ML20129C535

Rev 0 to Incore Instrumentation Operability Requirements

ML20129C535
Person / Time
Site:	Fort Calhoun
Issue date:	03/01/1996
From:	Luikens R, Stafford C OMAHA PUBLIC POWER DISTRICT
To:
Shared Package
ML20129C526	List: LIC-96-0157, Provides Change to Util Procedure OI-NI-2 W/Supporting 10CFR50.59 Evaluation Which Allowed Continued Operation W/Less than 75% But Greater than 28% Operable Incore Detector Strings ML20129C535
References
OI-NI-2, NUDOCS 9610240052
Download: ML20129C535 (34)

Text

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FORT CALHOUN STATION FC-154 GENERAL FORM R15 NUCLEAR SAFETY EVALUATION SEE N00-QP-3 FOR INSTRUCTIONS p/**

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,, ID No. 4 7773 SECTION A Page / of 8 (from 9.1) 10 CFR 50.59 Applicability Screening 9.1 Activity Identification Procedure Change No. 4~7 '13 3 affecting Procedure dI- NE .2 Modification Request No. Design [ ] Installation [ ] Testing ( )

l Temporary Modification No. Engineering Change Notice No.

Other Document Tit 1e: TNCME TM tream eurA ro W MffA A ft / r y REQutyEsmEnis

} Nuclear Safety Evaluation Conclusion

[] This activity is not a 10 CFR 50.59 activity, because it:

I e Does not change the facility as described in the USAR.

e Does not change procedures as described in the USAR.

• Does not involve conductino tests or experiments not described in the USAR.

• Does not affect Nuclear Safety in a way not previously evaluated in the USAR.

[X] This activity is being done pursuant to 10 CFR 50.59. This activity jitjilu be approved by the PRC. ,

o This safety evaluation must be reviewed by SARC; ref Tech. Spec. 5.5.2.7.

• This activity must be reported in the annual report; ref 10 CFR 50.59, Item b, Paragraph 2.

[] This activity involves an Unreviewed Safety Question. The activity must be canceled, or revised and re-evaluated, or NRC authorization is required prior to implementation; ref 10 CFR 50.59, Item c.

~

We hereby certify that this Nuclear Safety Evaluation is complete and accurate to the best of our knowledge.

Prepared by M M cu s 5'. G visa Date 2h9/H. Time 14/f p@' Print Name IEi 4* M' -

'72~17

(/ Signature 3 Extensioft/% lSL5 Revicwed by 4 O r2t li N M I di d ate J- M 9d Time /6/d Print Name Mh77)s#-/unkt. Extension //dY

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FORT CALHOUN. STATION FC-154 GENERAL FORM ,

R15 I

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NUCLEAR SAFETY EVALVATIDE SEE NOD-QP-3 FOR INSTRUCTIONS *I'/w 10 No. #h'l 33 SECTION A Page .Z of (from9.1) 10 CFR 50.59 Applicability Screening 9.2 What (specifically) is being done?

tv.sneo m b de.o Revi.sdag fAe operrbHnY y Refonte.~seN.s of the t^'coee

.,2, /g allow cou lo"ved operedsom wow Aess 7% 7sp; (tc1) 49 05 - Nr bv f greal-er 149-&s% op.rs.ble T ^'Coe E insraem e,v kbw .s /roxy s ,

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9.3 Why is this being done (briefly)?

7e Jc.sh yf the re/Ayalool> of TAe oper+blbly regvere ~en b forTAr EC C s'pjen uc! allow a he9 'ker Nurke.~ o$ ICIs k 6t gMpyr5bfl GH) $he ll u

frodde the y ab,'l.l fi y debobl nous lk YAa feaerdes/rikfrh /5lls reg v}<d by % Tecuasu Spe4Wiesa ud descroird la USAA selsou.

7. 5. 4. 3, 9.4 Does the activity involve a change to the Technical Specifications other than the Basis Sections?

[x] NO -

• This activity meets the requirements of current Technical Specifications. The following sections were reviewed:

3 /O. / , D. i b . .),.?.10. ( S.2 .

• Continue with 9.5 4

[ ] YES -

• Technical Specification Section must be revised prior to performing this activity, o Exit this procedure and continue with N0D-QP-7.

A-2 (50-G-30,N00-QP-3)

FORT CALHOUN, STATION GENERAL FORM FC-154 ,

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NUCLEAR SAFETY EVALUATION SEE N00-QP-3 FOR INSTRUCTIONS N ID No. 411 M IS i SECTION A Page .,3_ of t4~ '

(from9.1) -

10 CFR 50.59 Applicability Screening L 9.5 Does the activity involve a change in the facility?

(X) NO - Go to 9.6

[ ] YES -

Is this aspect of the facility described in the USAR?

List USAR Sections / Figures reviewed: l

[ ] NO -

Go to 3.5

[]YES - list USAR Sections / Figures

• Does the USAR description require any changes or revis;cr.s ::ue to this activity?

( ) NO -

continue with 9.6

( l 'fES -

• 10 CFR 50.59 applies to this activity

• Section 8 of the Nuclear Safety Evaluation must also be completed.

• continue with 9.6 9.6 Does the activity involve changes to procedures?

[] NO -

Go to 9.7

[>d YES Are related procedures sincluding definitions or descriptions of activities or contro;s over functions) outlined, summarized, completely described, or implied in the USAR?

List USAR Sections / Figures reviewed: 7. / , 7. 5 , / u.14'

[ ] NO -

Go to 9.7

[)d YES - list USAR Sections / Figures 7. S. '/. 3 e Does the USAR description require any chan9es or revisions due to this activity?

[ ] NO - Continue with 9.7

(>0 YES -

• 10 CFR 50.59 applies to this activity

• Section 8 of the Nuclear Safety Evaluation must also be completed 1

• Continue with 9.7 l A-3 l

l (S0-G-30,N00-QP-3)

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_ FORT CALHOUN' STATION FC-154 GENERAL FORM R15 NUCLEAR SAFETY EVALUATION 7[hw SEE N00-QP-3 FOR INSTRUCTIONS I >-

ID No. @% SECTION A Page Y of &

(from 9.1) 10 CFR 50.59 Applicability Screening 9.7 Does the activity involve tests or experiments?

[y] NO - Go to 9.8 ,

[ ] YES - Is the test / experiment one which has been previously anticipated in the USAR?

{ ] YES

• list USAR Sections e Go to 9.8

[ ] NO - (i.e., it is not described in the USAR; includin one-of-a-kind tests or new system configurations Could this test / experiment degrade the margins of safety during normal operations or anticipated transients, or could it degrade the adequacy of structures, systems or components to prevent accidents or mitigate accident conditions?

[ ] NO -

- Continue with 9.8

[ ] YES -

• 10 CFR 50.59 applies to this activity

• Section B of the Nuclear Safety Evaluation must also be completed.

• Continue with 9.8 A-4 (50-G-30,N00-QP-3)

.i FORT CALHOUN STATIUN FC-154 GENERAL FORM R15 NUCLEAR SAFETY EVALUATION SEE N00-QP-3 FOR INSTRUCTIONS

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ID No. 4'7 73'3 ifCTION A Page 6~ of M (from 9.1) 10 CFR 50.59 Applicability Screening ,

i 9.8 Could the activity adversely affect nuclear safety?(Consider System Interactions)

D() NO - Explain do ' cctv u e-t- / <> /4. o/

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- Go to Nuclear Safety Evaluation Conclusion or continue with Section B of the Nuclear Safety Evaluation, if required.

[ ] YES - How l

l Has this effect been previously evaluated in the USAR?

[ ] YES -

discussed in USAR Section

- Continue with Nuclear Safety Evaluation conclusion

[ ] NO - e 10 CFR 50.59 applies to this activity j e Continue with Section B of the Nuclear Safety Evaluation A-5 (50-G-30,N00-QP-3)

e FORT CALHOdN STATION FC-154 GENERAL FORM R15 NUCLEAR SAFETY EVALUATION SEE NOD-QP-3 FOR INSTRUCTIONS ID No. 47733 Page 6 of.15 (from 9.1)

ATTACHMENT SHEET 9.8 at any of the detector levels. Additionally. USAR 7.5.4.3 states that the azimuthal tilt calculation is valid for a minimum of two ooerable ICI strinas ner cuadrant.

For Cycle 6. ABB-CE analyzed a similar situation of failures. An exolicit analyses of current and Droiected detector failure natterns wasl oerformed (CEN-150(O)-P. " Analysis of CECOR Power Peakina Uncertainties for Fort Calhoun Unit I Cycle 6", February 1981). This analysis was formally submitted to the NRC in the form of an Interim TS. and the NRC aranted the TS chances which allowed ooeration with a reduced ,

compliment of incore detectors (Amendment 55). The chances crocosed for Cycle 16 are the same as those reauested in Cycle 6 except this reauest does not recuire a TS chance and the minimum allowed operable strinas is 28%. Since there have been no chances made to the desian of the ICI System since Cycle 6. the analysis utilized in Cycle 6 remains valid.

This is based uoon the use of the same methodoloov. the same ICI System conficuration, and the same anolication of peakina factor measurement uncertainties. The current Deakina factor uncertainties are slichtly '

different in value that the peakina factor uncertainties used in cycle 6 due to current use of the latest revised CECOR tooncal methodoloav (CENPD-153-P. Rev. 1-P-A) which was implemented after Cycle 6. The ceakina_ factor uncertainties derived from the use of the current prediction codes ( CASMO / SIMULATE) are bounded by both the Cycle 6 and Cycle 16 ocakina factor uncertainties derived from CENPD-153-P. The CENPD-153-P osakina factor uncertainties are currently apolied in the FCS reload methodoloav. Therefore, the orincioal difference between the Cycle k_and the current Cycle 16 reauest is usina the 10 CFR 50.59 4 process now versus the TS amendment in Cycle 6. As a result, the aonlication of an additional uncertainty of 1% to the ocakina factors and PLHR is conservative since the Cycle 6 analysis determined that the increase in uncertainty to 80h failed was less than 1%. ,

Apolvina the 1% increase will result in the followina peakina factor I' total uncertainties:

5.3% (*) + 1% = 6.3% for Fn

• Prom CENPD-153-P. Rev. 1-P-A.

5.0% (*) + 1% = 6.0% for F. " INCA /CECOR Power Peakina ll 6.2% (*) + L% = 7.2% for F:

Uncertainty." May 1980.

The 1% increase in F uncertainty for monitorina linear heat rate will

~

be accomplished by uodatina the card value for the measurement calcula-tional uncertainty factor in CECOR and Mini-CECOR. For F the lll l

1% uncertainty will be conservativolv ano1 Led to the untY. _aDQ_Ea lted Deakina factor measured values usina the followina formulas! ,

P = 1.01 F," , l F = 1.01 P p Peak Linear byeat Rate is calculated by (F

• core averace LHR) .

The current trend of the neakinc fact' ors for Cycle 16 is decreasina.

havina Deaked at accroximatelv 6 GWD for F u and F: and 2 GWD for F. llll 3'

(50-G-30,N0D-QP-3)

FORT dALHOIfN STATION FC-154 i GENERAL FORM R15 NUCTRAR SAFETY EVALUATION SEE NOD-QP-3 FOR INSTRUCTIONS ID No. 47733 Page 7 of 15 ,

(from 9.1) )

ATTACHMENT SHEET 9.8 Continued This neakina factor trend will continue for the remainder of Cycle 16. i ADplication of the 1% uncertainty to current values of Deakina factors will not exceed the TS limits. The azimuthal nower tilt will be unaffected.as the reauirement for two strinas Der cuadrant remains unchanced. This ensures adeauate core coverace. Durina the 1996 i Refuelina outaae, all currently inoperable ICI strinas are scheduled to be replaced with new instruments. Therefore, Cycle 17 will becin with at least 75% onorable strinas, as described in USAR 7.5.4.3. ,

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- GENERAL' FORM R15 NUCLEAR SAFETY EVALUATION Y.s SEE NOD-QP-3 FOR INSTRUCTIONS g jg ID No. 41'l33 Page & of d l (from 9.1) SECTION B Unreviewed Safety Question Determination 10.1.1 Identify Plant Specific Design, Operating and Technical Documents .

Document Title ID Number Revision )

U.SA/E. Sec f/0N  ?, L il. ] _ f DfD OZ-NT .1 10.1.2 Identify Applicable NRC Documents / Industry Standards ID Number Revision j Title kr+ 9er. Antw$ne SS~ 1 l

reu 9ec h*we.d 47 10.1.3 Identify Related Drawings Title ID Number Revision 7An fu:ves I. A . 7 1 'Y 10.2 List safety functions the affected structures or component;s perform: Teen f/ec. mm ru.wa of peae uwear-JJerd' rn+ 9. PLANAe AAAtAt switMc fearve. . rnrecrul</A%daal 3 Stub nd is 9 tuortsst rit r-Pen ,k List applicable accidents for which these safety functions are required: NONE B-1 (S0-G-30,N00-QP-3)

( - - .= - - - _ . . __-- -~ . . _ - - _

FORT CALHOUN STATION FC-154

- GENERAL FORM R15 l NUr'r.FAR SAFETY EVALUATION 6 SEE NOD-QP-3 FOR INSTRUCTIONS i h-ID No. M SS Page -@~ of A (from 9.1) <

I SECTION B ij Unreviewed Safety Question Determination i

1 10.3 System Interactions Analyses Criteria Aeolicable criteria Aeolicable l

j .

Fire Protection [ ] Structural Impact [ ]

~

Electrical Equipment Separation Criteria ()

Qualifications ()

' Single Failure Criteria ()

! High Energy Line Break i Review () Possibility of Operator Error ()

{ Seismic Interaction and Qualification [ ] Heavy Loads ()

Electrical Systems Analysis ( ) Impact on HVAC ()

1 Human Factors Review [ ] System / Component Performance ((

! Security Review ()

J Natural Phenomena ()

Environmental i

, Radiological Release () Installation of Temporary '

1 Modifications ()

Materials Compatibility [ ]

Testing of Temporary i i

Contairaent Intagrity [] Modifications []

]

Control Room Habitability () Other: ()

1 i Missile Protection [ ]

l _ _ _ .

Discussion of Applicable Systens Interactions Analyses

(Include Attachment Sheet as needed) rer q, jam _/5 Josr/F/e/) re4 (19 TD Bo '/o tuoPersble sir Nr,$ to?r $bh o f aa cUa bru

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FORT CALHOUM STATION FC-154 2- GENERAL FORM R15

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NUCLEAR SAFETY EVALUATION es**K' SEE NOD-QP-3 FOR INSTRUCTIONS g g l

. ID No. WM Page 4 of -12f (from 9.1) SECTION B j Unreviewed Safety Question Determination Could the proposed activity increase the probability 2

10.4 YES ( )

of occurrence of an accident previously evaluated in the USAR? .

NO (X)

Explain: No Ofkeet hee Aeoh's /M$e to affu/ Yde coa /lnurxbh o.e spexf) o f tat J y Ter't or //Ad . TAere lo're, ff e Ars(abo' ol'occ>re w er ot' ap aec 0!N DresinoysI., easfua au 'fke US*%2. A No T

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4 10.5 Could the proposed activity increase the consequences YES ( )

of an accident previously evaluated in the USAR? NO %j g Explain: AAbbiY Fes-soce+ewhs a.uS Ne aplacNtok of gea 0ei ou Use ya/ses ef #

l kraere/ k, Ike iou re efelnldri an'cr h corne.n'scas w,'& Yke /,d ab /ke Tec.eSpeca o.ud #4.4 Lh e mor t (U Nt [v g] hesse s s 3 Alt VCOY 6MO Yt Cert Onust' $;.sso c 5q;sulsb tkg

^ * '")2,.G. reed't== 3/ 4/44 **$W V!'*b nak,ses reeas'c, uN0 so u'N. os ib -GO% XC1 sInn,<,5 MI*0. TAerefore, 7ter-e s1 Ho overuse su lhe l

i consegcemos of w ucadout pressosty 54wsid Ca l*ce USht.

10.6 could the proposed activity increase the probability YES [ ]

of occurrence of a mal' .uiction t of equipment important, to safety previously eval'1ated in the USAR? NO (,5()

Explain: 7b s,anad h @ wl4 ~/~ ^ Ap'~'4 le & M '4=4-

• • -B.t.% in'of.iq 1% w:

Ey ' .F mble Im Jumarws. AG Q:c.I 4/M h*s. Is Hm flC S*we*~ s.-lIl o<

' hs.*s. #<.ar:<. w l des,% wae.:r: " %.s 6L hrt' JC2' %s M s.~'!/be ~ % A % d M k 'N v ,

het sow *a 4 can=* Jas J zJ.nsm/ nnf.c + Ai SCf La<, N +,4,.W/,w el as

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10.7 tVw.sCould

, ,j egy;an,,,4 the propose  ;,,,,,,dactivity increase the consequencesu,,p. ,l, s. YESpsp., [ ] ast f.e.{ ,% jsc of a malfunction of equipment important to safety previously evaluated in the USAR{ *** 5/'/% , g .py.jty y 0 Explain: LSAita mw.,isaan !!/b< n'rh emes, vel.1 wit. c.~/.'n a un m.e, y'w s,u Isrs Asn 75's

• Fins IAset (<>wmat s art U di)ti.'s Hu I:sss.h sisk:F:n in the Ge4. t'en.I+;u s ,'W 4e <wlaJ fo d+ w. lou rouawmA

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

l FORT CALHOUF STATION FC-154  !

GENERAL FORM R15

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NUMRAR SAFETY EVALUATION 3i% l SEE NOD-QP-3 FOR INSTRUCTIONS !I g ID No. D '> '2/ Page M of M l (from 9.1) 1 SECTION B.

Unreviewed Safety Question Determination 10.8 Could the proposed activity create the possibilit;y of YES ( )

an accident of a different type than any previously evaluated in the USAR? NO ()c]

Explatn: % dmrru l <Jon <e ' I

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10.9 Could the proposed activity create the possibility of YES [ ]

a malfunction of equipment important to safety of a different type than any previously evaluated in the NO ()c)

USAR?

Explains $*ha rhit & y<et das not itk. n c.,,k._ <

4,% ,,o,s ,,s+,w nf- rha ,1/sor,}2e c witosJ f.*x Ih: thauv& 6w 1ks W .hesM u.il(

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10.10 Does the proposed activity reduce the margin of safety YES [ ]

as defined in the basis for any Technical Specification ~l NO ()d Explain: The, is e <s.

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FORT CALIiOUN. STATION FC-154 GENERAL FORM R15 NUCLEAR SAFETY EVALUATION y,,

i SEE NOD-QP-3 FOR INSTRUCTIONS 4 ID No. #f'71 % Page O of M (from 9.1)

SECTION B Unreviewed Safety Question Determination 10.11 Summarize USAR changes which are needed 2r attach marked-up copy of affected pages: 1

$4 c, H2.=4 ed ()MS ~7 f. </. 3 M <ek -w

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10.12 Annual report of 10 CFR 50.59 changes, tests and experiments.

Ch u ,t. - 9me/, .

Provide a brief description of the activity: '

< ./ )

[a, r .Iyv.bw 0.C

• Atf 'Z h> e !/o - he SG Sm+.,n .+o n . ,, '

Ess MG Osm b te t,v /J4, les s /1me ~1Sle -.s 1 d*~ h c n Me,x J/ i&e t: Th) ss 6 aII dct er.n sours % k .M n ~.%t< o 4 C.x t- L J ' I f$

JJh /ar +$M3

• Summarize the safety evaluation: M( A $ss A e ..t-inn u vau:na s.Aeq me,+w

• Go to the Nticlear Safety Evaluation Conclusion B-5 4

(50-G-30, N00-QP-3) 4

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

t

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a.y ~% a 5,wv

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7.5.4.2 ,

System, Description  ;

.. 1H-w I

The in-core instrumentation system consists of 28 fixed in-core detector i assemblies inserted into selected fuel assemblies. Each assembly contains four rhodium detectors, and one thermocouple. Outputs may be read on the terminals and printers in the control room. These units with their cabling are contained inside an Inconel sheath.

Assemblies are inserted into the core through six instrumentation ports in the reactor vessel head. Each assembly is guided into position in an empty CEA 1 tube in the center of the fuel assembly via a fixed stainless steel guide I tube. The seal plug forms a pressure boundary for each assembly at the l reactor vessel head as does the Gralock adaptor hub to the reactor vessel flange assembly.

The neutron detectors produce a current proportional to neutron flux by a l neutron-beta reaction in the detector wire. The emitter, which is the central <

conductor in the coaxial detector, is made of Rhodium 103 and has a high  ;

themal neutron capture cross section. The rhodium detectors are provided to '

measure flux at four axial locations in the fuel assemblies. 1 The data from the detectors are read by the Emergency Response Facilities (ERF) plant computer which scans all assemblies and prints out the data periodically or on demand. The computer continually computes integrated flux at each detector to update detector sensitivity factors to compensate for detector burnout.

7.5.4.3 ICI Requirements for Monitoring Technical Specifications On July 16, 1993, the USNRC issued a Final Policy Statement on Technical Specification Improvements for Nuclear Power Reactors. The Final Policy Statement contains four criteria which can be used to determine which constraints on the design and operation of nuclear power plants are appropriate for inclusion in the plant's Technical Specifications. The ICI System does not meet any of those four criteria. Subsequently, on February 10, 1995, OPPD requested the elimination of Technical Specification 2.10.3 and the relocation of the Technical Specification limitations on the use of the ICI System to the Fort Calhoun Station Updated Safety Analysis Report (USAR).

The USNRC issued a Safety Evaluation Report (SER), dated June 26, 1995,  !

approving OPPD's request (Reference Amendment No. 167). The SER stated that i in order to change the requirements concerning the number and location of l functional detectors, a successful 10 CFR 50.59 safety evaluation with a rigorous evaluation and justification is required. The following considerations must be included in a 10 CFR 50.59 evaluation if changes to the ICI System requirements are proposed:

1) How an inadvertent loading of a fuel assembly into an improper location will be detected,

2) How the validity of the tilt estimates will be ensured,

3) How adequate core coverage will be maintained,

4) A list of the measurement uncertainties and why the added uncertainties are adequate to guarante0 that measured peak linear heat rates, peak pin powers radial peaking factors, and azimuthal power tilts will meet TS limits, and

5) How the ICI System will be restored to at least 75 percent prior to the beginning of a new cycle.

7.5-14 R510/95

}.o.16.'.!LT13) O' y

.Th3 following inf:rmation represents the ICI requirements for measuringMe

• ' Y Technical Specification values: MS Mhe ICI-Systs, shall ba Oper:ble--with>

At 12:str75% of .li in cui e instr;=nt:, :nd -

e A ::inir- Of wh:::ver thetrIci-Systee in ; Ora detector-steing: pr--fuM-ax4;l k esed te c.eni.er the phaar rlar gth qu&drant-

,ial penth-

-facter (i ", the integreted rediali = king f::ter M, the r:did

-power-dist" Mon, the p :k linear-heat rete, and the'al= thal p;ir n [

trauf -

g An operable in-core instrument shall consist of three or more operable rhodium detectors.

A quadrant symmetric in-core instrument location shan consist 4 of a location with a symmetric counterpart in any other quadrant. s Following each fuel loading:

- The ICI System must have at least 75f, of the in-core instruments operable, and The initial measurement of the linear heat rate, F,7 F,7 and azimuthal power tilt shall consist of the first full core power distribution calculation based on in-core detector signals made at a power level greater than 40 percent of rated power.

For recalibration of the ex-core detectors, a minimum of four in core instrument locations at each detector level (or a total of 16 detectors) with at least one location in the center seven rows of fuel assemblies and at least one location outside the center seven rows of fuel assemblies shall be operable.

With the ICI System inoperable, do not use the ICI System for 1) recalibration of the ex-core detector inputs to the axial power distribution trip calculator, and 2) monitoring of peak linear heat rate and radial power distribution.

The linear heat rate shall not exceed the limits of the Allowable Peak Linear Heat Rate vs. Burnup Figure provided in the COLR when the following uncertainties are appropriately applied:

- A flux peaking augmentation factor as shown in Technical Specification Figure 2-8,

- A measurement calculational uncertainty factor of 1.062 / fu mort +6 7fil. og 4 Tc6 epimuk sna I.on. E,r um qsci, sa.t tgs7, ,g g., n

- An engineering uncertainty factor of 1.03, y ,# ,

- A linear heat rate uncertainty factor of 1.002 due to axial fuel densification and thermal expansion, and 9[ -

A power measurement uncertainty factor of 1.02.

-Using-the root e.een sq r d =th ,d; e combination of the above uncertainties yields a -h072 multiplier to the measured peak linear hear ratef mwww.,,c.

'"8 6 r,s + % 7M. 4 m. cec,neu w o.t< l.f 2s5 MUfHf lint & VeWw ~lC% cn t tg4, of- % fcf5 of.akJt 7.5-14a R0'10/95 l

t

. 't'.9: 4. Wan - fy ,F g n

>-I -1.

g The ICI System shall be operable with either:

1) At least 75% of all in-core instruments and a minimum of two in-core detector strings per full axial length quadrant whenever the ICI (F , System is used toradial monitor the planar radial g %b g

), the integrated peaking factor (F '),peaking the totalfactor pehing factor (F,'), the radial power distribulion, the peak linear heat rate, and the azimuthal power tilt, or

2) At least 28% but less than 75% of all In-core Detector Strings and:

At least two In-core Detector Strings are operable per Axial Quadrant whenever the ICI System is used to monitor the planar radial peaking factor (Fy '), the integrated radial peaking factor (F,'), the total peaking factor (F,'), the radial power distribution, the peak linear heat rate, and the azimuthal power tilt, and An increase of 1% to the total u certainties applied to the planar radial peaging factor (F ), the integrated ra ial peaking factor (F, ), and the t al peaking factor (F,p),

and, The frequency of performing RE-ST-RX-0001 is changed to a minimum of once every 15 days.

^

e V. .

i.

. FORT CAUiOUN STATION FC-68A GENERAL FORM Rio i

Training Requirements j l

PART 1 - Traln!ng Requirements and Preferred Method of Training (compieted by cogrdzant PRC Member or Qualified neviewer)

Setpoint/ Procedure Change No. M77 33 Date: 2f Commitment item No. Commitment Date [)N/A I

identify the OPPD Training Group / Department affected by this procedure. (Cirde 'A' if the training is to be dono l After the effecitvo date of implementation or 'B' if Training is to be done Before the effective date of implementation.

NOTE: It is the responsibilty of the affected Department Head, in conjunction with the Training Supervisor for that l

Crea to determine applicability to personnel in that department.  ;

Trainina Grouos:

Lic Staff . (B) l&C (A) (B) RP . . (A) (B) ,

LO .. (B) EM .

.(A) (B) Chem . .. .. (A) (B)

NLO (B) MM . .. . (A) (B) Dept Supv .. ... (A) (B)

Shift Supv (B) PE (A) (B) PRC - (A) (B)

GM (A) (B) Con Mgt Eng .. . (A) (B)

Securtty (A) (B) Const Mgt Crafts (A) (B) Sys Engr . (A) (B)

Maint Plan (A) (B) Des Engr LA) (B) 1 GET .(A) (B) Asbestos (A) (B) STA .. ... @ (B)

E Plan -(A) (B) Cent Maint -(A) (B) NSRG .

. (A) (B)

Fire Brig -(A) (B) QA/QC .(A) (B)

Other (A) (B) A Preferred Method of Trainina dk,9 [

equired Reading [)O hift [ ] Regularly Scheduled PfMXIGr h y e.,

% Hotline y gp v~-~r ,

CognizanLpAC Member /Qdailtleofteviewer C Date j

- Changes to Training Requirements Part[leted (Comp by Cognizant PRC Member or Qualified Reviewer and Trainingl Departme Changes to proposed requirements _

Notified:

Cognizant PRC Member / Qualified Reviewer Date l

Changes by-Training Department Representative Part 3 - Completion of Training Requirements Training Completed Before Implementation of Procedure:

Manager - Training or Attomate NOTE: A Completed TAP 10A form must be attached to the FC-68A form. _

4

@ G

- FORT CALHOUN STATION FC-68H ADMINISTRATION FORM R0 TRAINING TRANSMITTAL DATE: March 5. 1996 TO: Dean Podoll TPCM Coordinator FP.OM: Document control J

The following procedure (s) have been identified as requiring training BEFORE x AFTER implementation.

FC-68 #47384 EPIP-RR-17A FC-68 #47727 EPIP-TSC-8 FC-68 #47733 OI-NI-2 N -

j

/03/05/96 F ECEIVED BY Date 6

(SO-G-30)

FC/ FORMS

A FORT CALHOUN S' TATION OI-NI-2 OPERATING INSTRUCTIONS PAGE 1 OF 4 IN-CORE INSTRUMENTATION OPERABILITY REQUIREMENTS SAFETY RELATED 6H E.URPOSE PAGE

1. Operability Requirements 2 PRECAUTIONS

1. Definition:

• Core Quadrant - An area containing seven in-core Detector Strings. Core Quadrants are not strictly defined

• Operable in-core Detector String - three or more operable Rhodium Detectors

• Quadrant Symmetric In-core Detector String Location - consist of a location with a symmetric counterpart in any other quadrant

• Tilt Groups - Sets of four approximately symmetric incore detectors used in the Mini CECORE/ BASS calculation of tilts

2. Loss of the ERF renders the In-core Detector System inoperable AND Technical Specification 2.10.4(1)(b) app!ies.

3. The minimum number of detectors and proper distribution must be met to ensure operation within the Limits used as initial Conditions for the Safety Analysis are met:

a. Radial Peaking Factors (Fx/ and F/) are less than the limits of Technical Specifications 2.10.4(2) and 2.10.4(3) as provided in the COLR.

b. Specified Kw/ft Limits are less than the Peak Linear Heat Rate vs. Bumup figure in the COLR AND ensured by actuating alarms set on each individual instrument.

c. To determine the Axial Shape Index for the periodic calibration verification of the Ex-core Detector System.

d. To determine azimuthal power tilt.

RO

-* FORT CALHOUN STATION OI-NI-2 OPERATING INSTRUCTIONS PAGE 2 OF 4 REFERENCES / COMMITMENTS

1. Technical Specification:

• 2.10.4: Power Distribution Limits

• Technical Specification Amendment No.167

2. Technical Data Book Figures

• I.A.6

• I.A.7.a, b, c .

3. Commitmonts:

None i

APPENDICES None 1

l 1

l RO

t

. FORT CALHOUN STATION OI-NI-2 OPERATING INSTRUCTION ,

PAGE 3 OF 4 Information Use Attachment 1 Operability Requirements PREREQUISITES y) INIT.

1. Procedure Revision Verification Master Revision Number Octe:

PROCEDURE j,. , , , , , , - , , . . . . . _ - -

1 w u m sunuwing are roet, o

-Tf4E4Hhe lncrc Octecict Cyctcm i: cent!derosoperable ~

gg -a. At least46%of-a44n-cor+0etector-Strings-aresperebie" g 3 +. At-least-two4n-coreCatccici Otririge are-operable-per-full-AxiaP

-Quadrant.^ g,-

2. IF the In-core Detector System is inoperable, THEN do NOT use the system to monitor Fx/, F/, Radial Power Distribution, and Peak Linear Heat Rate.

3. IF calibrating the Ex-core Detectors, THEN a minimum of four in-core Locations at each in-core Detector Level (16 detectors total) with at least one location in the center seven rows AND one location outside the center seven rows of fuel assemblies shall be operable.

R0

i i

i-I N 11~.

l

1. WHEN either of the following conditions are met.

THEN the In-core Detector System is considered operable:

a. At least 75% of all In-core Detector Strings are operable and at least two In-core Detector Strings are operable per full Axial Quadrant, or 4% Ptik E s t **%

b. Between-2M and 75% of all In-core Detector Strings are operable and:

1) At least two In-core Detector Strings are  ;

operable per full Axial Quadrant,

2) An increase of 1% to the tota 1 3 Sb"" b6 y,p fp uncertainties # applied to the planar radial peaking factor (Fxyl ), the integrated l radial peaking factor (Frr ), and the total . j peaking factor (FaT ), and j

3) The frequency of performing RE-ST-RX-0001 )

is changed to a minimum of once every 15 l I

days.

i 8

) ,

-* FORT CALHOUN' STATION OI-NI-2 OPERATING INSTRUCTION , PAGE 4 0F 4 Information Use Attachment i Operability Requirements PROCEDURE (contint:ed) M INIT.

CAUTION Reactor Power shall be restricted to less than 75% of Peak Linear Heat Rate when initial measurements cannot be made.-

4. The initial measurements of Fxy', Fa', Linear Heat Rate, and Azimuthal Power Tilt after each fuel loading shall be made with the following:

a. An operable in-core Detector System with the following:

1) At least 75% of all in-core Detector Strings operable.

2) At least two Quadrant Symmetric in-core Detector String Locations per Core Quadrant.

b. Power Level greater than 40% for the first Full Core Power Distribution Calculation based on in-core Detector Signals.

Completed by Date/ Time /

R0 O

0 1 .

4 Fort Calhoun Station Unit No.1 Ol-NI-2 i OPERATING INSTRUCTION

Title:

IN-CORE INSTRUMENTATION OPERABILITY REQUIREMENTS FC-68 Number: 47244 Reason for Change: This procedure was developed to replace Technical Specification 2.10.3 that was removed by Amendment No.167.

I Contact Person: Robert Ross i

I I

1 RO ISSUED: 08-10-95 4:00 pm l

h g.

FORT CALHOUN STATION Ol-NI-2 OPERATING INSTRUCTIONS PAGE 1 OF 4 IN-CORE INSTRUMENTATION OPERABILITY REQUIREMENTS l

SAFETY RELATED 61I PURPOSE PAZ

1. Operability Requirements 2 PRECAUTIONS

1. Definition:

. Core Quadrant - An area containing seven in-core Detector Strings. Core Quadrants are not strictly defined e Operable In-core Detector String - three or more operable Rhodium Detectors e Quadrant Symmetric in-core Det scior String Location - consist of a location with a i symmetric counterpart in any othar quadrant 1

l e Til: Groups - Sets of four approxima,ely symmetric incore detectors used in the Mini CECORE/ BASS calculation of tilts

2. Loss of the ERF renders the In-core Detector System inoperable AND Technical Specification 2.10.4(1)(b) applies. l l

3. The minimum number of detectors and proper distribution must be met to ensure operation within the Limits used as initial Conditions for the Safety Analysis are met:

a. Radial Peaking Factors (Fy and F n') are less than the limits of Technical Specifications 2.10.4(2) and 2.10.4(3) as provided in the COLR. (

b. Specified Kw/ft Limits are test th'an the Peak Linear Heat Rate vs. Burnup figure in the COLR AND ensured by ac'uat;ng alarms set on each individual instrument.

c. To determine the Axial Shape Index for the periodic calibration verification of the Ex-core Detector System.

d. To determine atimuthal power tilt.

R0

g. _.-_-. - . - . . . - . - _ - .. .. ..

' - ? -

J FORT CALHOUN STATION OI-NI-2 OPERATING INSTRUCTIONS PAGE 2 OF 4 REFERENCES / COMMITMENTS l

1. Technical Specification.

e 2.10.4: Power Distribution Limits i

e Technical Specification Amen.dmant No.167

2. Technical Data bok Figures e I.A.6 e I.A.7.a, b, c

3. Commitments:

None

&fENDICES None l

l l

l t

R0 ,

1

- mus e

~

FORT CALHOUN STATION OI-NI-2 OPERATING INSTRUCTION PAGE 3 OF 4 Information Use Attachment 1 Operability Requirements M) INIT.

PREREQUISITES

1. Procedure Revision Verification Master Revision Number Date:

PROCEDURE

1. WHEN the following are met, THEN the in-core Deittetor System is considered operable;

a. At least 75% of all in-core Detector Strings are operable,

b. At least two In-core Detector Strings are operable per full Axial Quadrant.

2. IF the In-core Detector System is inoperable, r

THEN do NOT use the system to monitor Fx/, Fa Radial Power

{ Distribution, and Peak Linear Heat Rate.

3. IF calibrating the Ex-core Detectors, THEN a minimum of four in-core Locations at each in-core Detector Level (16 detectors total) with at least one location in the center seven rows AND one location outside the center seven rows of fuel assemblies shall be operable.

t RG

-- - -- m

~

FORT CAlllOUN STATION Ol-NI-2 OPERATING INSTRUCTION PAGE 4 OF 4 Information Use Attachment 1 Operability Requirements

.1/J. INIT.

PROCEDURE (continued)

CAUTION Reactor Power shall be restricted to less than 75% of Peak Linear Heat Rate when initial measurements cannot be made.

4. The initial measurements of Fx/, F/, Linsar Heat Rate, and Azimuthal Power Tilt after each fuel loading shah ne made with the following:

a. An operable In-core Detector System with the following:

At least 75% of allin-core Detector Strings operable.

1)

2) At least two Quadrant Symmetric In-core Detector String Locations per Core Quadrant.

b. Power Level greater than 40% for the first Full Core Power Distribution Calculation based on in-core Detector Signals.

Date/ Time /

Completed by _

R0

i ,

LIC-96-0157 Attachment 2 i

j I

l l

1 l

0 .

l

~

NUCLEAR OPERATIONS DIVISION

- N00-QP-31 QUALITY PROCEDURE PRC Ralend PAGE '21 of 22 Appendix B PRC Mtr. Mineta i PAGE 1 of 2 1 Operability Evaluation Form AUG 151996 CR No. Ici'iG o o ar ye l

Initiating Event: Date: 6/S / c.,

(Maint., Surveillance,etc.) '

I. Affected Item (SSC): Equipment Tag No.

I c f re . tor:. g y. (Cs(e.<.)

l II. Describe the operability concern associated with the affected item, including a discussion of the requirement or commitment established for the item and why the requirement or commitment may not be met: 4.,4 3 ,, f5 p.se G ), -c ,.,r m i +:it l es-o r b<. Assv :,sa ,n w su v r e o -- i, a.n 't o '7. ,

• I<. , r ass., ps, a., s.,,.:o.,

1 t+t< t.e. ,*.s er th~ ino, a y . C/ l%aoo97> swks.s s f ;kso ICi ' ^ " *e de ts.W

k. s c.e,s e p., w A. +ns e.n.,re ., ra p, , p., ,,a t 3 77 p , ,, 9.<,, ,7. . m.

a

.s= 9: & - .3. % . .,e m.7 4 ,< t ,.e ; g ,a<<. a .,,e 4 7 III.

Describe 7.4.2): ik r<the intended f sy.4.m A ., c,,safety %g L:3 functionr h (s) .. of ,o the affected

'r # it s,._ , y _..~ , . < . -

, r. ,., . .,. _ . s ,., .

W e A d h H-c P<<, 7: .ra.e; g 7,.,r a , s r **t T> Hrasem 4,m.w n

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3 "'-1 . '%ill. ,. / % :. e the IV. List references used (See Section 7.4.3): -

prfs.ym.,,,J,.,,,,f

< les ,< , c. , ,. m ] g M^

,g,,7,,,.#

OMA '7. 5 . +. '3 */

T1 s.so U.)s. sad b. 466 k t%,

,4 ;., ,, j,, g,4 g f<. . F2 . oon f> 2. I o. .t (4) A . *s 6 . ,* C . N'~ Pi>re.brea (, i., ,s.,',.c.4 ,,. ' " '"

• [

V. Based on a review of the intended safety function (s) of the affected item and the operability concern (See Section 7.3.2):

, X , The item is OPERABLE, or The item is IN0PERABLE (IMMEDIATELY NOTIFY THE SHIFT SUPERVISOR)

• • • • • • 0PTIONAL******

Preliminary determination is as above. Further determination / evaluation must be made by the Production Engineering Division. EAR #

has been issued.

R10

• e  ? ., .

NUCLEAR OPERATIONS DIVISION N00-QP-31

(, QUALITY PROCEDURE' PAGE 22 of 22 Appendix B PAGE 2 of 2

, Operability Evaluation Form 4

CR No. / 4 % o .2 9 7 a VI. Justification of Decision (See Section 7.4.4): .

.> q.i,<. trs A s y, + Ya t:It omt<..t. +:., l,.

+- .a = u a-< t ,

cdc.r. :ns,a w s:.,n.v . toy , ;r Q + c<'c tu.-

, c.,, >a.,

ve:, s ,;ee ,;en 4a ,,,, s a ge n + s7 y ,<,,, .i. u a u <, n :., . ,, , e m -:. --- , % n :. .; c n ; e 3 m __. *~

w- -

% c_;: .m:  ;,a._- :_

(F n ,,;; s ,,: ,. ., .w. s. u z u e , as 4 g .. .

n_

VII. List any recommended corrective actions to correct the initial operabili y -

concern. for A, ,e ,r:.ua ascp tvop.,y;. n;.am +,n % rs. s,u M A :n- s a g : n .,n..( , % */ t: st, e.;w Hi+ ;t'~

N ** "' ' ' s lls.sa.o

• Prepared by: MM< ' 9 l -

Date: 6/9/are '

Time: /o 5 f ~O 3.p(Qq ! .

l (10 CFR 5pfPreparer Q'ualified) , A6i e,..,,..,.

,s .,

w )'

/ M'W[. /

i Reviewed by: @ Date: d > !d G Time: n 1 6:,% ,

(10 CFR 50.59 Reviewer QuaJified) i'

//  ;, ta,, *) i i

IndependentA M (4 % )

e Review by: V)

// f sm , . o i

@A Date: 6/ f/ /6 Time: // '3J

$10 CFR 50.5VReviewer Qualified)

W#

/ ,

c.ow 4.r. n t ISI Coordinator: NfA N Date: Time: "N

, h C M < .:

Concurrence: . m D Date: 92 9 #10. Time: /W6 I d # '( s w j' (Licensed Senior Operator) 458, Concurrence:~ D/// A / Date: f-O '/-//a Time: /9#8 >

/ (Shift Supfvisor)

Concurrence: Date: D- N Time: \N (Plant Nanager/Supv-Operations)

• NOTE: All Operability Evaluations REQUIRE PRC Review.

t ISI Coordinator's signature ONLY required for ISI-related SSC.

1kseqn. , y M19 r.~ t:s+ un ta. t t:.a :, su,;J4 i., art:a, . r . ., yp t:ssu

%* (%x GkllbreMan oEr t% e, Cacu n s We't% /syts# fo RCles*4%I(nst t"lk.

W elt "U'O fst *l pron.cyr , ins,tsJ:. ML, fe Y, P9 hso (. A + C , R10

t c he Isf..t

<sJap.4J4,,p ,p p d i t r te.,gg,;.4 ,, ; f.o;g ,,g; g h css t

• • Y

    0.03 but <0.10, correct the power tilt within two hours or determine within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and at least once per subsequent 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, that the total integrated radial peaking factor, Fa', is within the limit of Specification 2.10.4(2) and that the total planar radial peaking factor, Fy r, s within the limit of 2.10.4(3), or reduce power to less than 70% of rated power within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of confirming T, >0.03.

(c) With the indicated power tilt determined to be ;> 10, power operation r

may proceed up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> provided Fa and F,' do not exceed the power limits of the Fa', F,' and the Core Power Limitations Figure provided in the COLR, or be in at least hot standby within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Subsequent operation for the purpose of measurement to identify the cause of the tilt is allowable provided the power level is restricted to 20% of the maximum allowable thermal power level for the existing reactor coolant pump combination.

2-57b Amendment No. 32,43,4h92,141

I* 0c %C 17:17 iD:QBB CORE CtJQLYS1S FQX: PQGE 2 ABR August 5,1996 FC-FE4062  ;

Mr Marcus Guinn Omaha Pubbc Power District P O Dox 399 Fi Calhoun, NE 6802LO199

Subject:

Power Distnbution Calculation with CECOR Dear Mr. Guinn-Based on the conversations with Mr. T Heng of OPPD, the following statements apply to the validity of the CECOR results with the current complement of operable ICIs at Ft Calhoun. The statements are based on the assumption that a suMcient number ofICIs arc operable to satisfy the requirements of the 10 CFR 50 59 already in place at Ft. Calhoun, for a minimum of 2R% ICIs operable Under this assumption, the 3 D powcr distribution calculated by CECOR is valid and the calculated power distribution mcludes the effect of tilt measured by the ICis. Therefore:

1 The power distnbution calculated by CECOR is salid although the tilt estimate may not be valid.

2 The calibration of the ex< ore ASI to the corc average ASI is appropriate as long as the l core is unrodded.

3. The calculation of the Alarm Limit Signals is valid sinoc the 3-D core power distribution calculation is valid 4 The core power peaks arc salid as talted values because the 3-D power distribution calculation uses all available ICI signals which result in a tilted core power distnbution Since the CECOR till estimate is not valid, the TS requirement on tilt has to be satisfied by the ex-core s) stem The information has not been independently reviewed and is not quality assured according to ABB CENO Quahty Assurance Program lf you have any questions, please call me at (R60) 285 5512.

Sincerely, USTION ENGINEERING, INC.

a S rvisor, Setpoint Analysis

c N L Shapiro ABB G I Vincent R C Whipple R T Pearce ABB Combustion Engineering Nuclear Operations

~

Comrate Eag>neovg 'ac PO Bos 500 Te'erhve (800) M8 19 ' '

in00 PespeCl Hels Road Fos (Ac0) 205 95t?

w.ndsor ennnect(w' 0609) 050n