Text

Proposed Tech Specs to Implement Cycle 9 Nuclear & Core thermal-hydraulic Design Methodologies
	ML20085J069
Person / Time
Site:	Wolf Creek
Issue date:	06/14/1995
From:	; WOLF CREEK NUCLEAR OPERATING CORP.
To:	;
Shared Package
ML19330G057	List: ML20085J065; ML20085J069; ML20085J076;
References
	NUDOCS 9506210409
	Download: ML20085J069 (30)
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,f Attachment IV to ET 95-0051-'

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l ATTACHMENT IV PROPOSED TECHNICAL' SPECIFICATION CHANGES

. Marked-up Technical' Specification Pages.

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9506210409 950614 PDR ADOCK 05000482 P.

PDR 1

Attachment IV to ET 95-0051 Page 2 of 17 POWER DISTRIBUTION LIMITS 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - FofX-%Z) l LIMITING CONDITION FOR OPERATION 3.2.2 Fo(X %Z) shall be limited by the following relationships:

1 F ( Z) s [F"") [ K(Z)] r;" (x, y, zj s If5[I [R(z)j for P > 0.5, and n

p F (2) s [F""] [ K (Z)] r;" (x, r, z; ; (Fn [x(zij for P s 0.5.

n 0.5

0.5 Where

Fy* $, Y, Z)

'hc me:::ur:d h :t "u:: het chann:! facter, T@X, Y, Z),-increased-by44to-account for manufacturing te! rance and fudher "> eased by 5% to ccccunt for mercurement uncede!ntyr F;"

= the Fo (Z) Limit at RATED THERMAL POWER (RTP),

as specified in the CORE OPERATING LIMITS REPORT (COLR),

Ti!ERMAL POWER P

RATED TIIERMAL POWER, and

=

= the normalized Fo(X %Z) limit as a function of l

K(Z) 1 core height, as specified in the COLR.

APPLICABILITY: MODE 1.

ACTION With Fo(Xr%Z) exceeding its limit:

l Reduce THERMAL POWER at least 1% for each 1% Fo"^(Xr%Z) exceeds the limit a.

within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 8/ hours; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; subsequent POWER OPERATION may proceed provided the Overpower AT Trip Setpoints have been reduced at least 1% for each 1% F (Z) exceeds the ilmit; and q

b. - Centre! the.^ FO te v !!hin neve ^ FO !!m!!c v;h!ch are detersned by reduc!n; the !!^viab!cTHER."^.L POWER et each point !cng the AFD

'm!! !!nec cf Spec! Scat!cn 2 2.1 et !e r.! 1% for ecch 1%

17 (X,%Z)+xceedtr4he !!m!! ei!!hin 2 heure and dec'eratheAFD men!!ct !:rm incpesab!e un"' the AFD !2rm cetpe!nM are reset to the m^dif d !!m!!:; and

-o-POWER OPERAT404mayfreceed4or+p40-atotal-ofJahours; cubsequen! POWER OPERAT!OM may preceed prev!ded the Overpce cr AT T ip Sctpc!nte have been reduced at ! cant-44for-each 1% Fy'(X,%Z) r exceeds-the4mit;-and WOLF CREEK - UNIT 1 3/42-4 Amendment No.44 l

s

Atttchment IV to ET 95-0051 iPage 3 of 17

POWER DISTRIBUTION LIMITS

'3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - Fo(X#.Z) l LIMITING CONDITION FOR OPERATION (Continued)

-db. Identify and correct the cause of the out-of-limit condition prior j

to increasing THERMAL POWER above the reduced limit required by ACTION a., above; THERMAL POWER may then be increased provided Fo(Xr%Z) is demonstrated through incore mapping to be within its l

limit.

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1 WOLF CREEK - UNIT 1 3/42-5 Amendment No.-64 l

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W 1Attochment IV to ET 95-0051

.Page-4 of'17' POWER DISTRIBUTION LIMITS

SURVEILLANCE REQUIREMENTS -

i 4.2.2.1 The provisions of Specification 4.0.4'are not applicable.

4.2.2.2 F70' V,Z) Fo(Z) shall be evaluated to determine if Fo(Z) F (XMZNs within its limit by:

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

b. Increasing the measuredMeasunng Fo(Z) Fy(X,Y,Z) et tM :--! ef: component of the power distribution map by 3% to account for manufacturing tolerances and further increasing the value by 5% to account for measurement uncertainties. Verify ~

l

. that the requirements of Specification 3.2.2 are satisfied.

4 A' '--

c -ce p^' 215"::S;- Fe!! F:.:: Dr;:, er

2. ^Mr e-^ etM; by 22% er - er: Of A.TED THEP"^.L PO'.^5P 'M TN.! v'.!i Ff(X,Y,Z) r :: !--! d:t;:."n;d*;

c. Srfe';!n;!6e : ' "'--i!;;::::nt:d 5 S;::""- : 2.2.2; i

cd. Satisfying the following relationship:

F7(X,Y,Z)flF M j

o

[F""][K(Z)] for P > 0.5 p

I F."(Z) s

[P][W(2)J

[F""][K(2)]-

o F."(Z) s for P s 0.5

[0.5][W(Z)]

.]

where Fo"(Z) is the measured Fo(Z) increased by the allowances fotiF (XMZ)Prepr r nt: the. W a

dM;n ;;":::d:-" M' n her?r!?d by :- ?!!r';. : f^he ON;M^d d*Yht!^- 5 tT ::.. *he.

nc. n ! d-d;n per;;r d:r'!Mbn and tM ----

.T.:nt end !:

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manufacturing tolerances and measurement uncertainty and W(Z) is the cycle dependent function that accounts for power distribution transients encountered during normal operation. This function is provided in the COLR.

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L-AttOchment IV'to ET 95-0051.

-Page 5 of 17

d. Measuring Fo"(Z) according to the following schedule:

1. Upon achieving equilibrium conditions after exceeding, by 10% or more of RATED THERMA,L POWER, the THERMAL POWER at which Fo(Z) was

- last determined, or

2. At least once per 31 Effective Full Power Days, whichever occurs first.

e. With measurements Indicating maximum' o

s K(Z)j overz has increased since the previous determination of Fo"(Z), either of the following actions shall be taken:

1. Fo"(Z) shall be increased over that specified in 4.2.2.2.c by an appropriate factor specified in the COLR, or

During power escalation at the beginning of each cycle, THERMAL POWER may be increased until a power level for extended operation has been achieved after whichand a l

power distribution map may be obtained.

WOLF CREEK - UNIT 1 3/4 2-6 Amendment No.4%

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^Attcchment'IV to ET'95-0051 Page 6 of 17 EQWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued)

2. Fo"(Z) shall be measured at least once per 7 Effective Full Power Days until two successive maps indicate that

'F"(Zi maximum o\\/

Is not increasing.

( K(Z)s l.

over z t

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f. With the relationships specifien.n 4.2.2.2.c above not be'ing satisfied:

.!' Attachment.IV to ET 95-0051

'Page 7.of 17

.1. Calculate the percent Fa"(Z) exceeds its limit by the following expression:

y maximum - Fy(Z) X W(Z)

X 100 for P 2 0.5 '

-1 over Z F'""

X K(Z) c P

,s maximum Ff(Z) X..W(Z)

-1 X 100 for P < 0.5 over Z F'""

X K(Z) 0.5

,s

(

2. Either one of the following actions shall be taken:

.I i

a. Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , control the AFD to within new AFD limits which are determined by tightening both the negative and positive AFD limits of

+

' Specification 3.2.1 by 1% AFD foreach percent Fo"(Z) exceeds its limit and declare the AFD monitor alarm inoperable until the AFD alarm setpoints are changed to the modified limits, or

b. Comply with the requirements of Specification 3.2.2 for Fo(Z) f exceeding its limit by the percent calculated above.

g. The limits in Specification 4.2.2.2.c,4.2.2.2.e and 4.2.2.2.f are not applicable in the following core plane regions as measured in percent of core height from the bottom of the fuel:

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

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

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F i

WOLF CREEK UMT 1 3/4 2-7 Amendment No.44 l

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Attachment IV to ET 95-0051 LPage 8 of 17 f

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. WOLF CREEK - UNIT 1 3/4 2-8 Amendment No.-84

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.Attechment IV to ET 95-0051 p

~ Page 9 of 17 POWER DISTRIBUTION LIMITS -

- 3/4 2 3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR - F $H E A Y4 LIMITING CONDITION FOR OPERATION i

3.2.3 F [g -F,%V) shall be limited by the following relationship:

Fyn caH"M@V)s F$7 [1.0 + PF s (1.0 P)) FAHR %Y)

L 3

where, F$7

=

The F$n limit at RATED THERMAL POWER (RTP) specified FAPoM%Y)2 themansmmmeasured refe' perh ret!c<lefined in the Core Operating Limits Report (COLR).

l the power factor multiplier for F $n specified in the COLR, FAPR %Y} =

b PFan

=

'M '-- % e.:t;;f'erefc'; 2h 'ates-defined and ;r0* d !^ 'he COLP TilERMAL POWER

,and P

=

RATED TilERMAL POWER F5 Measured values of F5 obtained by using the movable incore detectors to

=

obtain a power distribution map. The measured values of F5 shall be used since an uncertainty of 4% for incore measurement of FE has been included in the above limit.

APPLICABillTY: MODE 1 ACTION With FE F,(XrY}exceedingitslimit:

a.

Within 44 hours5.092593e-4 days 0.0122 hours 7.275132e-5 weeks 1.6742e-5 months , either

1. Restore F5 to within the above limit, or

2. rReduce the 2!!cedi THERMAL POWER to less than 50% of from-RATED e

THERMAL POWER and reduce the Power Range Neutron Flux - High Trip Setpoint to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

et '--g moug. fer 739 u4 32 p3pq"(XrY) erreed the

" " end

b. '"% 9 heur: eith^';

o fe : F ^"9"(X,v) te ="':- 'he !: " 8^' RATED THEP PO?SR, er e

2.

etre tM PC;;r 9:n;: Meet'^ c!ux

":;h 'r! S:!;e!rt et 'r:9

?

oougy7 egg 4g.ge r3gou(y,v)erreeg ;3 e_a,3 3

- be. -Within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of initially being outside the above limit, verify through incoreeether, flux mapping that F5 has been restored to within the above limit, or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Attachment IV to ET 95-0051 Page 10 of 17 rr'e e FAHR"(X,v) te e "h!n the ! :t for-RATED THEP"^L POM'ER, or 4

o

2. o der-'he fe!!cMng 2^*!cn::

e 2.

edure the OTAT K,4erm by et !:::t TRH** for e:ch 1% that FAHR *(X,v) exceed! the ! ", and

b. Ver!Fj threugh tre^re m:pp!ng th:t FAHR"(X,Y) !: re:!cred to

.,aw:-.g !: a ye,.3e 19gp62 aL poi,ago 2"aefed by ACT!ON 2, er educe THEP^.L POWEP te !rce than 5% of RATED THEP*' AL i

oOMEP ~"- 'M ner! 2 heure, end c.

Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced limit required by Actions a. or b., above;

^

subsequent POWER OPERATION may proceed provided that F$g s i

I demonstrated through in-core flux mapping to be within its limit at a nominal 50% of RATED THERMAL POWER prior to exceeding this THERMAL POWER, at a nominal 75% of RATED THERMAL POWER prior to exceeding this THERMAL 1

POWER and within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after attaining 95% or greater RATED THERMAL POWER.

RRH !! the emeunt of THEP^.L POMEP MuMc requ: red te compene te fer each 1% th2! FAHRM(X,v) exceed: FAHR'(X,Y) and ie :p-!SM in the COLA e4RH !e the emeent Of OTAT K, cetpcint reduct!cn required te ccmpensate4er each 1% +het CAHRM(XrY) exceede the !:-:t and !: sp-! Sed in the COLA I

WOLF CREEK - UNIT 1 3/4 2-9 Amendment No. 24,64 l

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Attcchment IV to'ET 95-0051 Page-11'of 17, POWER DISTRIBUTION LIMITS

- LIMITING CONDITION FOR OPERATION MEElGNKreelems8

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' ^TED.THERMALrPOWER, 4

l SURVEILLANCE REQUIREMENTS - 4.2.3.1 F h shall be determined to be within its limit by using the movable incore detectors to obtain a power distribution map:

a.

Prior to operation above 75% of RATED THERMAL POWER after each fuelloading, and b.'

At least once per 31 Effective Full Power Days, and c.

The provisions of Specification of 4.0.4 are not applicable.

T M pr r! 9 ^e e' e-r!S e e - 4.0.4 ere net 2- "cr5 "

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c",d the m- ~um mer;!n fer !!!'^et!^^ - 2 ned !=

' 2.2.2.5.4, c'^re. "*he m": mum m r;!r ! !e:: then 0, ecmp!y witMhe.^CT!ON re'9ement: Of Spee! Set ~t 2.2.2.

WOLF CREEK - UNIT 1 3/4 2-10 Amendment No.44

Attcchment IV to ET 95-0051 Page 12 of 17 ADMINISTRATIVE CONTROLS I I I II I

H CORE OPERATING LIMITS REPORT (COLR) '

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

OPERATING LIMITS REPORT (COLR) before each reload cycle or any remaining part of a reload cycle, for the following:

l

.1.

Specification 3.1.1.3: Moderator Temperature Coefficient (MTC) EOL limits

2. Specification 3.1.3.5: Shutdown Rod' insertion Limit

3. Specification 3.1.3.6: Control Rod Insertion Limits

4. Specification 3.2.1: Axial Flux Difference (AFD)

5. Specification 3.2.2: Heat Flux Hot Channel Factor - F (Xr%Z) l o

6. Specification 3.2.3: Nuclear Enthalpy Rise Hot Channel Factor -

F,9;rY) F5

7. Specification 3.9.1.b: Refueling Boron Concentration The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents.

a. NRC Safety Evaluation Report dated October 29,1992, for the " Core Thermal Hydraulic Analysis Methodology for the Wolf Creek Generating Station"(ET-90-0140, ET 92-0103)

(Methodology for Specification 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor-F$ F,(4V) 1

b. NRC Safety Evaluation Report dated January 17,1989, for the

" Acceptance for Referencing of Licensing Topical Report WCAP-11397, Revised Thermal Design Procedure."

(Methodology for Specification 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor-F$.

cb. NRC Safety Evaluation Report dated September 30,1993,(ua^n burnce) for the " Transient

.l Analysis Methodology for the Wolf Creek Generating Station"(ET 0026 ET 92-0142, WM 93-0010, WM 93-0028)

(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient [MTC))

g

m_

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WOLF CREEK-UNIT 1 6-21 Amendment No 42 64 l

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Attachment IV to ET 95-0051 Page 14 of 17 ADMINIST_RATIVE CONTROLS CORE OPERATING LIMITS REPORT (COLR) (Continued) l

d. NRC Safety Evaluation Report dated November 26,1993, k

" Acceptance for Referencing of Revised Version of Licensing Topical Report WCAP-1021G-P-A, Relaxation of Constant Axial Offset Control - Fo Surveillance Technical Specification" (TAC No. M88206).

(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor - Fo(Z): Specification 3.1.1.3 - Moderator Temperature g

Coefficient (MTC): Specification 3.1.3.5 - Shutdown Rod insertion J

Limit: Specification 3.1.3.6 - Control Rod insertion Limits:

{

Specification 3.2.1 - Axial Flux Difference: Specification 3.2.3 Nuclear Enthalpy Rise Hot Channel Factor-F$: Specification 3.9.1.b -

Refueling Boron Concentration).

ed. NRC Safety Evaluation Report dated March 10,1993, for the " Reload

=

Safety Evaluation Methodology for the Wolf Creek Generating Station" (ET 92-0032, ET 93-0017)

(Methodology for Specification 3.1.3.6 - Control Rod insertion Limits; Specification 3.2.1 - Axial Flux Difference) fe. NRC Safety Evaluation Report dated March 30,1993, for the " Revision l

to Technical Specification for Cycle 7" (NA 92-0073, NA 93-0013, NA 93-0054)

(

(Methodology for Specification 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor - Fj ((4V), [Use of WRB-2 Correlation with VIPRE-01 Code])

gf. NRC Safety Evaluation Report dated November 13,1986, for "The 1981 l

-/

Version of the Westinghouse ECCS Evaluation Model Using the BASH Code" (WCAP-10266-P-A, Rev. 2)

(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Facics - F (%Z) o h.

NRC Safety Evaluation Report dated May 17,1988," Acceptance for Referencing of Westinghouse Topical report WCAP-11596 -

Qualification of the Phoenix-P/ANC Nuclear Design System for Pressurized Water Roactor Cores."

(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor-F (Z): Specification 3.1.1.3 - Moderator Temperature o

Coefficient (MTC): Specification 3.1.3.5 - Shutdown Rod Insertion Limit: Specification 3.1.3.6 - Control Rod insertion Limits:

Specification 3.2.1 - Axial Flux Difference: Specification 3.2.3 Nuclear Enthalpy Rise Hot Channel Factor - F$ : Specification 3.9.1.b -

Refueling Boron Concentration).

b

Attcchment-IV to ET 95-0051 Page 15 of 17 l} NRC Safety Evaluation Report dated June 23,1986," Acceptance for Referencing of Topical Report WCAP 10965-P and WCAP 10966-NP-ANC: A Westinghouse Advanced Nodal Computer Code."

(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor-F (Z): Specification 3.1.1.3 - Moderator Temperature o

Coefficient (MTC): Specification 3.1.3.5 - Shutdown Rod Insertion Limit: Specification 3.1.3.6 - Control Rod Insertion Limits:

Specification 3.2.1 - Axial Flux Difference: Specification 3.2.3 Nuclear Enthalpy Rise Hot Channel Factor-Fj,: Specifica% 3.9.1.b -

Refueling Boron Concentration).

The core operating limits shall be determined so that a'l applicable limits (e.g., fuel thermal-hydraulic limits, core thermal-hydraulic imits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator of the NRC Regional Office within the time period specified for each report.

6.10 RECORD RETENTION In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated.

6.10.1 The following records shall be retained for at least 5 years:

a. Records and logs of unit operation covering time interval at each power level;

b. Records and logs of principal maintenance activities, inspections, repair and replacement of principalitems of equipment related to nuclear safety; WOLF CREEK - UNIT 1 6-21a Amendment No.44 l

s Attcchmemt IV to ET 95-0051"'

LPage 16 of'17 3/4.2 POWER DISTRIBUTION LIMITS

-BASES-The specifications 'of this section provide assurance of fuel integrity during Condition I (Morn ~,al Operation) and 11 (Incidents of Moderate Frequency)

events by: (a) mautWing the minimum DNBR in the core greater than or equal to the DNBR de.,gn limit specified in the CORE OPERATING LIMITS REPORT (COLR) during normal operaton and in short-term transients, and (b) limiting the fission gas release, fui l pallat temperature, and cladding mechanical proper-

' ties to within assumed design criteria. In addition, limiting the peak linear power density during Condition i events provides assurance that the initial.

'i conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200*F is not exceeded.

The definitions of certain hot channel and peaking factors as used in these specifications are as follows:

F(4%2)'

Heat Flux Hot Channel Factor, is defined as the local heat l

o flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux,29"Mng Sr en69 ting L;m,n~e er, t.;l ;;;";t :.-d at, :t erremb!y (?,Y); '

F,(4VF F$ Nuclear Enthalpy Rise Hot Channel Factor, is defined as the I

ratio of the integral of linear power along the rod with the highest integrated power to the average rod power. t cremb!y "fM&P 3/4.2.1 AXIAL FLUX DIFFERENCE The limits on AXtAL FLUX DIFFERENCE (AFD) assure that the F (4%Z) and o

E,(4V) F$ limits are not exceeded during either normal operation or in the j

event of xenon redistribution following power changes. The AFD limits have

'l been adjusted for measurement uncertainty.

Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitor Alarm. The computer deter-

)

mines the 1-minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for at least 2 of 4 or 2 of 3 OPERABLE excore channels are outside the AFD limits and the THERMAL POWER is greater than %% of RATED THERMAL POWER.

i 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR AND Nt; CLEAR ENTHALPY RlSE HOT CHANNEL FACTOR The limits on heat flux hot channel factor and nuclear e'ithalpy rise hot channel factor ensure that: (1) the design limits on peak local power density and minimum DNBR are not exceeded, and (2) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200*F ECCS acceptance criteria limit.

WOLF CREEK-UNIT 1 B 3/4 2-1 Amr.ndment No. 4, St l

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+ i +4 1

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jAttqchment?.IV to ET'95-0051 i

VPage 17:of 17 POWER DISTRIBUTION LIMITS m

1 m

d' 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR AND NUCLEAR ENTHALPY RISE m

liQI. CHANNEL FACIOR (Continued).

p

Each of these is measurable but will normally only be determined.

! periodically as specifled in Specifications 4.2.2 and 4.2.3. This periodic '

?

'i E surveillance is sufficient to insure that the limits are maintained provided:

B

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

insertion differing by more than 112 steps, indicated, from the '

group demand position, b

b.' Control rod groups are sequenced with overlapping groups as -

described in Specification 3.1.3.6,

~

c. The control rod insertion limits of Specification 3.1.3.6 are maintained, and

d. The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.

' Fg(XrV) F[ will be' maintained within its limits provided Conditions a.

through d. above are maintabed. The limits on the nuclear enthalpy rise hot

^

n

' channel factor,4g(XrV) F[, are specdied in the COLR,-asMaximumAllowable

_ m " ' P::S =^ L:. __. 2:- nd bj d' :d:n; 5 P__

.2

^": rd': Fxt m,m.

. u, m _.___.. w ____

...___..__,y vs o,, 2 -- __ m.

. Usu

'.; ' x: --t :: 5 rc :!x' 5 5 C"?" -- f 9 51: ';.

Fg(v.,v) 'f&r :;::!':I 5 5 COLR :.d e :

..;._!_:xx =ialpewer 7u 4

Fo (XrV,Z)'and4MR (XrV) F$ are measured periodically to provide assurance M

M 4'

that they remain within their limits. A peaking margin calculation is performed, when necessary, to provide the basis for reducing THERMAL POWER,'

or for reducing the width of the AFD limits.r.ansfer64(A!) :t '.

e'N OTAH4 ::' :'.::. ~ The hot channel factor F"(z) is measured periodically and iner ased by a cycle

+

g and height dependent factor, W(z), to provide assurance that the

- limit of F (z) is met. W(z) accounts for the effects of normal operation transients g

and is determined from expectsd power control maneuvers over the full range of bumup conditions in the core. The W(z) functions are specified in the Core

~ Operating Limits Report.

3/4.2 4 QUADRANT POWER TILT RATIO

. The QUADRANT POWER TILT RATIO limit assures that the radial power

' distribution satisfies the design values used in the power capability

= analysis. Radial power distribution measurements are made during STARTUP -

. testing and periodically during power operation.

e

_. The limit of 1.02, at which corrective ACTION is required, provides DN8 and linear heat generation rate protection with x-y plane power tilts. A limit of 1.02 was selected to provide an allowance for the uncertainty

. associated with the indicated power tilt.

WOLF CREEK. UNIT 1 B 3/4 2-2 Amendment No. 4r64r 64 l

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!AttGchment V to ET 95-0051 Page'1:of 13?

. :7. ;

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J ATTACHMENT V PROPOSED' TECHNICAL SPECIFICATION CHANGES Draft (Clean Copy) Technical Specification Pages'

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iL Attcchment V to ET 95-0051 Page 2 of 13 POWER DISTRIBUTION LIMITS 3/4 2.2 HEAT FLUX HOT CHANNEL FACTOR - Fgg) l LIMITING CONDITION FOR OPERATION 3.2.2 Fo(Z) shall be limited by the following relationships:

F (Z) s [F""] [ K(Z)] for P > 0.5, and q

P-F (2) s [F"""] [ K(Z)] for P 5 0.5.

q 0.5 Where:

F'"

= the Fo (Z) Limit at RATED THERMAL POWER (RTP),

l as specified in the CORE OPERATING LIMITS REPORT (COLR),

TIIERMAL POWER P

=

,and RATED TilERMAL POWER K(Z)

= the normalized Fo(Z) limit as a function of core height, as specified in the COLR.

APPLICABILITY: MODE 1.

ACTION:

With Fo(Z) exceeding its limit:

l

-l Reduce THERMAL POWER at least 1% for each 1% Fo(Z) exceeds the limit a.

l within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; subsequent POWER OPERATION may proceed provided the Overpower AT Trip Setpoints have been reduced at least 1% for each 1% F (Z) exceeds the limit; and q

i WOLF CREEK - UNIT 1 3/4 2-4 Amendment No. 64 l

q

'Attcchment V to ET 95-0051 Page 3.of 13 POWER DISTRIBUTION LIMITS 3212.2 HEAT FLUX HOT CHANNEL FACTOR - Fg@

l.

LIMITING CONDITION FOR OPERATION (Continued)

b. Identify and correct the cause of the out-of-limit condition prior l

~ to increasing THERMAL POWER ebove the reduced limit required by ACTION a., above; THERMAL POWER may then be increased provided Fo(Z) is demonstrated through incore mapping to be within its l

. limit.

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1 WOLF CREEK UNIT 1 3/4 2-5 Amendment No. 64 l

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AttEchment V to ET 95-0051

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. Page 4;of 13.:

POWER' DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

4.2.2.2 Fo(Z) shall be evaluated to determine if Fo(Z) is within its limit by:

l

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

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

.l

b. Increasing the measured Fo(Z) component of the power distribution map by 3% to account for manufacturing tolerances and further increasing the value by 5% to account for measurement uncertainties. Verify that the requirements of Specification 3.2.2 are satisfied.

c. Satisfying the following relationship:

[F*"][K(Z)] for P > 0.5 g

F."(Z)s

[P][W(Z)]

[F "][K(Z)]

F."(Z) s for P s 0.5

[0.5][W(Z)]

where Fo"(Z) is the measured Fo(Z) increased by the allowances for manufacturing tolerances and measurement uncertainty and W(Z) is the cycle dependent function that accounts for power distribution transients encountered during normal operation. This function is provided in the COLR.

d. Measuring Fo"(Z) according to the following schedule:

1. Upon achieving equilibrium conditions after exceeding, by 10% or more of '

RATED THERMAL POWER, the THERMAL POWER at which Fo(Z) was last determined,* or

2. At least once per 31 Effective Full Power Days, whichever occurs first,

e. With measurements indicating

F"(Z)'

maximum y

K(Z),

(

has increased since the previous determination of Fo"(Z), either of the following actions shall be taken:

1. Fo"(Z) shall be increased over that specified in 4.2.2.2.c I

by an appropriate factor specified in the COLR, or i

During power escalation at the beginning of each cycle, THERMAL POWER may be increased until a power level for extended operation has been achieved after which a l

power distribution map may be obtained.

WOLF CREEK-UNIT 1 3/4 2-6 Amendment No. 64 l

Attcchment V to ET 95-0051 Page 5 of 13

. POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued)

2. Fo"(Z) shall be measured at least once per 7 Effective Full Power Days until two successive maps indicate that

' F" (Z)'

o is not increasing.

maximum over z K(Z),

s

f. With the relationships specified in 4.2.2.2.c above not being satisfied:

1. Calculate the percent Fo"(Z) exceeds its limit by the following expression:

(

maximum F"(Z) X W(2) o X 100 for P 2 0.5

-1 over Z F"" X K(2)

(

P

_s

.g 3

maximum F"(Z) X W(2) o

-1 X 100 for P < 0.5 over Z F""

X K(Z)

(

_ 0.5 s

2. Either one of the following actions shall be taken:

a. Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , control the AFD to within new AFD limits which are determined by tightening both the negative and positive AFD limits of Specification 3.2.1 by 1% AFD for each percent Fo"(Z) exceeds its limit and declare the AFD monitor alarm inoperable until the AFD alarm setpoints are changed to the modified limits, or

b. Comply with the requirements of Specification 3.2.2 for Fo(Z) exceeding its limit by the percent calculated above.

g. The limits in Specification 4.2.2.2.c,4.2.2.2.e and 4.2.2.2.f are not applicable in the following core plane regions as measured in percent of core height from the bottom of the fuel:

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

)

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

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I WOLF CREEK UNIT 1 3/4 2-7 Amendment No. 61

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Attichment V to ET 95-0051 Page 6 'of 13 -

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i WOLF CREEK - UNIT 1 3/4 2-8 Amendment No. 64 l

i.v-Attcchment V to ET 95-0051.

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Page 7 of.13-EQWER DISTRIBUTION LIMITS 3/4 2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR - F $g -

LIMITING CONDITION FOR OPERATION l

3.2.3 F $g shall be limited by the fo!!owing relationship:

F $g 5 F $7 [1.0 + PF H (1.0 - P)]

3

where, F$7

=

The F$n limit at RATED THERMAL POWER (RTP) specified in the Core Operating Limits Report (COLR).

the power factor multiplier for Fyn specified in the COLR, PF s

=

4 TIIERMAL POWER P

=

,and RATED TilERMAL POWER

=

. Measured values of FU, obtained by using the movable incore detectors to F$i obtain a power distribution' map. The measured values of Ffg shall be used since an uncertainty of 4% for incore measurement of FE, has been included in the above limit.

i APPLfCABILITY: MODE 1 ACTION With Fji exceedingitslimit:

a.

Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , either l

1. Restore F$i within the above limit, or l

to

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

b.

Within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of initially being outside the above limit, verify through incore flux mapping that Ffu has been restored to within the above limit, or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the.

the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c.

Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced limit required by Actions a. or b., above; subsequent POWER OPERATION may proceed provided that F yn is

(

demonstrated through in-core flux mapping to be within its limit at a nominal l

50% of RATED THERMAL POWER prior to exceeding this THERMAL POWER, at 3

a nominal 75% of RATED THERMAL POWER prior to exceeding this THERMAL i

POWER and within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after attaining 95% or greater RATED THERMAL POWER.

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WOLF CREEK - UNIT 1 3/4 2-9 Amendment No. 23, St l

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AttachmInt V to ET 95-0051 Page 8 of 13 POWER OlSTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENTS N

4.2.3.1 F3H shall be determined to be within its limit by using the movable incore detectors to obtain a power distribution map:

a.

Prior to operation above 75% of RATED THERMAL POWER after each fuelloading, and b.

At least once per 31 Effective Full Power Days, and c.

The provisions of Specification of 4.0.4 are not applicable.

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WOLF CREEK - UNIT 1 3/4 2-10 Amendment No. 64 l

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" Attachment V'to'ET-95-0051 1

-Page 9 of'13 l

ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT fCOLR)

]

. 6.9.1.9 Core operating limits shall t'e established and documented in the CORE OPERATING LIMITS REPORT (COLR) before each reload cycle or any remaining

.; part of a reload cycle, for the following:

~

' 1. Specification 3.1.1.3: Moderator Temperature Coefficient (MTC) EOL limits

2. Specification 3.1.3.5: Shutdown Rod insertion Limit

3. Specification 3.1.3.6: Control Rod insertion Limits -

4. Specification 3.2.1: Axial Flux Difference (AFD) j

5. Specification 3.2.2: Heat Flux Hot Channel Factor-Fo(Z) l

. 6. Specification 3.2.3: Nuclear Enthalpy Rise Hot Channel Factor -

F$

7. Specification 3.9.1.b: Refueling Boron Concentration.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents.

a. NRC Safety Evaluation Repott dated October 29,1992, for the " Core Thermal Hydraulic Analysis Methodology for the Wolf Creek Generating Station" (ET-90-0140. ET 92-0103) -

(Methodology for Specification 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor-F$

b, NRC Safety Evaluation Report dated January 17,1989, for the

" Acceptance for Referencing of Licensing Topical Report.

WCAP-11397, Revised Thermal Design Procedure." -

(Methodology for Specification 3.2.3 - Nuclear Enthalpv Rise Hot Channel Factor-F$.

c. NRC Safety Evaluation Report dated September 30,1993, for the " Transient l

'j Analysis Methodology for the Wolf Creek Generating Station" (ET 0026, ET 92-0142, WM 93-0010, WM 93-0028)

(Methodology for Specification 3.1.1.3 - Moderator Temperature j

Coefficient [MTC))

j i

WOLF CREEK - UNIT 1 6-21 Amendment No. 42784

Attcchment V to ET f5-0051 Page 10 of 13 ADMINISTRATIVE CONTROLS C. ORE OPERATING LIMITS REPORT (COLR)(Continued) l

d. NRC Safety Evaluation Report dated November 26,1993,

" Acceptance for Referencing of Revised Version of Licensing Topical Report WCAP-10216-P-A, Relaxation of Constant Axial Offset Control - Fo Surveillance Technical Specification" (TAC No. M88206).

(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor - Fo(Z): Specification 3.1.1.3 - Moderator Temperature Coefficient (MTC): Specification 3.1.3.5 - Shutdown Rod Insertion Limit: Specification 3.1.3.6 - Control Rod insertion Limits:

Specification 3.2.1 - Axial Flux Difference: Specification 3.2.3 Nuclear Enthalpy Rise Hot Channel Factor-F$: Specification 3.9.1.b -

Refueling Boron Concentration).

e. NRC Safety Evaluation Report dated March 10,1993, for the " Reload Safety Evaluation Methodology for the Wolf Creek Generating Station" (ET 92-0032, ET 93-0017)

(Methodology for Specification 3.1.3.6 - Control Rod lnsertion Limits; Specification 3.2.1 - Axial Flux Difference)

f. NRC Safety Evaluation Report dated March 30,1993, for the " Revision l

to Technical Specification for Cycle 7" (NA 92-0073, NA 93-0013, NA 93-0054)

(Methodology for Specification 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor - F$, [Use of WRB-2 Correlation with VIPRE-01 Code])

g. NRC Safety Evaluation Report dated November 13,1986, for "The 1981 l

Version of the Westinghouse ECCS Evaluation Model Using the BASH Code" (WCAP-10266-P-A, Rev. 2)

(Methodology for Specification 3.2.2 - Heat Flux Hot Channel I

Factor - Fo(Z) h.

NRC Safety Evaluation Report dated May 17,1988,' Acceptance for Referencing of Westinghouse Topical report WCAP-11596 -

Qualification of the Phoenix-P/ANC Nuclear Desigr System for Pressurized Water Reactor Cores.*

(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor-Fo(Z): Specification 3.1.1.3 - Moderator Temperature Coefficient (MTC): Specification 3.1.3.5 - Shutdown Rod Insertion Limit: Specification 3.1.3.6 - Control Rod insertion Limits:

Specification 3.2.1 - Axial Flux Difference: Specification 3.2.3 Nuclear Enthalpy Rise Hot Channel Factor - F$ : Specification 3.9.1.b -

Refueling Boron Concentration).

WOLF CREEK - UNIT 1 6-21a Amendment No. 61 l

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- Attcchment V to ET 95-0051 ePage 11'of 13 ADMINISTRATIVE CONTROLS

COBE OPERATING LIMITS REPORT (COLR)(Continued) -

i) NRC Safety Evaluation Report dated June 23,1986, " Acceptance for Referencing of Topical Report WCAP 10965-P and WCAP '

' 10966-NP-ANC: A Westinghouse Advanced Nodal Computer Code?

o (Methodology for Specification 3.2.2 - Heat Flux Hot Channel L

Factor-Fo(Z): Specification 3.1.1.3 Moderator Temperature

- Coefficient (MTC): Specification 3.1.3.5 - Shutdown Rod Insertion Limit: Specification 3.1.3M - Control Rod Insertion Limits:

Specification 3.2.1 - Axial Flux Difference: Specification 3.2.3 Nuclear Enthalpy Rise Hot Channel Factor-Fy",: Specification 3.9.1.b -

Refueling Boron Concentration).

The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-hydraulic limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

The CORE OPERATING UMITS REPORT, including any mida:ycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle,

- to the NRC Occument Control Desk with copies to the Regional Administrator and Resident inspector.

1 SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator of the j

NRC Regional Office within the time period specified for each report. '

6.10 RECORD RETENTIOff

.I In addition to the applicable record retention requirements of Title 10, Code 1

of Federal Regulations, the following records shall be retained for at least the minimum period indicated.

-6.10.1 The following records shall be retained for at least 5 years:

1

a. Records and logs of unit operation covering time interval at each power level;

b. Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety;

c. All REPORTABLE EVENTS;

d. Records of surveillance activities, inspections, and calibrations required by these Technical Specifications;

e. Records of changes made to the procedures required by Specification 6.8.1;

f. Records of radioactive shipments;

g. Records of sealed source and fission detector leak tests and results; and h.- Records of annual physicalinventory of all sealed source material of record.

WOLF CREEK - UNIT 1 6-21b Amendment No. 6%

l

Attrchment V to ET 95-0051}

.Page 12 of 13-3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section provide assurance of fuelintegrity

- during Condition I (Normal Operation) and ll (Incidents of Moderate Frequency) events by: (a) maintaining the minimum DNBR in the core greater than or equal

- to the DNBR design limit specified in the CORE OPERATING LIMITS REPORT (COLR) during normal operation and in short-term transients, and (b) limiting the.

fission gas release, fuel pellet temperature, and cladding mechanical proper-ties to within assumed design criteria. In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA snalyses are met and the ECCS acceptance criteria limit of 2200'F is not exceeded.

The definitions of certain hot channel and peaking factors as used in

- these specifications are as follows:

l-F(Z) _ Heat Flux Hot Channel Factor, is defined as the local heat l

a l

flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, l

F$

Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the integral of linear power along the rod with the highest integrated power to the average rod power.

l 3/4.2.1 AXIAL FLUX DIFFERENCE

.l The limits on AXIAL FLUX DIFFERENCE (AFD) assure that the F (Z) and o

Fj limits are not exceeded during either normal operation or in the

)

event of xenon redistribution following power changes. The AFD limits have been adjusted for measurement uncertainty.

Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitor Alarm. The computer deter-mines the 1-minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for at least 2 of 4 or 2 of 3 OPERABLE excore channels are outside the AFD limits and the THERMAL POWER is j

greater than 50% of RATED THERMAL POWER.

l 3/4.2.2 and 3/4.2 3 HEAT FLUX HOT CHANNEL FACTOR AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR i

The limits on heat flux hot channel factor and nuclear enthalpy rise hot

!I channel factor ensure that: (1) the design limits on peak local power density and minimum DNBR are not exceeded, and (2)in the event of a LOCA the peak fuel clad temperature will not exceed the 2200'F ECCS acceptance criteria limit.

WOLF CREEK-UNIT 1 B 3/4 2-1 Amendment No. 4, St l

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7 l

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-Att chment V to ET 95-oo51 Page 13 of 13 POWER DISTRIBUTION LIMITS BASES 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR AND NUCLEAR ENTHALPY RISE HQT CHANNEL FACTOR (Continued)

Each of these is measurable but will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3. This periodic I

surveillance is sufficient to insure that the limits are maintained provided:

a. Control rods in a single group raove together with no individual rod insertion differing by more than i 12 steps, indicated, from the group demand. position,

b. Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.6,

c. The control rod insertion limits of Specification 3.1.3.6 are maintained, and

d. The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.

Fj, will be maintained within its limits provided Conditions a.

through d. above are maintained. The limits on the nuclear enthalpy rise hot channel factor, Fj,, are specified in the COLR.

F (Z) and Fj, are measured periodically to provide assurance a

that they remain within their limits. A peaking margin calculation is performed, when necessary, to provide the basis for reducing THERMAL POWER or for reducing the width of the AFD limits. The hot channel factor Ff"(Z) is measured periodically and increased by a cycle and height dependent factor, W(Z), to provide assurance that the limit of F (Z)is met. W(Z) accounts for the effects of normal operation transients o

and is determined from expected power control maneuvers over the full range of burnup conditions in the core. The W(Z) functions are specified in the Core Operating Limits Report.

3/4.2.4 QUADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that the radial power distribution satisfies the design values used in the power capability analysis. Radial power distribution measurements are m?de during STARTUP te4 ting and periodically dur:ng power operation.

The limit of 1.02, at which corrective ACTION is required, provides DNB and linear heat generation rate protection with x-y plane power tilts. A limit of 1.02 was selected to provide an allowance for the uncertainty associated with the indicated power tilt.

WOLF CREEK - UNIT 1 B 3/4 2-2 Amendment No. 4 &h 64 l

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ML20085J069

Text

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