ex:cre =ccitering syste= and in establishi:g the l J

i sxial shape index li=its, the f:11:vi:g assu=;t10:s are =ade: (1) the CIA inser.ic limits of Specifiestion 2.10.2.(6) sad 10 g ter=

insertien limits of Specifi:stics 2.10.2(7) tre satisfied. (2) the

! flux peaking sus =entatics fa:: Ors are ss shev: 1: 71rr e 2-3, (3) j the a:isuthal ;:ver tilt restricticus of Specifiestics 2.10.k.(k)

I tre satisfied, s:v. (L) the ::tal plazar rsdial ;eali:g fa:ter does l ::: ex:eed .he ' *-d :s of Specifi:stics 2.10.L.(3) .

I i

Os *::cre De*.e:*.cr M::itoring 3ys.e: pr

rides a direct = essure of the pesking fse .crs sad the ala =s whi:h have bee: es atiished f:r the indivihal 1:ccre dete cr seg=ents ensu e ths: the ;esh linear test estes vill be ::::1 :usly =aintai ed vi-his the sil:vab'.e li=its Of Firre 2-5. "he setpci :s f:r these slar=s include al-1:vsa:es, set is the :::servstive directi::s, f:r the fac Ors listed

.. ,.. ......). .

A end:ent 3c. 22,17 2-57:

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J J

i

I i

2.0 LDf!TI'TG CONDITIC'IS FOR CPERA"' ION i 2.lk Engineered Safety Features Syste= Initiation Instrumentation Settings i (Continued) 1 The set points for the isolation function have been selected I i to linit radioactivity concentrations at the boundary of the

, restricted area to approximately 0.25 of 10 CFR 20 li=its, as-1 su=ing existence of annual average =eteorology.

Each channel is supplied from a separate instru=ent a.c. bus j and each auxiliary relay requipes power to operate. On failure l l of a single a.c. supply, the A and B =atrices vill assu=e a '

one-out-of-two logic.

4 (k) Lov Stea= cenerator Pressure A signal is provided upon sensing a low pressure in a stea=

generator to close the main stea= isolation valves in order to =inimize the temperature reduction in the reactor coolant syste= vith resultant loss of water level and possible addi-4 tion of reactivity. The setting of 500 psia includes a +22 _

psi 'vertainty and was the setting used in the safety analy-sis. I (5) SIEW Tank Lev Level

Level svitches are provided en the SIRW tank to actuate the

valves in the safety i=Jection pu: ..p suction lines in such a

=anner so as to svitch the water supply frc= the SIRW tank to the containment su=p for a recirculation = ode of cperation after a period of approxi=ately 2h minutes following a safety injection signal. The svitchover point of 16 inches above tank bottc= is set to prevent the pu ps '-- -"- dng dry during the 10 seconds required to stroke the valves and to hold in reserve l borated approximately vater. Se FSAR 28,000 gallons loss of of at coolant least 1700 accident pp= (k) analysis assu=ed the recirculaticn started when the -*-4---

usable volu=e of 233,000 ss11ces had been pu= ped frc= the ta.%.

I i

Amend =ent No. f, 3d 43 2-62

m 2.0 LIMITI!!G C0!!DITIC!!S 703 OPERATIO!!

2.15_ Inceru entation anc control svetens (Continued)

. assumes a tripped conditien (except high raye-of-change of power, high power level and high pressuriser pressure),tl) which results in a one-out-of-three channel logic. If in the 2 of k logic syste= of the reactor protective system one channel is bypassed and a second channel =anually placed in a tripped condition, the resulting legic is 1 of 2. At rated power, the =ini=um operable high-power level cha=nels is 3 in order to provide adequate pcver tilt detection. If only 2 channels are operable, the reactor pcver level is reduced to TC% rated power which protects the' reactor frc= possibly exceeding design peaking facters due to undetected flux tilts and frc= exceeding drcpped CM peaking facters.

All engineered safety features are initiated by 2-out-of L Icgic =atrices except contain=ent high radiatica which operates en a 1-out-of-5 basis. l The engineered safety features systes provides a 2 of h Icgic en the signals used to actuate the equipnent connected to each of the two e=crgency diesel generator units.

The rod block syste= aute=atically inhibits all CM =ction in the event a Limiting Ccndition for Cperation (LCO) on CH insertion, CM deviation CIA overlap or CM sequencing is apprcached. The installation of the red bicek systes ensures that no single fa""-

n the control ele =ent drive control syste: (other than'a dropped C M) can cause the CEA's to nove such that the CIA insertion, deviation, sequencing er overlap li=its are exceeded. Acceriingly, with the red bicek syste= installed, only the dropped CIA event is considered an ACO and factored into the derivation of " f * '** g Safety Syste= Settings and Li=iting Cenditiens for Operation.

With the red block '"~+'- cut-of-service several additional CIA deviation events nust be censidered as ACO's. Analysis of these incidents indicates that the single CM vithdraval incident is the

=ost liniting of these events. An analysis of the at-pcver single CEA withdraval incident was perfer=ed for Fort Calhoun for varicus initial Grcup k insertions, and it has been ecceluded that the Lisiting Cceditions for Cperatica (LCO ) and Li=iting Safety Syste:

. Settings (L335) are valid fer a Group h insertien of less than er i equal to 15"..

I

! References f (1) FSAR, Section 7.2 7 1 A=endment ::c. p'jd, , 22 33- 2-66 l

l l l t l

1 --

wv

8 h

t k.0 DESIG'i FENI'JRES_

4.k Fuel Storare ,

k. h.1 !!ev Fuel Storace s

l The new unirradiated fuel bundles vill normally be stored in the dry new fuel storage rack with an effective =ultiplication factor of less than 0 9 The open grating floc,r below the rack and the '

i covers above the rachs, along with generous provision for drain-age, precludes flooding of the new fuel storage rack. l Ilev fuel =ay also be stored.in shipping containers or in the spent [

fuel pool racks which have a =aximum effective multiplication face .

l tor of 0.95 vith Fort Calhoun ?ype C fuel and unberated water. ,

I '

t The new fuel storage racks are designed as a Class I structure.

l r

h.h.2 Spent Fuel Storare Irradiated fuel bundles vill be stored prior to off-site shi;=en*

in the stainless steel lined spent fuel pool. The spent fuel pool is normally filled with berated water with a concentration of at j least 1700 pp=.

l i The spent fuel racks are designed to =aintain fuel in a geomet:7 [ '

l vhich ensures an effective =ultiplication factor of 0 95 or less I

vith new fuel containing not = ore than h0.1 gra=s of U-235 per -

axial centimeter of fuel assembly when the pcci is filled with un-borated water.

The spent fuel racks are designed as a Class I structure. [

3 : Tor: ally the spent fuel pool cooling syste: vill =aintain the bulk I l vater te=;erature of the pocl belev 12007. Under other conditicas '

4 of fuel discharge, the fuel peci vater te=perature is =aintained ,

belev 140C7. i

I i ,

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Amendment 2: o . 'd. O , ,

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- 0.4 - 0.3 - 0.2 - 0.1 0 0.1 02 0.3 0.4 AXIAL SHAPE INDEX, Y 7 r FORT CALHOUN i.!MITING CON 0iTION FOR CPE R ATION FOR FGURE TECHNICAL EXCOR E MONITORING OF LINEAR HEAT RATES 26 SPECIFICATIONS UP TO A MAXIMUM OF I4.7 KW/FT.

Amend ent n. J. *

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