ML20044B868

Proposed TS Figure 2.1-1, Reactor Core Safety Limit - Four Loops in Operation

ML20044B868
Person / Time
Site:	Wolf Creek
Issue date:	03/08/1993
From:	WOLF CREEK NUCLEAR OPERATING CORP.
To:
Shared Package
ML20044B867	List: ML20044B866 ML20044B868
References
NUDOCS 9303110183
Download: ML20044B868 (12)

Text

_

Attachment to NA 93-0054L l

"'1

' Page 1 of 12-l

. wt ;90 c N0!T!CNS r0R OPERATION AND SURVE!LLANCE REQUIREMENTS-

!CT:CN PAGE-i

'i 3/4.0 ApoLICABILITY...............................................

3/4 0.

3/.1 REACTIVITY CONTROL SYSTEMS j

3/4.1.1 BORATION CCNTROL Shutdown Margin - T 200*F............................

3/4 1 '

avg 3

Shutdown Margin - T,,,i 200*F...........................

3/4 1-3 Moderator Temperature Coefficient........................

3/4 1-4 l

um Teesterg_ture f ar c ri ti_ cal i ty..................... m 3/4 1-6 l

F tCrttR 5.1-1 BOL MOWWOK TEr6PEiEhiT8 LRE. CCEFACfDJT

~3g g.g*

3/4.1.2 80 RATION SYSTEM 5(VS. POW CE L5r_v tt_

Flow Path -

Shutdown.....................................

3/4 1-7 Flow Paths - Operating..................

3/4 1 -

Charging Pump - Shutdown.................................

3/4 1-9 Charging Pumps - Operating.....................,..........

3/4: 1-10

.l Bo rated Wate r Sou rce - 5hu tdown..........................

3/4 1-11 Borated Water Sources - Operating........................

3/4 1-12:

a 14.1.3 MOVA8tf CONTROL ASSEMBLIES-1 t

Group Height.............................................

3/4 1-1 l

TABLE-3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION l

IN THE EVENT OF AN IN0PERA8LE. FULL-LENGTH R00.........................................

3/4 1-15 j

Pos i tion Indication Systems - Operating....~..............

3/4'l-17 Position Indication System -

Shutdown....................

3/4 1.

t Rod Orop Time................,...........................

3/4 1-10 i

S hu tdown Rod Inse rtion Limi t.............................

3/4 1-20 Conhiel.RedInsertionLimits.............................

3/4 1- *..

.[

.FI% 4.1-1 R00 BANK INSERTION LIMITS VERSUS THERMAL POWEA-FOUR LOOP OPERATION............

3/4 1-30 Delate.

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CREEC - UNIT 1 IV i

9303110183 930308

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1 Attachment to NA 93-0054 Q pegce cairn Fo LLO kmJG PAGE

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OPERATION Mi 660

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0 0.0 '/.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 1.1 1.2 FRACTION OF RATED THERMAL PO

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FIGURE 2.1-1

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REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION j

WOLF CREEK - UNIT 1 2-2 Amendment No.

51

l Attachment to NA 93-0054 Page 3 of 12 680 1

UNACCEPTABLE 660 OPERATION 2400 PSIA N

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N N

N 640 Lu 2000 PSIA N

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2250 PSIA b 620

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560 O.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 1.1 1.2 FRACTION OF RATED THERMAL POWER l

FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION Wolf Creek-Unit 1 2-2 Amendment No.

~.

Attachment to NA 93-0054

^

4 gF. fiece m t6e CORE Page 4 of 12 REP 0PT (COLTC) oPCRnTING t_lmtTS 2.1 SAFETY LIMITS BASES I

2.1.1 REACTOR CORE The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforation which would result in the releas of fission products to the reactor coolant.

Overheating of the fuel claading is prevented by restricting fuel operation to within the nucleate boiling legime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boil ng regime could result in excessive cladding temperatures because of the on et of departure from nucleate boiling (DNB) and the resultant sharp reducti n in heat transfer coefficient.

DNB is not a directly measurable parameter du ing operation and therefore THERMAL POWER and Reactor Coolant Temperature an Pressure have been related to DNB through DNBR correlations.

DNBR correlatio s have been developed to predict the DNB flux and the location of DNB for axiall uniform and nonuniform heat flux distributions.

The local DNB heat f ux ratio (DNBR) is defined as the ratio of the heat flux that would cause DN at a particular core location to the local heat flux, and is indicative of th margin to DNB.

The DNB design basis is as follows:

there must be t least a 95 percent probability that the minimum DNBR of the limitin rod d ing Condition I and tepanareauagh mit of the DNB correlation g Ley @use. (the WRB one4ation in t..n-a he correlation DNBR g ain-TiinIMubTTsliEifDWon7 tie-EinitTrrappTic15Te tiper}imental data set such b

that there is a 95 percent probability with 95 percent c fidence,that DNE_-

11L.no1Jc ur when the minimum DNBR is at the DNBR limi (1.17fortheWRB-1}

- ccerclot4onk Fo at conditionwhich,.DLll outside th Panse of v v c16fTigof thefPj-1 correlation, the W~3] correlation is used.

app DTmkn~~~iDaiiitaIneaoyNerformingsafetyanalysesto In addit"e ion alv han the correlation limit, called the safety analysis limit a highe DNBR.

T e margin between the safety analysis limit DNBR and the correlation limit DNBR 1s Me_d_.p_cqyer_.ADEn DliBR_panaltinsandJJ_ ovide _maIginAt des.19%

flexibilit The coe-t.3 a na hras I Nit DN G R 4 speci-91ce in tAe CotR The curveUf%ure 2Mthe To'ci ofloTn'tT3TTRERRhTT0h, Reactor Coolant System pressure and average temperatura for which the minimum DNBR is no less than the applicable safety analysis limit DNBR, or the average l

enthalpy at the vessel exit is. qual,to the_._enthalpyJf._ saturated _1.igui%

p he. Jes p F w seecWieJ in These curves are based,on-en-enthcipy hot-channel-fac-torr ye Co F

p _refere ith,a k,qf_Jg5AMF sgep-An-aMowance -

i; included for on inc+cc:e in F at-reduced-power--based-on the expression:

,N

_,er r,.

n,,, _

(

' AH ~ ' ' " ' ' ' " ' ' ' ' ' "

-Where P is the -fract4en-of-RATED-THERMAL-POWER WOLF CREEK - UNIT 1 B 2-1 Amendment No. 23, 51

Attachment to NA 93-0054 i

Page 5 of 12 m

REACTIVITY CONTROL SYSTEMS ied n Fieure 3. H MODERATOR TEMPERATURE COEFFICIENT % Nm8s Spec LIMITINGCONDITIONFOROPERATION[

~

3.1.1.3 The moderator temperature coefficient (MTC) shall be:

Less positive than 0 fk/k/"r for the all rods withdrawn, beginning a.

of cycle life (BOL), het :cre THERMAL POWER condition,-ee- -

the got Hmit specifiece in the COLR, Less negative thank 1 - 10

• ak/k/*F for the all rods withdrawn, j

b.

end of cycle life (EOL), RATED THERMAL POWER condition.

/

APPLICABILITY Specification 3.1.1.3a. - MODES 1 and 2#*.

Specification 3.1.1.3b. - MODES 1, 2, and 3#.

ACTION:

A A

With the MTC more positive than the limit of Specification 3.1.1.3a a.

above, operation in MODES 1 and 2 may proceed provided:

1.

Control rod withdrawal limits are established and aintained sufficient to restore the MTC to less positive tha 0 i rf "

within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in HOT STANDBY within the next Iio u rs.

I These withdrawal limits shall be in addition to the insertion limits of Specification 3.1.3.6; 2.

The control rods are maintained within the withdrawal limits established above until a subsequent calculation veriffes that the MTC has been restored to within its limit for the all rods withdrawn condition; and 3.

A Special Repo,rt is prepared and submitted to the Commission pursuant to Specification 6.9.2 within 10 days, describing the value of the measured MTC, the interim control red withdrawal l

limits, and the predicted average core burnup necessary for restoring the positive MTC topithin itt ll(t.__for_ tyte_aKR withdrawn condition.

gOL. Spec't.[iec{ rrs -tke Cog

' i. tithe MTC more negatiye than the 1imit 4-f-6 pee 4 Meet 4en :.1.1. :

b.

N bove-be in HOT SHUTDOWN within 1 h

)

i

• With Kgf greater than or equal to 1.

1. See Special Test Exception Specification 3.10.3.

WOLF CREE < - UNIT 1 3/4 1 :

)

i

' Attachmer.t to NA 93-0054 Page 6 of 12

'REA: TIVI W CONTROL SYSTEMS I

SU:XEILLANCE REOUIRE5'ENTc i

4.1.1. 3 The MIC shall be determined to be wit i

'+e eci-ng te fuel cycle as follows:

i C ef zero Te#

specificc0 i^

h-Fig ure 3.H e

g a.

F F (talJ h ured and compa ed to the BOL limit of 50cci' ice-

' -14en 2.1.1. 3a., c.bcve, prior to init1 operation above 5% of RATED

'?EEMAL 6 te ch fuel loading; and C

the MTC shall be measured at any THERMAL POWER and compared to c.

i

.2 e 10 '

.e../"I- (all rocs withdrawn, RATED THERMAL POWER condi-i

' tion) within 7 EFPD after reaching an equilibrium boron concen-tration of 300 ppm.

In the event this comparison indicates the MTC is more negative than y.2.10 ' 2L/k/"F, the MTC shall be remeasured, and compared to the E0L MTC limitgef Jpccificction 2.1.1.:5.-, at least once per 14 EFPD during the remaincer of the fuel cycle.

)

i 1

h, goo vem sucvei knm Lid i

-t h e COLR y 7 ci m e d) m k

I J.

1 2

i.

WOLF CREEK - UNIT 1 3/4 1-5 i

~

Attachment to NA 93-0054 Page 7 of 12 MTC (pcm/ dog. F)

ARO at BOL 8

UNACCEPTABLE OPERATION 6.0, 70%.........

g _

ACCEPTABLE OPERATION r

a _.....................................

i 2

t l

l 1

1 I

I I

1 I

o 0

10 20 30 40 50 60 70 8G 90 100-

% of Rated Thermal Power Figure 3.1-1 (Page 1 of 1)

BOL Moderator Temperature Coefficient vs. Power Level WOLF CREEK - UNIT 1 3/4 1-Sa

l Attachment to FA 93-0054 Page 8 of 12

?

INSERT 1 (Add to page 3/4 2-14) b.

With the RCS total flow rate outside the region of acceptable j

operation shown on Table 3.2-1:

l 1.

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

a.

Restore the total flow rate to within the above limit, or b.

Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER.

2.

Within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> either:

a.

Restore the total flow rate to within the above limit, or b.

Reduce the Power Range Neutron Flux - High Trip Setpoint to less than or equal to 55% of RATED THERMAL POWER.

3.

Within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of initially being outside the above

limit, verify that the RCS total flow rate is restored to within the above limit, or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />; and 4.

Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above

(

the reduced THERMAL POWER limit required by ACTION 1.b and/or 3, above; subsequent POWER' OPERATION may proceed provided that the indicated RCS total flow rate is demonstrated to be within the region of acceptable operation prior to exceeding the following THERMAL POWER levels:

a.

A nominal 50% of RATED THERMAL POWER, b.

A nominal 75% of RATED THERMAL POWER, and c.

Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of attaining greater than or equal to 95% of RATED THERMAL POWER.

Attachment to NA 93-0054 Page,9 of 12 TABLE 3.2-1 DNB PARAMETERS i

LIMITS Four Loops in PARA'4ETER Operation i

1. ndicated Reactor Coolant System T,yg

< 592.5 F q

ndicated Pressurizer Pressure

> 2220 psig*

3. Reactev Cmla.st S sic

• Fimbe 2 38.4 x 109 GPM b

+

/

r i

v r

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

WOLF CREEK - UNIT 1 3/4 2-15

Attachment to NA 93-0054 Page 10 of 12 9e Ett$

2-Abb N ?

• 3*

5

~

p (zy) per 3pecibcaOn ?. 2.3,

r m

Lbe COLK.

(

WhenRCSflowrateandha] remeasured,noadditionalklseft.ncesare, Mprior to ccapartiin WiDi the limits

'h E 3.h 3F N surene P M "o 2.5% for RCS total flow rate and 43(forkhave been allowed for in determinationofthq,deggo,DISRvalue.

g (p (gyj

%nr.ae+ni&y)

The nessurement 4eeer for RCS total rate is bas upon performing a precision heat balan(ce75iPUIT,ng the resu to Salibrate the RCS flow rate indicators.

Potential fouling of the feedwatla wenture which sight not be etected could bias the result from the precisi6n %

lance in a non-to Therefore, an inspection is performed of the feedwater ventur$%.nservative manner.

s t:Weach refueling ostage.

The 12-hour periodic surveillance of indicated RCS flow is sufficient to detect only flow degnda(tan.Jehich40uldJead to operation outside the acceptable region of operation 9:r = 't;= 0.0-3.

This surveillance also provices adequate monitoring toMtect any core Crud build @,

(C 7,4a e, m coa.

3

i Attachment to fiA 93-0054 Page 11 of 12 DESIGN FEATURES i

I 5.6 FUEL STORAGE CRITICALITY i

5.6.1.1 The spent fuel storage racks are designed and shall be maintained i

wi th:

t a.

Ak equivalent to less than o e ual to_D.95 when flooded with etf s

m.

,_unharAted water, which include e cen:crs;tiveAallowance :f 2.0%-

-MtAt-Jor uncertainties as describMITs 4.3 of the M O "

TIis based on new fuel with an enrichment of-4-Wwei ht percent l

JU-235 in Region 1 and on spent fuel with combinatiqn, gitia l

enrichment and discharge exposures, shown in Figur ) 5. C - 1, )i n Region 2, and 3.9 - 1 l

b.

A nominal 9.236 inch center-to-center distance between fuel assemblies.

{

placed in the storage racks.

i 5.6.1.2 The k,ff for new fuel for the first core loading stored dry in the I

spent fuel storage racks shall not exceed 0.98 when aqueous foam moderation is

{

s assumed.

~

i P

DRAINAGE i

P 5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent

+

inadvertent draining of the pool below elevation 2040 feet.

CAPACITY i

5.6.3 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1344 fuel assemblies.

i 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT 5.7.1 The components identified in Table 5.7-1 are designed and'shall be maintained within the cyclic or transient limits of Table 5.7-1.

l i

j:

-l WOLF CREEK - UNIT 1 5-7 Amendment No. 15

{

i s

93-0054 Attachment to NA Page.12 of 12

(. -lh Io s r vn-( Aso ' Te p A c, E.

C_ ORE OPERATING LIMITS REPORTS fCOLR) 6.9.1.9 Core operating limits shall be 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:

l.

Specification 3.1.1.3

Moderator Temperature Cocificient (MTC) EOL

-p limits 2.

Specification 3.1.3.5

Shutdown Rod Insertion Limit 3.

Specification 3.1.3.6

Control Rod Insertion Limits i

4.

Specification 3.2.1

Axial Flux Difference (AFD) l l

S.

Specification 3.2.2.

11 cat Flux Hot Channel Factor - F (X, Y, Z) q 6.

Specification 3.2.3

Nuclear Enthalpy Rise Hot Channel Factor - Fai(X, )

j 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

-i approved by the NRC, specifically those described in the following documents:

l A.

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

B.

NRC Safety Evaluation Report dated

, for the " Transient Analysis -

l Methodology for the Wolf Creek Generating Station" (ET 91-0026, ET 92-0142. WM 93-0010,'

W M 93-0028).

l C.

NRC Safety Evahmtion Report dated

, for the " Steady State Core Physics Methodology for the Wolf Creek Generating Station" (ET 92-0011, WM 93-0038).

i D.

NRC Safety Eva.*nha Report dated

, for the " Reload Safety Evaluation Methoe bg for the Wolf Creek Generating Station" (ET 92-0032, ET 93-0017).

t E.

NRC Safety Evaluation Report dated

, for the " Revision to Technical i

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

,j i

F.

NCR Safety Evaluation Report dated November 13,1986, for "The 1981 Version of the Westinghouse ECCS Evaluation Model Using the basil Code" (WCAP 10266-P-A Rev 2).

l l

The core operating limits shall be determined so that all applicable limits (e.g., fuel thennal-hydraulic -

limits, core thermal-hydraulic limits, ECCS limits, nuc! car limits such as shutdown margin, and transient j

and accident analysis limits) of the safety analysis are met.

The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, l

shall be provided upon issuance, for cach reload cycle, to the NRC Document Control Desk with copics to j

the Reg,ional Administrator and Resident inspector..

J

,, _