Text

Proposed Tech Specs Relocating Certain Numerical Values for Several cycle-specific Core Operating Limits
	ML18009A664
Person / Time
Site:	Harris
Issue date:	09/19/1990
From:	; CAROLINA POWER & LIGHT CO.
To:	;
Shared Package
ML18009A665	List: ML18009A664; ML18022A808;
References
	NUDOCS 9009210259
	Download: ML18009A664 (79)
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ENCLOSURE 5 SHEARON HARRIS NUCLEAR POWER PLANT DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT CYCLE 4 RELOAD SUBMITTAL TECHNICAL SPECIFICATION PAGES 9009210259 900919 POR ADOCK 05000400 PDC

INDEX LIMI;INC CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS SECTION PAGE 3/4 ~ 2 POWER DISTRIBUTION LIMITS 3/4 ~ 2 ~ 1 AXIAL FLUX DIFFERENCE ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 2-L FICURE 3. 2- 1 (DELETED)..............'........,..., 3/4 2-4 3 /4 2 2 HEAQF LUX g T CHANNEL FACTOR (Z) ~ ~ 3I4 2-5 f Q FICURE 3.2-2 fl K(Z) - LOCAL AXIAL PENALTY FUNCTION FOR F (Z) ....... ,3/4 2-8 3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4

~.................,.......,.

2-9'/4 3/4.2.4 QUADRANT POWER TILT RATIO.. ~ . 2-11 3/ 4.2.5 DNB PARAMETERS ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 2-14 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION.. . . ~ ~ ~ ~ ~ e ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ .. 3/4 3-L TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION............ e. o.... 3/4 3-2 T ABLE 3.3-2 (DELETED)......,,..........,..... 3/4 3-9 TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE RKQUIRKMENTS ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 3-11 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION~ ~ ~ ~ ~ ~ e ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ i ~ ~ ~ ~ 3/4 3-16 TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION',.....o........o..o..oo..o..o......... 3/4 3-18 TABLE 3.3-4 KNCINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ . ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 3"28 TABLE 3.3"5 (DELETED)............... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 3-37 TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM h INSTRUMENTATION SURVEILLANCE REQUIREMENTS'......o.oo.. 3/4 3-41 3/4. 3. 3 MONITORING INSTRUMENTATION

~

Radiation Monitoring for PLant Operations ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 3-50 SHEARON HARRIS - UNIT 1 Amendment No. g,J3,1

DEFINITIONS CONTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when'.

a. All penetrations required to be closed during accident conditions are either:

1. Capable of being closed by an OPERABLE containment automatic isolation valve system, or 2., Closed vaLves by manual secured valves, blind flanges, or deactivated automatic i their closed positions, except as provided in able 3.6-1 o Specification 3.6.3. I

b. All equipment, hatches are closed and sealed, C ~ Each air lock is in compliance with the requirements of Specification 3.6.1.3,

d. The containment leakage rates are within the limits of Specification 3.6.1.2, and

e. The sealing mechanism associated with each penetration (e.g., welds, bellows, or 0-rings) is OPERABLE.

CONTROLLED LEAKAGE 1.8 CONTROLLED LEAKAGE shall, be that seal water flow supplied to the reactor coolant pump seals.

CORE ALTERATION 1.9 CORE ALTERATION shall be the movement or manipulation of any component within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position.

CORE OPERATING LIMITS REPORT 1.9.a The CORE OPERATING LIMITS REPORT is the unit-specific document that provides core o'perating limits for the current operating reload cycle. These cycle"specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9.1.6. Plant operation within these core operating limits is addressed within the individual specifications, DIGITAL CHANNEL OPERATIONAL TEST 1.10 A DIGITAL CHANNEL OPERATIONAL TEST shall consist of exercising the digi-tal computer h'ardware using data base manipulation to verify OPERABILITY of alarm andlor trip functions.

SHEARON HARRIS " UNIT 1 1-2 Amendment No. p

2.1 SAFETY LIMITS BASKS 2.1.1 REACTOR -CORE (Continued)

These curves are'based on an enthalpy h t channel factor, F , o g egAedin 4H H

Euel and 1.65 Eor VANTAGE 5 fuel and a reference cosine vith a peak of

'OPAR ppg~vIH6 pg f/'g&8T 1.55 Eor axial power shape. An allowance is included for an increase in (NLR) calculated F4H at reduced over based on the expression:

F'F~

bH F4H . [L + 0. (1-P)] for LOPAR fuel, and F4H

~ 1.65 [L + 0,35 (L-P)] Eor VANTAGE 5 fuel Where P is the Eraction of RATED THERMAL POWK >

These limiting heat flux conditions are higher than those calculated Eor the range oE aLL control rods fully vithdravn to the maximum alLovable control rod insertion assuming the axial power imbalance is vithin the Limits of the fl (4I) function of the Overtemperature trip. When the axial power imbalance is not vithin the toLerance, the axial power imbalance effect on the Over-temperature 4T trips will reduce the Setpoints to provide protection consistent with core Safety Limits.

F " = F limH uf 8A'Tfb THERMAL POWER SpegiAed- fn+edblR, I'd pF = pzz<> Fmfor Wd<~pt<~ Pot'~< spciAed inde QOI-R.

SHKARON HARRIS - UNIT 1 8 2"la Amendment No. 15

REACTIUITY CONTROL SYSTEMS ~oin4oinc J willin +c H~+ spcciAc~ '" +c MODERATOR TEMPERATURE COEFFICIENT PIP or eggs~ ~

LIMITING CONDITION FOR OPERATION 3.L,L.3 The moderator temperature coefficient (MTC) shall be

a. Less osxtxve tha +5 pcm/'F for power Levels up to 70Z RATED THERMAL POWER and a Linear ramp from that point to 0 pcm/'F at 100Z RATED THERMAL POWER'n

b. Less negative than -49 pcm/'F for all rods withdrawn, end of c cle Life (EOL), RATED THERMAL POWER condition.

PositivE'RTe umit APPLICABILITY: pecxfication 3.1.1.3a. - MODES 1 and 2'nly~.

S ecification 3.1 '.3b. - MODES 1, 2, and 3 only~.

ge~IIVC tvtTC l.i~~+

ACTION:

p,slflve i4rC LirnH speci &cd in We C.OLR With the MTC more positive than the limit of S ecification 3.1.1.3a.

Iabo~ve operation in MODES 1 and 2 may proceed provided:

1. Control rod ~ithdra~al limits are estabLished and maintained sufficient to restore the MTC to within, the ve 1 within 24 hour~or be in HOT STANDBY within the next 6 i withdrawal Limits shall be limits of Specification 3.1.3.6; in addition to the insertion hours.'hese

2. The controL rods are maintained within the ~ithdra~al Limits established above until a subsequent calculation verifies that the MTC has been restored to within its limit for'the alL rods withdrawn condition', and

3. A Special Report 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 rod withdrawal limits, and the predicted average core burnup necessary for restoring the positive MTC to within its Limit for the aLL rods withdrawn condition. i +s gcgativc MTC t.imi

'c inde Spcccr6cd in COLR>

With the MTC more negative than the imit of S ecification 3.1 ~ 1.3b.

bove e in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

With keff greater than or equal to 1.

~See Special Test Exceptions Specification 3.10.3.

~ ~ ~ ~ ~

~

SHEARON HARRIS - UNIT 1 3/4 1-4 Amendment No. 8

REACTIVITY CONTROL SYSTEMS SURVEILLANCE RE UIREMENTS 4.1.1.3 The MTC shall be determined to be within its limits during each fuel cycle as follows: t pas'ave MTC limit5pcciRd in AC CON, plank praulure a ~ The MTC shall be measured nd compared to the xt cation . . . a. a ove prior to initial operation above SZ of RATED THERMAL POWE , after each fuel loading, and

b. The MTC shall e measured at any'HERMAL POWER and corn ared to

@c,3MPQ i'~i

. -41.5 cm F all rods wxt rawn, RATED THERMAL POWER condition) within

$ g pyeigancc spec(F'icginAc, ca 7 EFPD after reaching an equx x rxum oron concentration of 300 ppm.

n the event th's comparison indicates the MTC is more negative th n AcsOo PPw. cm/'F the MTC shall be remeasured, and compared to the 0 ~

5ur vci4ncc, liini4 Spcifi'e8 in &e ceca MTC limit of S ecification 3.1 1 'b. at

~ least once per 14 EFPD Hepta 8+".g<~

during the remainder of the fuel cycle. i~ eccccc, SHEARON HARRIS - UNIT' Amendment No. g

REACTIVITY CONTROL SYSTEMS 3/4.1.3 MOVABLE CONTROL ASSEMBLIES GROUP HEIGHT LIMITING CONDITION FOR OPERATION 3.1.3.1 All shutdown and control rods shall be OPERABLE.and positioned within

+ 12 steps (indicated position) of their group step counter demand position.

APPLICABILITY: MODES 1<< and 2<<.

ACTION:

a. With one or more rods inoperable due to being immovable as a result of excessive friction or mechanical. interference or known to be untrippable, determine that the SHUTDOWN MARGIN requirement of Specification 3.1 ~ 1.1 is satisfied within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and be in HOT STANDBY within 6 hours.

b. With more than one rod misaligned from the group step counter demand position by more than + 12 steps (indicated position), be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c. With more than one rod inoperable, due to a rod control urgent failure alarm or obvious eLectricaL problem in the rod control system existing for greater than 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months , be in HOT STANDBY within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

d. With one 'rod trippable.but inoperable due to causes other than addressed by ACTION a., above, or misaligned from its group step counter demand heightimmit + 12 steps (indicated position),

orth' than by more POWER OPERATION may continue provided that within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .'.

The rod is restored to OPERABLE status within the above alignment requirements, or

2. The rod is declared inoperable and the remainder of the rods in the group with the inoperable rod are aligned to within + 12 ste s of the ino erable rod while maintaining the rod sequence s i >4 and insert con 1 n cont ro an t e ore g 5pgg[f)~Ho~

0 eratin Limits Re s

e an s specs ae eve s a e restricted pursuant to Specification 3.1.3.6 during subsequent operation, or

3. The rod is declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied. POWER OPERATION may then continue provided that:

a) A reevaluation of e'ach accident analysis of Table 3.1-1 is performed within $ days; this reevaluation shall confirm that the previously analyxed results of these accidents

<<See Special Test Exceptions Specifications 3.10.2 and 3.10.3.

SHEARQN HARRIS - UNIT 1 3/4 1-14 Amendment No. j, P5

~ g REACTIVITY CONTROL SYSTEMS SHUTDOWN ROD INSERTION LIMIT LIMITINC CONDITION FOR OPERATION 3.1.3.5 ALL shutdown rods shall be full withdrawn as s ecified in the or 0 eratin Limits Re ort ~ GARE OpSRATINQ g.(I4)~ QEpgRT (col.R)> lang proudue pu-Io@

APPLICABILITY: MODES 1* and 2* ~.

ACTION:

as spccigcdih +c COLR With a maximum of one shutdown rod not fully withdrawn, except Eor surveillance testing pursuant to Specification 4.1.3.1.2, within 1 hour either'.

a. FulLy withdraw the rod, or

b. Declare the rod to be inoperable and apply Specification 3.1,3.1.

SURVEILLANCE RE UIREMENTS as speciFiad ln Ac, 4.1.3.5 shutdown rod shall be determined to be fu y withdra Doc,a'ach Within 15 minutes prior to withdrawal of any rods in Control Bank A, B, C, or D during an approach to reactor criticality, and r

b. At least once per 12 hours thereafter.

""See Special. Test Exceptions Specifications 3.10.2 and 3.10.3.

- With Keff greater than or equal to l.

SHEARON BARRIS - UNIT 1 3/4 1-20 Amendment No. ]8

REACTIVITY CONTROL SYSTEMS CONTROL ROD INSERTION LIMITS LIMITINC CONDITION FOR OPERATION 3.1.3.6 The control banks shall be Li ited in physicaL insertion as specified in the 0 eratxn Li CORE OPERA'fIN'IHlOS'EPOM (~LR)q pl arrl pf'Ocedw'+ P APPLICABILITY: MODES 1* and 2>> '"".

ACTION:

spcciGcd in 4c doLR With the control banks inserted beyond the bov insertion limit except for surveillance testing pursuant to Specification 4.1.3.1.2:

inse.a]ion y spec ice'ed in kc Co<"

a. Restore the control banks to within the Limit) within 2 hours, or

b. Reduce THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 'o less than or equal to that fraction of RATED THERMAL POWER which is allowed b the bank posi-tion using the insertion limits specified in the Core Operating Limits eport, or QOi.P,

c. Be in at Least HOT STANDBY within 6 hours.

SURVEILLANCE REQUIREMENTS 4.1.3.6 The position of each controL bank shall be determined to be within the inserti'on Limit t least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the rod insertion limit monitor is inoperable, then verify the individuaL rod positions at least once er 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

s~~i.j~ in a LR

<See Special Test Exceptions Specifications 3.10m2 and 3.10.3.

"-'-With Keff greater than or equal to 1.

SHEARON HARRIS - UNIT 1 3/4 1-21 Amendment Ne. 7, g

FIGURE 3.1-2 DELETED.

ROD 6ROUI XNSERT10lU L IVjlT8 VER$ QS THERMALP4IVER THREE LOOP DPERATlOhl This &pure ig deleted from Technical SpeciFicoh'one, cond is oon~roPcd by the QOFlEOPERAYIQG LIMLW REPORT,plan+ procedure. PLF'-t~>.

SHEARON HARRIS - UNIT 1 3/4 1-22 Anendnent No.

I

~

3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE LIMITINC CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be m '.ned vithi p-I p4 Qpg& ppeRAAhlci LIB>rs REpoRT colR), la~+ procccKi-c I

a. the acce table o erational space as specified in t e o i imits Re or for Relaxed Axial Offset Control (RAOC) operation, or !

b. within a band about the target AFD during Base L ad operation as specified in the Core 0 eratan Laments Re ort. I C,OLR APPLICABILITY: MODE- 1 a ove 50X of RATED THERMAL POWER<.

ACTION:

a ~ For RAOC operation with the indicated AFD outside of the Limits specified in the ore eratxn xmxts e or , either'.

OL,g.

1. Restore the indicated AFD to vithin the limits specified in the Core 0 eratin Limits Re or't within 15 minutes, or CO

2. Reduce THERMAL POWER to less than 50X of RATED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux-High Trip setpoints to Less than or equaL to 55X of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b. For Base Load operation above APL with the indicated AXIAL FLUX DIFFERENCE outside of the a Licable tar et band about the target AFD, either'.

1. Restore the indicated AFD to vithin the target band limits vithin 15 minutes, or

2. Reduce THERMAL POWER to less than APL of RATED THERMAL POWER and discontinue Base Load operation within 30 minutes.

C ~ THERMAL POWER shall not be increased above 50X of RATED THERMAL POWER unLess the indicated AFD 's vithin the Limits specified in the Core 0 eratxn Laments Re ort for RAOC operation.

dot.R.

""See Special Test Exception 3.10.2 T'r*APLND iS th minimum aLLowable over level for Base Load o eration and b v xn the 0 er tin Limits Re ort er S ecification 6. .1.6.

is s~ikccI A OAA SHEARON HARRIS - UNIT 1 3/4 2-1 Amendment No. g, g

FICURE 3.2"1 DELETED AXIAL FLHX DIFFERENCE LI I 8 FuHQTION OF RA'%b 7IIEWAL Pb'lnlERFOR AK I

This figure is dcldcd 4'rom Techniml pccit'icah'one and 4 GOnfrollcd b'/.Hc QoIIE, oPERATlhl6 I-sfITs REPORT, Plank Proccdurc, PLP-lM.

SHEARON HARRIS - UNIT 1 3/4 2-4 Amendment Ne. l7, Pf

POWER DISTR1BUTION LIMITS 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR F (Z)

=LIMITING CONDITION FOR OPERATION 3.2.2 F~(Z) shal Limited by the foLlowing relationships:

RTP F x FO(Z) < ~2.4 (Z)N FOR P > 0 ~ 5 P

5k )C FO(Z) < (4.90) (Z)g FOR P 4 0.5 fgP 0'%b THERhQL POWER SPCjAtd i<+< <OREOPQAAj(AfCj Where: +C F@ liFF)i+ at Lt+f75 QEPORT (coLR)p plan4 pF Dcedgrc. PLP- J 04>

p THERMAL POWER , and RATED THERMA P WE normalize( F x) as a fz)F5<<'oF) 07 K(Z) = the -2 0 ve core height ocatio spccificd in++ CON APPLICABILITY: MODE 1 ~

ACTION:

With F~(Z) exceeding its limit-'

~ Reduce THERMAL POWER at least 1X for each LX F((Z) exceeds the limit within 15 minutes and similarLy reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-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 'ubsequent POWER OPERATION may proceed provided the Overpower hT Trip Setpoints have been reduced at Least lX for each LZ F~(Z) exceeds the Limit.

b. Identify and correct the cause of the out-of-Limit condition prior to increasing THERMAL POWER above the reduced limit re" quired by ACTION a49 above; THERMAL POWER may then be increased provided F~(Z) is demonstrated through incore mapping to be within its limit.

SHEARON HARRIS - UNIT 1 3/4 2-S Amendment No. )F, +

POWER DISTRIBUTION LIMITS SURVEILLANCE RE UIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

4.2.2.2 For RAOC operation, F~(Z) shall be evaluated to determine if it is within its limit by'.

a. Using the movable incore detectors to obtain a power distribution map at any THERMAL POWER greater than 5Z of RATED THERMAL POWER.

b. Increasing the measured Fz(Z) component of the power distribution map by 3X to account for ihanufacturing tolerances and further increasing the valu'e by 5X to account for measurement uncertainties. Verify the requirements of Specification 3.2.2 are satisfied.

C ~ Satisfying the lowing relationship: '@

F (Z) < 4 x K(Z) for P > 0.5 Fq F (Z) < ~9 RH x K(Z) for P < 0.5 RTP

~Q where F<(Z) is the measured F (Z) increased by the allowan es for gyral>xc J F~(>> manufacturing tolerances and measurement uncertainty, .4 is the F~

as a AnchD0 OF core. h~g"~ limit, K Z i iven in Fi ure 3.2-, is the fraction of RATED THERMAL POWER, and W(Z) is the cycle dependent function that accounts for ower distribution transients encountered durin normal F" K(z) c'~J W( 0 o eration. This function is given in the Core 0 eratin L 't ore spcificdin+eS ~ =

Re ort as er S ecification 6.9.1.6.

d. Measuring- F (Z) according to the following schedule:

1 ~ Upon"achieving equilibrium conditions after exceeding by 10X or more of RATED THERMAL POWER, the THERMAL POWER at which F~(Z) was last determined,* or

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

"During power escalation at the beginning of each cycle, power level may be increased until a power level for extended operation has been achieved and a power distribution map obtained.

SHEARON HARRIS - UNIT 1 3/4 2-6 Amendmant No. 1 ~

POWER DISTRIBUTION LIMITS SURVEILLANCE RE UIREMENTS (Continued)

e. With measurements indicating FM maximum KKKz has increased since the previous determination of FQ (Z) either of the following actions shall be taken'.

1

1) FQ (Z) shell be increased by 2X over that specified in Specification 4.2.2.2c, or

2) F (Z) shall be measured at least once per 7 Effective FulL Q

Power Days until two successive maps indicate that FM maximum Q (Z) is not increasing.

KKKz

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

1) Calculate the percent FQ(z) exceeds its limit by the following expression:

F (Z) x W(Z) maximum x 100 for P R 0.5 FqRTP x K(z) I F (Z) x W(Z) maxim x 100 for P < 0.5 F T x K(Z) 5 2} One of the following actions shall be taken'.

a) Within 15 minutes, controL the AFD to within new AFD limits which are determined b reducin the AFD Limits specified in the o e 0 eratin Limi s R or by LZ AFD COUR for each percent FQ exceeds its limits as determined in Specification 4.2.2.2f.l). Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , reset the AFD alarm setpoints to these modified limits, or b) Comply with the requirements of Specification 3 '.2 for F (Z) exceeding its limit by the percent calculated above, 0

c) Verify that the requirements of Specification 4.2.2.3 for Base Load operation. are satisfied and enter Base Load operation.

SHEARON HARRIS - UNIT 1 3/4 2-7a Amendsent No. 7 PS ~

POWER DISTRIBUTION LIMITS SURVEILLANCE R UIREMENTS (Continued)

g. The limits specified in Specifications 4.2.2.2c, 4.2.2.2e, and 4.2.2.2f above are not applicable in the following core'lane regions:

1 ~ Lower core region Erom 0 to 15X, inclusive.

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

4.2.2.3 Base Load operation is permitted at powers above APL if the following conditions are satisfied:

a. Prior to entering Base Load operation, maintain THERMAL POWER above APL and less than or equal to that allowed by Specification 4.2.2.2 for at least the 'previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 'aintain Base Load operation surveillance (AFD within the limits specified in the Core Operating Limits Report) during this time period. Base Load operation is then permitted providing THERMAL POWER is maintained between APL and APL or between APL and 100X (whichever is most limiting) and Fg surveillance is maintained pursuant to Specification 4. 2.4. APL " is defined as:

sTP F

APL BL ~

minimum x K(Z) x 100Z

[ ]

F (Z) x W(Z) eg'~ is +.

where! F~(Z) is the measured F~(Z) increased by th allowances for

~e porn alizeJ FqC>)

manufacturing tolerances and measurement u certainty. he F~ limit as a funehon oE core height, and dependent function that accounts Eor limited power distribution transients encountered durin Base Load o eration. The function is P~~, y(z), on~ given in the Core Operatin Limits Re ort as er W(z)z< a' S ecification 6.9.1.6.

ih+

5tMcificd bo Durj~g Base Load operation, if the THERMAL POWER is decreased below APL then the conditions of 4.2.2.3.a shall be satisfied before re-entering Base Load operation.

4.2.2.4 During Base Load operation F~(Z) shall be evaluated to determine if it is within its limit by'.

a~ Using the movable incore detectorsgo obtain a power distribution map at any THERMAL POWER above APL bo Increasing the measured F~(Z) component of the power distribution map by 3X to account Eor manufacturing tolerances and further increasing the value by 5Z to account for measurement uncertainties. Verify the requirements of Specification 3.2.2 are satisfied.

SHEARON HARRIS - UNIT 1 3/4 2-7b Amendment No. 1 ~ 9

~s POWER DISTRIBUTION LIMITS SURVEILLANCE RE UIREMENTS (Conti,nued)

C ~ Satisfy'he P following relationship'.

F K(Z) RTP.

FM(Z) " ~

for P > APLN Fq is%he. Fq liPlit~

Q P x W Z)BL where: M FQ(Z) is the measured FQ(Z) The F limit is 2.45.

~e poronalixed Fg(~)

as a /+dion oF cow "<'P s K(Z)

POWER"

~ iven in Fi ure 3.2" ~ P is the fraction of RATED THERMAL W(Z)BL is the cycle dependent function that accounts for F,R~P limited pouer distribution transients encountered durin orms K(x))o~nd W'C4SL operation. Thas unction as given an the Core Operating Limits ad~ specified in the Cot R Re ort as er S ecification 6.9.1.6.

d ~ Measuring F'(Z) in conjunction with target flux difference determination according to the following schedule:

1. Prior to enteri.n 8 Base Load o P eration af t e r sat i sfyang Section 4.2.2.3 unless a fuLL core flux map has been taken in the previous 31 EFPD with the reLative thermaL power having been maintained above APLND for the 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to mapping, and 2 ~ At. Least once per 31 effective full power days.

e. With measurements indicating maximum (~i F (Z) has increased since the previous determination F'Z)

Q either of the foLLowing actions shaLL be taken:

M FQ(Z) shaLl be increased by 2 percent over that specified in 4.2.2.4.c; or

2. F Q

(Z) shall be measured at Least once per 7 EFPD until 2 successive maps indicate that (z) maximum -is not increasing.

i~Kz> i With the relationship specified in 4.2.2.4.c above not being satisfied, either of the following actions shalL be taken.'.

Place the core in an equilibrium condition where the Limit in 4.2.2.2.c is satisfied, and remeasure FQ(Z), or SHEARON HARRIS - UNIT 1 3/4 2-7c Amendment No, 7

POWER DISTRIBUTION LIMITS SURVEILLANCE RE IREMENTS (Continued)

2. Comply with the requirements of Specification 3.2.2 for F~(Z) exceeding its limit by the percent calculated with the following expression:

(z) x w(z)

[(max f [ J ) -1] x 100 for P? APL F nt x K(z)-

The limits specified in 4.2.2.4.c, 4.2.2.4.e, and 4.2.2.4.f above are not applicable in the following core plane regions'.

1. Lower core region 0 to 15 percent, inclusive.

2. Upper core region 85 to 100 percent, inclusive.

4.2.2.5 When F~(Z) is measured for reasons other than meeting the requirements of Specification 4.2.2.2 an overall measured F (Z) shall be obtained from a power distribution map and increased by 3Z 3o account for manufacturing tolerances and further increased by 5X to account for measurement uncertainty.

SHEARON HARRIS - UNIT 1, 3/4 2-7d Amendment No. 7, +

4 I MRK E18fhllON (FED)

FIGURF 3.2-2 (Z)

K(Z) LOCAL AXlAL PENALTY FUNCTION FOR Fa

FIGURE 3.2-2 K(Z) THE NORMALIZED (Z) AS A FUNCTION OF CORE HEIGHT FQ This figure is deleted from Technical Specifications and is controlled by the CORE OPERATING LIMITS REPORT, plant procedure PLP-106.

SHEARON HARRIS - UNIT 1 3/4 2-8 AMENDMENT NO. l)

POWER DISTRIBUTION LIMITS 3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR LIMITING CONDITION FOR OPERATION 3.2.3 The in ic Reactor Coolant System (RCS) total flow rate and F<H shall be maintained as fol ows'.

a. re

~l flo ate Meas ggP RCS PF~

> 293,540 gpm x (1.0 + C 1 ), and 8H be F4H ~ 6 [1 ~ 0 + 0 le0 P)] fOr LOPAR fu F>H < 1.65 (1.0 + 0.35(1.0-P)) for VANTAGE 5 fuel.

lhere:

P ~ THERMAL POWER , an RATED THERMAL POWER F<H

~ nthalpy rise hot channel factor obtained by using the movable incore detectors to obtain a power distribution map, with the measured value of the nuclear enthalpy rise hot N

channel factor (F>H) increased by an allowance of 4Z to attount for measurement untertaintyfy~s oad Cl ~ Measurement uncertainty for core flow as described in the Bases.

APPLICABILITY: MODE 1 ~

ACTION:

'With RCS total flow rate or F<H outside the above limits:

CL ~ Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either!

Restore RCS total flow rate and F<H to within the above limits, or 2~ Reduce THERMAL POWER to less than 50Z of RATED THERMAL POWER and reduce the Power Range Neutron Flux - High Trip Setpoint to less than or equal to 55Z of RATED THERMAL POWER within the next 4 hours.

F RTP F4H limit at RATED THERMAL POWER specified in the CORE OPERATING LIMITS REPORT (COLR), plant procedure PLP-106, PF~H - Power Factor Multiplier for Fz,H specified in the COLR, SHEARON HARRIS - UN[ f 1 3/4 2-9 Amendment No. g> Q

POWER DISTRIBUTION LIMITS LIMITING CONDITION FQR OPERATION ACTION Continued:

b. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of initially being outside the above limits, v rify through incore flux mapping and RCS total flow rate o arioso that

~~ and RCS total flow rate are restored to within the above liaits, or reduce THERMAL PQWFR to less than 5" of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

C. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced THERMAL PCWER limit required by ACTION a.2. and/o b., above; subsequent POWER OPERATION may proceed provided that and indicated RCS total flow rate are JCggf ffi<fiQ demonstrat d through incore flux mapping and RCS total flow rate o , to be ~ithin acceptable limits prior to exceeding the following THERMAL POWER levels:

1. A nominal 50% of RATED THERMAL POWER,

2. A nominal 75 of RATED THERMAL POWER, and

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

SURVEILLANCE RE UIREMENTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicab'Ie.

4.2.3.2 F~ shall be datensined to be within accaptab'li liaits

a. Prior to operation above 75" of RATED THERMAL POWER after each fuel loading, and

b. At least once per 31 Effective Full Power Days.

4.2.3.3 The i i d RCS total flow rate shall be verified to be within the I acceptable limit:

a. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by the use of main control board instru-mentation or equivalent, and

b. At least once per 31 days by the use of process computer readings or digital voltmeter measurement.

4.2.3.4 The RCS total flow rate indicators shall be subjected to a CHANNEL CALIBRATION at least once per 18 months.

4.2.3.5 The RCS total flow rate shall be determined by precision heat balance-measurement at least once per 18 months. The measurement instrumentation shall be calibrated within 7 days prior to the performance of the c-lorimetric flow measurement.

SHEARON HARRIS - UNIT 1 3/4 2-10

I

~ ~

3/4. 3 INSTRUMENTATION 3/4.3. 1 REACTOR TRIP SYSTEM INSTRUMENTATION LIMITING CONOITION FOR OPERATION 3.3.1 As a minimum, the Reactor Trip System inst~umentation channels-.and interlocks of Table 3.3-1 shall be OPERABLE.

APPLICABILITY: As shown in I'able 3.3-1.

ACTION: As sbnsn ln Teble 3.3-1.

SURVEILLANCE RE UIREMENTS 4.3.1.1 Each Reactor Trip System instrumentation channel and interlock and the automatic trip logic shall be. demonstrated OPERABLE by the performance of the Reactor Trip Sys r entati u I e Re uirements specified in Tabl e 4. 3-1. SpcciAed in &c rcc le

prOCedurC, PLtb-INn>

4,3.1.2 The REACTOR TRIP SYSTEM RESPONSE TIME of each Reactor trip function shall be demonstrated to be within its Iimi t least once per 18 months. Each test shall include at least one train such t at both trains are tested at least once per 36 months and one channel per function such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific Reactor trip function as shown in the "Total No. of Channels" column of Table 3.3-1.

SHEARON HARRIS .- UNIT 1 3/4 3-1

TABLE 3.3-2 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES This table is deleted from Technical Specifications.

The info ation in this table is controlled by plant protedure P1P-10f6y~

Technical Specxfxcatxon quxpment xst rogram>

PAGE 3/4 3-10 HAS BEEN DELETED.

SHEARON HARRIS - UNIT 1 3/4 3-9 Amendment No.

INSTRUMENTATION ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE RE UIREMEHTS 4.3.2.1 Each ESFAS instrumentation channel and interlock and the automatic actuation logic and relays shall be demonstrated OpERASLE by performance of the ESFAS Instrumentation Surveillance equirements specified in Table 4.3-2.

Ne 4.3.2.2 The ENGINEERED SAFETY FEA RES RESPONSE TIME of each ESFAS function shall be demonstrated to be within limit at least once per IS months. Each I test shall include at least one train such that both rains are tested at least nnce per 36 months and one channel per function such that all channels are tested at least once per N times 18 months where N is the tota> numb>> of redundant channels in a specific ESFAS function s shown in the "Total Ho. of Channels" column of Table 3.3-3.

gpqqjpJ in +a Tcchnied Sfcciiien+'>n Q~'P p

tank p f'oc58w c lPL ~

pygmy~~

SHEARON HARRIS - UNIT 1 3/4 3-L7

TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES This table is deleted from Technical Specifications.

The informati n ' is ta is controlled b plant procedure PLP-106f Technical Specification Equipment List Program>

PACES 3/4 3-38 THROUGH 3/4 3-40.HAVE BEEN DELETED.

SHEARON HARR I S - UNIT I 3/4 3-37 Amendment Nn. ]If

~~

REACTOR COOLANT SYSTEM REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.9.2 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the Limit Lines shown on Figures 3.4-2 and 3.4-3 during heatup, cooldown, and inservice leak and hydrostatic testing with:

a. ~ A maximum heatup rate as shown on Table 4.4-6.
b. A maximum cooldown rate as shown on TabLe 4.4-6.
c. A maximum temperature change of less than or equal to 10'F in any 1-hour period during inservice hydrostatic and leak testing opera-tions above the heatup and cooldown limit curves.
APPLICABILITY: MODES 4, 5, and 6 with reactor vessel head on.
ACTION:
With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes', if the pressure and temperature limit lines shown on Figures 3.4"2 and 3.4"3 were exceeded, perform an engineering evaluation to determine the effects of the out"of-Limit condition on the structural integrity of the Reactor Coolant System; .determine that the Reactor Coolant System remains acceptable for continued operation or maintain the RCS Tag and pressure at less, than 200'F and 500 psig, respectively..
SURVEILLANCE RE UIREMENTS 4.4.9.2.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operations'Nt c
4.4.9.2.2 Deleted fro echnic 1 ec'cati ns. er to Plant Procedure PLP-106+ echnica Specification Equipment List Programs SHEARON HARRIS - UNIT 1 3/4 4-34 Amendment No. N
TABLE 4e4 S REACTOR VESSEL HATERIAL SURVEILLANCE PROCRAM This table is deleted from Technical Specifications. ~
~ ~ ~ ~ ~ ~
e in ormation
~ ~
'his
~ ~
List
~
e ble is controlle by plant procedure PLP-106 Technica 5gchnica'pecification Equipment Program>
SHEAROM HARRIS - UNIT 1 3/4 4-37 haendeenr. Ho. ~
CONTAINMENT SYSTEHS gpniFil,d in +t'. T8+~i+ ~p~~i~~
3/4. 6. 3 CONTAINMENT ISOLATION VALVES E)~'prnerd /iN pry'~ p>"n~ P~e~"
PLP-IOl, LIMITING CONDITION FOR OPERATION 3.6.3 Each containment isolation valve shall be OPERABLE with isolation times less than or equal to required isolation times'.
APPLICABILITY: NODES 1, 2, 3, and 4.
ACTION:
With one or more of the containment isolation valve(s) inoperable, maintain at least one isolation valve OPERABLE in-each affected penetration that is open and:
a. Restore the inoperable valve(s) to OPERABLE status within 4 hou~s,-
or
b. Isolate each affected penetration within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least one deactivated automatic valve secured in the isolation position, or
c. Isolate each affected penetration within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least one closed manual valve or blind flange, or
d. Be in at least HOT STANDBY within the next 6 hours and in COLO SHUTOOW within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
SURVEILLANCE RE UIREHENTS 4.6.3.1 Each isolation valve shall be demonstrated OPERABLE prior to returning the valve to service after maintenance, repair or r'eplacement work is performed on the valve or its associated actuator, control or po~er circuit by performance of a cycling test, and verification of isolation time.
SHEARON HARRIS - UNIT 1 3/4 6-14
CONTAINMENT SYSTEMS CONTAINMENT ISOLATION VALVES SURVEILLANCE RE UIREMENTS (Continued) 4.6.3.2 Each isolation valve shall be demonstrated OPERABLE during the COLD SHUTDOMN or REFUELINC MODE at least once per 18 months by'.
a. Verifying that on a Phase "A" Isolation test signal, each Phase "A" isolation valve actuates to its isolation position',
b. Verifying that on a Phase "B" Isolation test signal, each Phase "B" isolation valve actuates to its isolation position; and
c. Verifying that on a Containment Ventilation Isolation test signal, each normal, preentry purge makeup and exhaust, and containment vacuum relief valve actuates to its isolation position, and
d. Verifying that, on a Safety Injection "S" test signal,, each containment isolation valve receiving an "S" signal actuates to its isolation position, and
e. Verifying that, on a Hain Steam Isolation test signal, each main steam isolation valve actuates to its isolation position, and
f. Verifying that, on a Hain Feedwater Isolation test signal, each feedwater isolation valve actuates to its isolation position.
4.6.3.3 The isolation time of each power operated or automatic valve shall be determined to be within its limit when tested pursuant to Specification 4iO F 5
'ppp> QQ >'f) +Q 7echn'ical Specie cab on Eq'uipnvn'> Lis" p,,~ pj',<+ pwcd~e. Rl'-<<4, SHEARON HARRIS - UNIT 1 3/4 6-15 Amendment No ~
TABLE 3.6-1 CONTAINMENT ISOLATION VALVES This table is deleted from Technical Specif
~ ~
tions.
~
~
The
~
o io i
~
is bl is controlled
~ ~
by plant procedure PLP-106 Technical
~~thn'Ca~ Specification Equipment List Program>
PAGES 3/4 6-17 THROUGH 3/4 6-29 HAVE BEEN DELETED.
SHEARON HARRIS - UNIT 1 3/4. 6-16 Amendment No. M
PLANT SYSTEMS 3/4.7.8
~ ~ SNUBBERS LIMITING'ONDITION FOR OPERATION 3.7.8 All snubber s shall be OPERABLE. The only snubbers excluded firom the requirements are those installed on nonsafety-related systems and then only if their failure or failure of the system on which they are installed would have no adverse effect on any safety-related system.
APPLICABILITY: MODES 1, 2, 3, and 4. MODES 5 and 6 for snubbers located on systems required OPERABLE in those MODES.
ACTION:
With one or more snubbers inoperable on any system, within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months replace or restore the inoperable snubber(s) to OPERABLE status and perform an engineering evaluation per the augmented inservice inspection program on the attached component or declare the attached system inoperable and follow the appropriate ACTION statement for that system.
SURVEILLANCE RE UIREMENTS 4.7.8 Each snubber shall be demonstrate BLE b erformance of the
~ T~<<~i
~ ~ P inser vice inspection progr . dpecigicd I'n
~
augmented cpsciAcak'nn Eg~~ Lisk prgr~, plan4 procedur< PLP-IQQ, PaCeS Zle V-ZO VaiinuSII Sl< T-Za WAVe gd SHEARON HARRIS - UNIT 1 3/4 7-19
V~cETE THIS PAGE e inservice inspection program for SNUBBERS is deleted from Technic Sp ifications and is controlled in plant procedure PLP-106, Technic Spe 'fication Equipment List Program.
PAGES 3/4 1 THROUGH 3/4 7-23 E BEEN DELETED.
SHEARON HARRIS - UNIT 1 3/4 7-20 Amendment No. l8
FICURE 4 ~ 7 1 SAHPLE PLAN(2) FOR SNUBBER FUNCTIONAL TEST byfhe This figure is del ~ om echn'c'fic procedure PLP-106 Technical Specification i 's co rolled n plant quipment List Program~
Q !
SHEARON HARRIS UNIT 1 3/4 7-24 Amendment No.
ELECTRICAL POSER SYSTEMS 3/4.8.4 ELECTRICAL E UIPMENT PROTECTIVE DEVICES CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES LIMITING CONDITION FOR OPERATION 3.8.4.1 Each containment etr ti c n u urren rot ive device shall be OPERABLE. speci icd iAAt. TechniccIf QptcIQI~'~ Egcup~ J july prygryrn>
pi~+ proce4m pgp ~gg APPLICABILITY: MODES 1, 2, 3, an ACTION:
14ith one or more of the containment penetration conductor overcurrent protective device(s) inoperable:
a~ Restore the protective device(s) to OPERABLE status or deenergize the circuit(s) by tripping the associated backup circuit breaker or racking out or removing the inoperable circuit breaker within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , declare the affected system or component inoperable, and verify the backup circuit breaker to be tripped or the inoper able circuit breaker racked out or removed at least once per 7 days thereafter; the provisions of Specification 3.0.4 are not applicable to overcurrent devices fn circuits which have their backup circuit breakers tripped, their inoperable circuit breakers racked out or removed, or
b. Be in at least HOT STANDBY within the next 6 hours and in COLD .
SHUTlXNN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
SURVEILLANCE RE UIREMENTS 4.&.4. 1 Each containment penetration conductor overcurrent'protective devices shall be demonstrated OPERABLE:
aO At least oned per 33 months:
1. By verifying that the 6900-volt circuit breakers are OPERABLE by selecting, on a rotating basis, at least 10K of the circuit breakers, and performing the following:
a} A CHANNEL CALIBRATION of the associated protective relays, b) An integrated system functional test which includes simulated automatic actuation of the system and verifying that each relay and associated circuit breakers and control circuits functfon as designed, and SHEARON HARRIS - UNIT 1 3/4 8-L9
TABLE 3.8-1 CONTAINHENT PENETRATION CONDUCTOR OVERCURRENT, PROTECTIVE DEVICES This table is deleted from Technical Specifications.
Techrita~
The i tio 'his Specification Equipment List t ble i controlle plant procedure PLP-10 , Technica Program~
PACES 3/4 8-22 THROUGH 3/4 8-38B HAVE BEEN DELETED.
SHEARON HARRIS - UNIT 1 3/4 8-21 Amendment No. Q
ELECTRICAL POSER SYSTEMS ELECTRICAL E UIPMENT PROTECTIVE DEVICES
$ pggiFied w %c. Tcchnica.l
'MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION SpectPicnh'on Eqgipme<+ ~+
p(~~ ~ p(~qgt~gfgrC.PLp-lOL LIMITING CONDITION FOR OPERATION 3.8.4.2 The thermal overload protection of each valv~ quiring bypass protec-tioqphall,be bypassed only under accident conditions 'by an OPERABLE bypass de-vice integral with the motor starter..
APPLICABILITY: Whenever the motor operated valve is required to be OPERABLE.
ACTION:
Nth the thermal overload protection for one or more of the above required valves not capable of being bypassed under conditions for which it is designed to be bypassed, restore the inoperable device or provide a means to bypass the thermal overload within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , or declare the affected valve(s) inoperable and apply the appropriate ACTION Statement(s) of the affected system(s).
SURVEILLANCE RE UIREMENTS 4.8.4.2 The thermal overload protection for the above required valves shall be verified to be bypassed only under accident conditions by an OPERABLE integral bypass device by the performance of a TRIP ACTUATION DEVICE OPERATIONAL TEST of the bypass circuitry:
a. At least once per M months for those thermal overloads which are normally in force during plant operation and are bypassed only under accident conditions; and
b. Following maintenance on the motor starter.
SHEARON HARRIS - UNIT 1 3/4 8-39
TABLE 3.8-2 MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION This cable is deleted from Technical Specifications.
xn ormetaon in this table s controlled by plant procedure PLP-10 , Technica
+C ~ital Specification quipment Last Programs
,PACES, 3/4 8-41 THROUGH 3/4 8-43 HAVE BEEN DELETED.
SHEARON HARRIS - UNIT 1 3/4 8-40 Amendment No. R
REACTIVITY CONTROL SYSTEMS BASES HODERATOR TEMPERATURE COEFFICIENT The most negative MTC, val.ue equivalent to the most positive moderator density coefficient (MDC), was obtained by incrementally correcting the HDC used in the FSAR anaLyses to nominal operating conditions. These corrections invoLved:
(1) a conversion of the HDC used in the FSAR safety analyses to its equivalent (Continued)'egdiv<
HTC, based on-the rate oE change oE moderator density with temperature at RATED THERMAL PO'MER conditions, and (2) subtracting from this value the largest differences in HTC observed between EOL, aLL rods vithdrawn, RATED THERMAL POWER conditions, and those most adverse conditions of moderator temperature and pressure, rad insertian, axial pover skeving, and xenon concentration that can occur in normal operation and lead to a significantly more negative EOL MTC at RATED THERMAL POMER. These corrections transEormed .
the MDC valu sed in the FSAR saEety analyses into the imitin HTC value of
-4 The TC value -41.5 cm F represents a conservative MTC vaLue
~vc>>+ at a core condition of 300 ppm equiLibrium boron concentration, and i obtained b makin correct'ons Eor burnup and soluble boron to the l.imitin MTC value of "4 cm F. swvcitJancc. limni+
gag ppw The SurveilLance Requirements Eor measurement of the HTC at the beginning and near the end of the fuel. cycle are adequate to confirm that the HTC remains within its Limits since this coefficient changes slowly due pcincipalLy ta the reduction in RCS boron concentration associated vith fuel, burnup.
3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor vilL not be made criticaL with the Reactor CooLant System average temperature Less than 551'F ~ This limitation is required to ensure'. (1) the moderator temperature coefficient is vithin its analyzed temperature range, (2) the trip instrumentation is vithin" its normaL, operating range, (3) the pressurizer is capable of being in an OpERABLE status with a steam bubble, and (4) the reactor vesseL is above its minimum RTNDT temperature.
3/4. 1. 2 BORATION SYSTEMS The Boron Injection System ensures that negative reactivity control is available during each mode of faciLity operation The components required to perform this function include! (1) boratcd water sources, (2) charging/safety injection pumps, (3) separate flov paths, (4) boric acid transfer pumps, and (5) an emergency pover supply from OPERABLE dieseL generators.
>ith thc RCS average temperature above 350'F, a minimum of tvo boron injection Elow paths are required to ensure single functional capability in the event an assumed fail,ure renders one of thc flow paths inopcrabl.c. The boration capa-bility of either ELav path is sufficient to provide the rcqui "ed SHUTDO&
MARGIN as defined by Specification 3/4.1.1.2 after xenon decay and cooldovn to 200'F. The maximum expected boration capability requirement occurs at BOL SHEARON HARRIS - UNIT 1 B 3/4 1-2 Amendment No. P> M
3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section provide assurance of fueL integrity during Condition I (Normal Operation) and II (Incidents of Moderate frequency) events by! (L) maintaining the minimum DNBR in the core greater than or equal to the design DNBR vaLue during normal operation and in short-term transients, and (2) limiting the fission gas release, fuel peLLet temperature, and cladding mechanicaL properties 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 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:
Fq(Z) Heat Flux Hot Channel Factor, is defined as the maximum Local heat flux on the surface of a fuel rod at core elevation Z divided by the average fueL rod heat flux, allotting for manufacturing tolerances on fuel peLlets and rods',
N dH 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; the FO limit specified in the CORE 3/4.2.L AXIAL FLUX DIFFERENCE OPERATING LIMITS REPORT (COLR)
The Limits on AXI I FLUX DIFFERENCE (AFD) assure that the F~(Z) upper bound envelope of times the normaLired axial peaking factor as not exceeded during either normal operation or in the event of xenon edistributio folLowing power changes. The normalized axia peaking factor is specified in the OLR.
Target flux difference (TARGET AFD) is determined at equilibrium xenon condi" tions. The rods may be positioned within the core in accordance with their respective insertion Limits and should be inserted near their normaL position for steady-state operation at high power leveLs. The value of the target flux difference obtained under these conditions divided by the fraction of RATED THERMAL POWER is the TARGET AFD at RATED THERMAL POWER for the associated core burnup conditions ~ TARGET AFD for other THERMAL POWER LeveLs are obtained by multipLying the RATED THERMAL POWER value by the appropriate fractionaL THERMAL POWER Level. The periodic updating of the target flux difference value is necessary to reflect core burnup considerations.
Fa,H 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, with an allowance to account for measurement uncertainty.
SHEARON HARRIS - UNIT l B 3/4 2-1 Amendment No. 7, g
POWER DISTRIBUTION LIMITS BASKS AXIAL FLUX DIFFERENCE (Continued) pygmy p~ Rhoc operation, At ower levels beLow APL, the limits on AFD are specified in the ore Operating Limits Re ort x.e. t at de oned b the RAOC o eratxn rocedure and Limits. These limits were calculated in a manner such that expected operational transients, e.g., load Eollow operations, would not result in the AFD deviating outside of those Limits. However, in the event such a deviation occurs, the short period of time aLlowed outside oE the Limits at reduced power Levels will'not result in significant xenon redistribution such that the envelope oE peaking factors would change sufficiently to prevent operation in the vicinity of the APL power level.
At power leveLs greater than APL , two modes of operation are permissible.')
RAOC with Eixed AFD Limits a= a Eunction of reactor power leveL and 2) Base Load operation which is defined as the maintenance oE the AFD within a band about a target value. Both the fixed AFD Limits Eor RAOC o erati n and the band Eor Base Load operation are specified for Lo d c cle in the OR
~LR, OPERATION RAOC craton roce ur above APL a the same as e Eor operation beLow APL
<<However, it is possible when following extended'Load following maneuvers that the AFD Limits may result in restrictions in the maximum. allowed power or AFD in order to guarantee operation with F~(Z) less than its Limiting value. To aLLow operation at the maximum permissible value, the Base Load operating procedure restricts the indicated AFD to a relatively small target band and power swings'or Base Load operation, it is expected that the plant wiLL operate within the target band. Operation outside of the target band for the short. time period allowed wilL not resuLt in significant xenon redistribution such that the envelope of peaking factors wouLd change suEficiently to prohibit continued operation in the power-region deEined above. To assure there is no residual xenon redistribution impact from past operation on the Base Load operation, a 24"hour waiting period at a power Level above APL and allowed by RAOC is necessary. During this time period, Load changes and rod motion are restricted to that alLowed by the Base Load procedure. AEter the waiting period, extended Base Load operation is permissible.
The computer determines the one-minute average of each oE the OPERABLE excore detector outputs and provides an alarm message immediateLy if or more OPERABLE excore channels are: 1) outside the allowed 4I power the AFD for two operating space (for RAOC operation), or 2) outside the acceptable AFD target band (Eor Base Load operation) ~ These alarms are active when power i~ greater than: 1) 50X of RATED THERMAL POWER (for RAOC operation), or 2) APL" (for Base Load operation). Penalty deviation minutes Eor Base Load operation are not- accumulated based on the short period of time during which operation outside oE the target band is allowed.
SHKARON HARRIS - UNIT 1 B 3/4 2-2 Amendment No. g, H
POWER DISTRIBUTION LIMITS BASES 3/4.2.2 AND 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR The limits on heat flux hot channeL factor, RCS flow rate, and u lear enthalpy rise hot channel factor ensure thatt (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.
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 surveiLLance is
~
sufficient to ensure that the limits are maintained provided:
a. Control rods in a single group move together with no individual rod insertion differing by more than + 12 steps, indicated, from the group demand position',
b. Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.63 SHEARON HARRIS - UNIT 1 B 3/4 2-2a Amendment No. 7, Q
POWER DISTRIBUTION LIMITS BASES HEAT FLUX HOT CHANNEL FACTOR', AND RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Concinued)
c. The control rod insertion Limits of Specifications 3.1.3.5 and 3,1e3.6 are maintained; and
d. The axial power distribution, expressed in terms oE AXIAL FLUX DIFFERENCE, is maintained vithin the limits ~
(
F H
wilL be maintained within its limits provided Conditions a. through d.
above are maintained. The combinations of the RCS Elov requirement and the measurement oE F<H ensures that the calculated DNBR vill noc be bel.ow the design DNBR value. The relaxation oE F<H as a Eunction oE THERMAL POWER allows changes in the radial power shape for al.L permissible rod insertion Limits.
F<H is evaluated as being less than or equaL to L.56 Eor LOPAR fuel and L.59 Eor VANTAGE 5 Euel ~ These vaLues are used in the various accident analyses N
where F<H inEluences parameters other than DNBR, e.g., peak clad temperacure, and thus is the maximum "as measured" vaLue alloved.
Margin is maintained between the .safety anal.ysis limit DNBR and the design
'Limit DNBR. This margin is more than sufficient to oEEset any rod bov "
penalty and transition core penalty.
s When an F< measurement is taken, an allovance Eor both experimental error and manufacturing tolerance must be made. An allovance of SZ is appropriate Eor a EuLL-core map taken with the Incore Detector Flux Mapping System, and a 3Z aLlovance is appropriate Eor manufacturing toLerance.
The hot channel factor F M (Z) is measured periodically and increased by a cycle and height dependent pov r factor appropriate to either RAOC or Base Load operation, W(Z) or W(Z)BLf to provide assurance that the Limit on the hot channel factor, F (Z), as mec. W(Z) accounts Eor the effects of normaL .
operation transients and was determined Erom expected power control maneuvers over the full range of burnup conditions in the core. W(Z)BL accounts Eor the more restrictive operating Limits allowed b Base Load operation vhich result in Less'severe transient values. The W(Z) function tor norma operatton ts provided xn the Core Operating Laments Re ort er v ecification 6 (
and .le(Z) funChOnS are, SPeCifjed in +C COLR.
BL When an FgH measurement is taken, an allowance for measurement error must be applied prior to comparing to the RTP F>H limit(s) specified in the CORE OPERATING LIMITS REPORT (COLR). An allowance of 4S is appropriate for a full-core map taken with the Incore Detector Flux Mapping System.
r SHEARON HARRIS - UNIT 1 B 3/4 2-4 Amendment No. 7p g
POWER DISTRIBUTION LIMITS BASES HEAT FLUX HOT CHANNEL FACTOR AND RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR Continued A,measurement error of 4Z for F H has been allowed for in determination of the design DNBR value. When RCS o rate is measured, no additional allovance is necessary prior to comparison vith the limit of Specification 3.2.3 ~ A normaL RCS Elowrate error of 2.1X wilL be incl.uded in Cl, vhich vilL be modified as discussed bel.ov.
The measurement error for RCS totaL ELov rate is based upon performing a precision heat balance and using the result to calibrate the RCS flow rate indicators. PotentiaL fouling oE the feedwater venturi which might not be detected couLd bias the result from the precision heat balance in a non-conservative manner. Therefore, a penal.ty of 0.1Z Eor undetected Eouling of the feedvater venturi, raises the nominal Elov measurement allowance, Cl f to 2.2Z for no venturi fouling. Any fouling which might bias the RCS Elov rate measurement greater than O.LZ can be detected by monitoring and trending various plant performance parameters. If detected, action shall be taken before performing subsequent precision heat balance measurements, i.e., either .
'the effect oE the fouling shall be quantified and compensated for in the RCS flov rate measurement or the venturi shall be cleaned to eliminate the fouLing.
The 12-hour periodic surveiLLance of indicated RCS Elov is sufficient to detect only Elov degradation that could Lead to operation outside the accept-able region of operation.
3/4 '.4 UADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO Limit assures that the radial po~er distribution satisfies the design values used in the pover capability analysis, Radial pover distribution measurements are made during STARTUP testing and period-ically during pover operation.
The limit of 1.02, at which corrective action is required, provides DNB and linear heat generation rate protection vith x-y plane pover tilts. A Limiting tilt of 1 ~ 025 can be tolerated before the margin for uncertainty in Fq is depleted. A limit oE 1 ~ 02 vas selected to provide an aLlowance Eor the uncertainty associated vith the indicated power tilt.
The 2-hour time allovance for operation with a tilt condition greater than 1.02 but Less than L.09 is provided to allow identification and correction of,a dropped or misaligned control rod. In the event such action does not correct the tilt, the margin for uncertainty on F~ is reinstated by reducing the maximum alloved pover by 3Z for each percent oE tilt in excess of 1 ~
SHEARON HARRIS - UNIT 1 B 3/4 2-5 Amendment No. N
ADMINISTRATIVE CONTROLS 6.9. 1.6 CORE OPERATING LIMITS REPORT operation (tel.R ~ an& rocc urc PLP-IO&
6.9.1.6.1 Core Limats shall be established and documented in the CORE OPERATING LIMITS REPORT prior to each reload cycLe, or prior to any remaining portion oE a reload cycle, for the foLlowing'.
a ~ The shutdown rod insertion Limits of Specification 3. 1.3.5 ~
(ace, graf b. The controL rod insertion Limits oE Specification 3 ~ 1.3.6.
~noh Znsere 4 c ~ The axial. fLux difEerence of SpeciEication 3.2.L.
(nW+Py)
d. The surveillance requirements of Specifications 4.2.2.2, 4 '.2.3, and 4.2.2.4.
6.9.1.6.2 The anaLyticaL methods used to determine the core operating limits shaLl. be those previously reviewed and approved by the NRC, speciEically those described in the EoLLowing documents'.
gcplace. a ~ WCAP-10216-P-A, ReLaxation of Constant AxiaL OEfset ControL F~
fg)l'fL Surveillance Technical SpeciEication, 1983.
Znzeri.k
b. WCAP-9272-P-A, Wes t inghouse Reload Sa fe ty Evaluat ion Me thodo Logy,
(~~ pg<) 1985.
6.9.L.6.3 The core operating Limits shall be determined so that all applicable Limits (e.g., Euel thermaL-mechanical limits, core thermaL"hydraulic limits, nucL'ear limits such as shutdown margin, and transient and accident analysis Limits) of the safety anaLysis are met.
6.9.1.6.4 The CORE OPERATING LIMITS REPORT, incLuding any mid-cycle revisions or supplements, shaLL be provided, upon issuance Eor 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 RegionaL OEEice of the NRC within the time period specified for each report ~
6. 10 RECORD RETENTION 6.10.1 In addition to the applicable record retention requirements oE TitLe 10, Code oE Federal Regulations, the following records shaLL be retained Eor at Least rhe minimum period indicated.
6.10.2 The EoLLowing records shall be retained Eor at Least 5 years:
a. Records and Logs of unit operation covering time interval at each po~er Level; SHEARON HARRIS - UNIT 1 6-24 Amendment No. J, J3, y5
INSERT A:
a~ Moderator Temperature Coefficient Positive and Negative Limits and 300 ppm surveillance limit for Specification 3/4.1.1.3,
b. Shutdown Bank Insertion Limits for Specification 3/4.1.3.5, C ~ Control Bank Insertion Limits for Specification 3/4.1.3.6,
d. Axial Flux Difference Limits, target band, and APL for Specification 3/4.2.1, RTP
e. Heat Flux Hot Channel Factor, F , K(Z), W(Z), APLND and W(Z)BL for Specification 3/4.2.2, RTP Enthalpy Rise Hot Channel Factor, F H
, and Power Factor Multiplier, PF for Specification 3/4.2.3.
INSERT B:
a ~ WCAP-9272-P-A, "WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY",
July 1985 (W Proprietary).
(Methodology for Specification 3.1.1 ~ 3 - Moderator Temperature Coefficient, 3.1.3.5 - Shutdown Bank Insertion Limit, 3.1.3.6 - Control Bank -
Insertion Limit, 3.2.1 Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 " Nuclear Enthalpy Rise Hot Channel Factor).
b. WCAP-11914, "SAFETY EVALUATION SUPPORTING A MORE NEGATIVE EOL MODERATOR TEMPERATURE COEFFICIENT TECHNICAL SPECIFICATION FOR THE SHEARON HARRIS NUCLEAR POWER PLANT", August 1988 (W Proprietary).
(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient).
C ~ WCAP-10216-P-A, "RELAXATION OF CONSTANT AXIAL OFFSET CONTROL Fq SURVEILLANCE TECHNICAL SPECIFICATION", JUNE 1983 (W Proprietary).
(Methodology for Specifications 3.2.1 Axial Flux Difference (Relaxed Axial Offset Control) and 3.2.2 - Heat Flux Hot Channel Factor (F~
Methodology for W(Z) surveillance requirements).)
d. WCAP-10266-P-A, Rev. 2, "The 1981 Version of the WESTINGHOUSE ECCS EVALUATION MODEL USING THE BASH CODE", March 1987 (W Proprietary).
(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor).
e. WCAP-8385, "POWER DISTRIBUTION CONTROL AND LOAD FOLLOWING PROCEDURES TOPICAL REPORT", September 1974 (W Proprietary).
(Methodology for Specification 3.2.1 - Axial Flux Difference (Constant Axial Offset Control).)
WCAP-11837-P-A, "EXTENSION OF METHODOLOGY FOR CALCULATING TRANSITION CORE DNBR PENALTIES", January 1990 (W Proprietary).
ENCLOSURE 6 SHEARON HARRIS NUCLEAR POWER PLANT DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT CYCLE 4 RELOAD SUBMITTAL SAMPLE CORE OPERATING LIMITS REPORT
SHEARON HARRIS UNIT 1 CYCLE g CORE OPERATiNG LIMITS REPORT REVISION 0
SHEARON HARRIS UNIT 1 CYCLE g 1.0 Core Operating Limits This Core Operating Limits Report for SHlCg has been prepared in accordance with the requirements of Technical Specification 6.9.1.6. The core operating limits have been developed using the NRC-approved methodologies specified in References The following cycle-specific core operating limits are included in this report:
1) Control Rod Insertion Limits (Figure 1). Fully withdrawn for all control and shutdown banks shall be 228 steps.
2) AFD Limits and W(z) Curves for RAOC Operation (Figures 2 through 5).
These W(z) curves are sufficient to determine the RAOC W(z) versus core height for Cycle P'burnups through the end of full power reactivity plus a po~er coastdown of up to 1180 MWD/MTU through the use of three point interpolation.
ND
3) AFD Limits, APL , and W(z) Curves fo" Base Load Operation (Table 1 and Figures 6 through 8) . These W(z) curves are sufficient to determine the Base Load W(z) versus core height for Cycle P burnups through the end of full power reactivity plus a power coastdown of up to 1180 MWD/MTU through the use of three point interpolation.
2N5E'g7 R )'f~c4~e~y 4 Agf-P 2.0 References
l. "We tingho e Reload afet Evaluation eth dology," W -9 2-P-A, July 985
2. "Rela ati of Const t Axi 1 Offset ntrol F Surv illance Tec ical S cific tion," W -10216 P-A, June 83.
Page I of ~(REV 0)
Attachment 1
4) Moderator Temperature Coefficient (MTC)
(Specification 3/4.1 '.3) 4.1 The Moderator Temperature Coefficient (MTC) limits are as follows:
The Positive MTC Limit (ARO/HZP) shall be less positive than
+5 pcm/ F for power levels up to 70K RATED THERMAL POWER and a linear ramp from that point to 0 pcm/ F at 100K RATED THERMAL POWER.
The Hegative MTC Limit (ARO/RTP) shall be less negative than
-49 pcm/ F.
.4.2 The MTC Surveillance limit is as folio'ws:
The 300 ppm/ARO/RTP-MTC should be less negative than or equal to -41.5 pcm/ F.
where: ARO stands for All Rods Out HZP stands for Hot Zero THERMAL POWER RTP stands for RATED THERMAL POWER
5) Heat Flux Hot Channel Factor - FO(Z)
(Specification 3/4.2.2)
FRTP 0
><(1> c x K(Z> for P > 0.5 P
FRTP FO for P < 0.5 0.5 THERMAL PO'WER where: P RATED THERMAL POWER 5.1 F
" 2 45 a
5.2 K(Z) is specified in Figure 9
l Attachment 1 Continued t
6) Nuclear Enthalpy Rise Hot Channel Factor -
F<H
<Specification 3/4.2 ')
RTP F~H <
F~H [1.0 +
PF~ H(1.0-P)]
THERHAL PONER where: P RATED THERNAL POMER RTP 6.1a F+H
1.62 for LOPAR fuel RTP
6. 1b F<H 1.65 for VANTAGE 5 fuel 6.2a PF~H -"0.3 for LOPAR fuel 6.2b PF<H = 0.35 for VANTAGE 5 fuel
I
~ ~
Attachment 2
1. MCAP-9272-P-A, "WESTINGHOUSE REI.OAD SAFETY e EVALUAT ION HETHODOLOGY", July 1985 (W P ropr i etary) .
(Hethodology for Specification 3.1.1.3 - Noderator Temperature Coefficient,3.1.3.5 - Shutdown Bank Insertion Limit,3.1.3.6 - Control Bank Insertion Limit, 3.2.1 - Axial Flux Difference,3.2.2 - Heat Flux Hot Channel Factor,and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor).
2. WCAP-11914, "SAFETY EVALUATION SUPPORTING A NORE NEGATIVE EOL HODERATOR TENPERATURE COEFFICIEHT TECHNICAL SPECIFICATION FOR THE SHEARON HARRIS NUCLEAR POWER PLANT", August 1988(W Proprietary).
(Hethodology for Specification 3.1.1.3 - Noderator
'Temperature Coefficient) ~
3 ~ MCAP 10216-P A, "RELAXATION OF CONSTANT AXIAL OFFSET CONTROL FQ SURVEILLANCE TECHNICAL SPECIFICATIOH", June 1983 (W Proprietary).
(Hethodology for Specifications 3.2.1 - Axial Flux Difference (Relaxed Axial Offset Control) and 3.2.2 - Heat Flux Hot Channel Factor (FG Hethodology for W(Z) surveillance requirements).)
4. WCAP.10266-P-A,REV.2,"THE 1981 VERSION OF THE MEST INGHOUSE ECCS EVALUATION NODEL USING THE BASH CODE", Narch 1987 (W Proprietary).
(Nethodol ogy for Specification 3 .2 .2 . Heat Flux Hot Channel Factor).
5 . WCAP 8385, "POWER DISTRIBUTION CONTROL AND LOAD FOLLOWING PROCEDURES - TOPICAL REPORT", September 1974 (W Proprietary).
(Nethodology for specification 3.2.1 - Axial Flux Difference (Constant Axial Offset Control).)
6. WCAP-11837 P-A, "EXTENSION OF NETHODOLOGY FOR CALCULATING TRANSITION CORE DNBR PENALTIES", January 1990 (W Proprietary).
FIGURE 1 SHEARON HARRIS UNIT 1 CYCLE g ROD GROUP INSERTION LIMITS VERSUS THERMAL POWER (TH REE-LOOP OPERATION) 220 .538 228 200 186 c 180 O
L BANK C
~ 160
~ '140 CL CP 0 128 120 CD BANK
~ 100 (l)
CO 40 20
0. 0
.00 .10 .20 .30 .40 .50 .60 .70 .80 .90 -
1.0 FRACTION OF RATED THERMAL POWER (Fully withdrawn shall be 228 steps)
This figure is referred to by Technica1 Specifications 3.1.3.1.d.2 3.1.3.5 3.1.3.6 Page +of Pf (REV 0)
I
FIGURE 2 SHEARON HARRIS UNIT 1 CYCLE jf AXIALFLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER FOR RAOC 120 110 100 7 100 100 I I
(
UN CCEPT BLE) U ACOEPT BLE 90 I !
l l 80 CEPiTAB E I 70 i
I I I I 60 I I I l 50 C)
-34, 50 28. 50 Z~ 40 30 20 10 05 0 -40 -30 -20 -10 0 10 20 30 40 50 AX(AL FLUX DlFFERENCE (Pg DELTA-I)
This figurc is referred to by Technical Specifications 3.2.1 4.2.2.2f B3/4.2.1 Pago P of P((REV 0)
e ~
a
s FIGURE 3 SHEARON HARRIS UNIT 1 CYCLE jf RAOC W(Z) AT 150 MWD/MTU Height 50L
~(F an't ~ls 2 I
e 0. 00 1. 0000 s s s e 0.20 1.0000.
0.40 1.0000 o'.5o 1.0000 0.80 1. 0000 1.00 1.0000 e 1.20 1. 0000 1.40 1.0000
!! i I I I 1 50~ 1 ~ OOOO!
I! ii ~
I
'!.80 2.00
. 3556
. 3335 2.20 . 3010 2.40 .2583 ls I 2.50 ~ 2356 I
2.80 ,2031 3.00 . 1717 3.20 1. 1455 I I I I 3.40 1, 1334 I I I I 3.50 . 1. 1295 3.80 . 1251 4.00 . 1232 s I 4.20 1. 1207 I! I! Il III: 1.40 1. 1172 I I I I I I 4.50 1. 1131
!4! 1. 80
5. 00
1. 1084
1. 1038 I ~ ~
5.20 1.0083 5.10 1. 0033 I! 5;do 1.0042
~
5.80 1 .OQQO I ~
i 5.00 1. 1051 5.20 1. 1134 5.40 1. 1100 d.do 1. 1234 5.8o 1. 12dd 7.00 . 1270 7.20 1. 1202 7.40 . 1314
~ s ~
4! I! I 7.50 .1325 II sl 7. 80 . 1325 8.00 . 1311 8.20 . 1285 8.10 1. 1241 8.50 1. 1181 I SS s I I I 8 80 F 1. 1204 I I s ! I II ! I I 0.00 1. 1270 s
0.20 1, 135Q I I } I I I I I I I I Q.40 1. 142Q
! ! I 0.50 1. 1192 9.80 1. 1553 10.00 1. 1592
! ! I II ;Is! e 10.20
10. 40
1. 1805 1 . 0000 ls, I 1o'.do 1. 0000 ss I e 10. 80 1.0000
'11.00 1. 0000 e 11.20 1.0000 11.40 1. 0000 s! II e 11.50 1. OOOO e 11. 80 1.0000 2 3 5 8 9 10 11 12 12.00 1. 0000 CORE HEIGH eet) Top and Notte 15' xcluded as peg Technical Spec10 1cat1 on 1.2.2.2.
This figurc is referred to by Techni cal Specifications 4.2.2.2c S3/4.Z.2 Page P'of P6 (REV 0)
~r 1 y T
I ~
FIGURE 4 F
SHEARON HARRIS UNIT 1 CYCLE pf RAOC W(Z) AT 7000 MWD/MTU Height IC)I.
~FIFE I I 0.00 1.0000
< ~
0.20 1. 0000 0.40 1. 0000
~ i s
0.50 1. 0000 IE Q ~ 8Q 1,0000 I 1.00 1. 0000 1.20 1. 0000 s 1. 0000
~
I s 1,50 1. 0000 1.80 .2148
2. 00 . 1873 2.20 . 1802 2.40 . 1d41 2.50 .1480 2.80 . 132d 3.00 . 1272 3.20 . 1277 3.40 . 1250 3.50 . 1241 3.80 . 1220 4.00 , 1177 4.20 . 1171 4.40 . 122d 4.50 . 1272
~ ~ I (('. 4.80 ~ 1308 4( I I I! 5.00 . 1340 5F20 , 1345
~
I
~
5.40 . 1378 I I 5.'50 . 1552
~
5.80 . 1733 d.00 . 1882
5. 20 .2032 5.40 ~ 2148
~ IEI 5.50 .2242 5.80 .230Q 7.00 ~ 2350 7.20 .2357 I! !! I I 7.40 .2357 I 7.50 .2322 I I ll 7.80 8.00
.2250
.2171 8.20 .2058 8.40 . 11td 8.50 . t747 I I . I 8.80 . 1515 fl (! ~
I ~ ('( Q.OO Q.20
. 1453
. 1413
~ ~
Q.40 . 1545 Q '.5O . 1587 Q. 80 . 1801 10.00 . 2051
10. 20 .2373
10. 40. 1 FQOOO
10. do: 1. OOOO
10. 80 1. 0000 s 1 '1 QQ 1 QQQQ 11.20 1.0000
~ I ~
I
~ s 11.40 1.0000 1 1 .50 1.0000 11.80 1.0000 4 5 6 7 8 9 IO 1I I 12.00 1. 0000 s Top snd bott~ 15%
CORE HEIGH T (F eet) excluded as pe(
Tecnn 1 ca I Spec 1 0 1 catt on 4.2.2 This figure ia referred to by Technical Specificationa 4.2.2.2c S3/4.Z.2 Page P of P4 (REV 0)
~
l' 0
FIGURE 5 SHEARON HARRIS UNIT 1 CYCI.E 8 RAOC W(Z) AT 13000 MWD/MTU Height EOL
~Faat ~Vz S O. 00 1. 0000 S 0.20 1.0000 S 0.40 1. 0000 S 0.50 1. 0000 S 0'.80 1. 0000 S 1. 00 1. 0000 S 1.20 1.0000 S 1. 40 1.0000 S 1.50 1. 0000
1. 80 .2390
2. 00 1.2196 2.20 . 1999 2.40 . 1798 2.50 1. 1592 2.80 . 1382 I I I I I I 3.00 . 1173 3.20 . 1085 3.40 . 1151 3.50 1. 12d5 3.80 1. 1378 4.00 1. 1487 lit jl 4.20 1. 1585 I III jl(I 4.40 4.50 1 1572
~
'I. 1741 4.80 'I. 1804 I I I I 5.00 1. 1834 i I 5.20 1. 1892 I! I I I I 6.40 5.50
1. 2029 1.21Q4 I! '
5.80 1. 2335
~
4 ~
I
~
5 00 F .2447 6.20 1.2533 5.40 1.25QO I I I ~ 5.60 .2517 II I 5.80 1. 2513 I 7.00 . 2579 ltl I 7.20 7.40 1.2517 1 2425 a
7.50 1. 2307 7.80 1.2150 8.00 1. 19Q2 8.20 1. 1780 8.40 1.'1581 8.50 1. 1532 I 'iji I! II 8.80 9.00 Q.20
1. 1544
1. '1541
1. 1523 9.40 1. 15'72 j(l '> I 'III I
Q.do 1 1753 jjjj
~
9.80 'I. 1938 it jifl tltl I 10. 00 I i I I I I 1. 2205 i 10.20 'I . 2496 S 10. 40 1. 0000
10. 50 1, OOOO S 10.80 1. 0000 S 11.00 1. OOOO S 11.20 1. 0000 S 11.40 1. 0000 S 11. 50 1. 0000 S 11.80 1. 0000 3 4 5 6 7 8 9 10 11 12 S 12.00 1.0000 CORE HEIGH( (F eet) s Top and bottoe 15%
excludaa as per.
Technical Speci f ication 1.2.2.2.
This figure to by is'eferred Technical
'pecifications 4.2a2.2c S3/4.2.2 Pafa j( of (t6 (RE> 0)
~
I t I k
SHEARON HARRIS UNIT 1 CYCLE P TABLE 1 BASE LOAD OPERATING LIMITS The following data is required to define the operating limits for Base Load Operation. These data are referred to in Technical Specifications 3.2.1, 4.2.2.3a, 4.2.2.4c, and B3/4.2.1.
Parameter Operating Limits AFD Limits within a i3X band about the target AFD 85 percent of rated thermal power W(z) Curves Figures 6 through 8 page /of /0'(REV 0)
fP FIGURE 6 SHEARON HARRIS UNIT 1 CYCLE gf BASE LOAD W(Z) AT 150 MWD/MTU "
FOR POWER LEVELS ABOVE 85% OF RATED THERMAL POWER Height 80I
~Face ~ll a 1.10 O. 00 1. 0000 0.20 1.0000
0. 40 1.0000 0.80 1.0000
$ 0.80 1. 0000 1.09 1.00 1.0000 1.20 1. 0000
$ 1.40 1. 0000
$ 1.50 1 .0000 1.80 1.0551
2. 00 1, 0552 1.08 2.20 1,0555
2. 40 1,0557 2.50 1.0557
2. 80 1. 0551
3. 00 1. 0541 I
3.20 '1
. 0527 I.07 3. 40 1. OSOQ I ~ 3.50 1.0491
3. 80 1. 0475
! I 4. 00 1.0459
~ I 4.20 1 .0445 4.40 1. 0433 1.06 4.50 1.0419 4.80 1 .0404 5.00 1.0388 6.20 1. 0371 5.40 1. 0353 5'.eo 1.0334
5. 80 1. 0316 e.'00 1.0294 8.20 1. 0272 I I I i 6.40 1 0248
~
5 eo
~ '1. 0222 5.80 1.0192 1.04 7. 00 1. 0170 7.20 1.0184 7.40 1.oie&
7.80 1.0171
7. 80 1.0188
&.00 1.0210 8.20 1.0234 1.03 I II II 8.40 1. 0258
~ 8. eo 1.0283 8.80 1.030Q 9.00 '1 . 0334 i I 9.20 1.0358 1.02
~
( I I I I ~
9.40 1. 0383 Q.eo 1. 0407 9.80 1. 0430
10. 00 1. 0452 10.20 1. 0472
~ 10. 40 1. 0000 1.01 10.60 1.0000
10. 80 1. 0000 11.00 1,0000 11.20 1. 0000 11.40 1. 0000
$ 11. eo 1, 0000 I 11.80 1.0000 1.000 12.00 1. 0000 2 3 4 5 6 7 8 9 10 11 12 CORE HEIGHT (F get) $ Top et@I Iottoa 15K excluded aa pe}-
Technical Speci f 1 cation 4e2.2.4.
This figure ia referred to" by Technical Specificationa 4.2.2.3a
4. 2,2. 4c S3/4.Z.2 Page (REV 0)
P afar
FIGURE 7, SHEARON HARRlS UNlT 0 CYCLE g BASE LOAD W(Z) AT 7000 MWD/MTU j
FOR POWER LEVELS ABOVE $ 5% OF RATED THERMAL POWER Height
~Feet I I I ~ 0. 00 1. 0000 0.20 1.0000 0.40 1.0000 0,50 1.0000 0.80 1.0000
1. 00 1.0000 .
1.20 1. 0000 1.40 1. 0000 E 1.50 1 . 0000
1. 80 1. 053B
2. 00 1. 0524 2.20 1.0813.
2.40 1. 0501 2.50 1.0585
'.80 3.00 1.0553
1. 0535 3'.20 1.0503 3.40 1.0471 3.50 1. 0453 3.80 1. 0441 4.00 1.0428 4.20 1. 0414 4.40 1.03$ 7 4.50 1.0380 4.$ 0 , 1.0351 5.00 1.0341 5.20 1,0321 5.40 1.0300 5.50 1.0278 S.80 1.0255 e.'oo 1.0233 5.20 e'.4o 1.0223 1 . 0227 5.50 1.0225 5.$ 0 1.0222 7.00 1.'O218 7.20 1.0213 7.40 1. 0210 7.50 1.0225 7.$ 0 1.0254 8.00 1.028d 8.20 1.031Q 8.40 1. 0352
$ .50 1. 0385 8.80 1.0417 Q.oo 1.044Q Q.20 1. 0480 Q.40 1. OS12 Q.eo 1. DS42 Q.80 1. 0572
10. 00 1,0SQQ 10.20 1,0524 10.40 1 . DODD
'O.'BO
1. 0000
10. 80 1. 0000 11.00 1. 0000 11.20 1. 0000 11.40 1. 0000 11.50 1. 0000 11.80 1. OOOO i 5 7 8 9 10 11 1 12.00 1. 0000 CORE HEIGHT (feet) Top Hxt Sottoa 15%
exc1LIdeo as pet Technl ca 1 Speci f 1Ca Son 4.2.2.4.
t This figure is referred to by Technical Specifications 4.2.2e3a 4.2.2.4c S3/4.Z.2 Peee)( of Ph (REV 0)
FIGURE 8 SHEARON HARRIS UNIT '1 CYCLE g BASE LOAD W(Z) AT 13000 MWD/MTU FOR POWER LEVELS ABOVE 85% OF RATED THERMAL POWER Height
~eeet
0. 00 1.0000 0.20 1.0000 0.40 1.0000 O.BO 1. 0000 0.80 1.0000 1.00 1.0000 1.20 1. 0000 1.40 1.0000 1.50 1. 0000
1. 80 1. 0776
2. 00 1.0745 2.20 1.0713 2.40 1. 057$
2.50 1. 0840 2.80 1. 05$ 5 3.00 1.0543 3.20 1.0487 3.40 1,0444 3.50 1.0434 3.$ 0 1, 0422 4.00 1.04OB 4.20 1. 03QO 4.40 1. 0371 4.50 1. 0351 4,80 1. 032Q 5.00 1. 0307 5.20 '1
. 0284 5.40 1. 0252 5.50 1.024B 5.80 1.0235 d.oo '.0235 5.20 '1 .0240 5.40 d.'BO 1.023Q 1.0235 5.80 1. 022$
7.00 1.0222 7.20 1. 020Q
'.40 1.0205 7tdo 1.0238 7.80 1. 0277 8.00 1. 031d 8.20 1. 0355 8 ~ 40 1.0387 Q.BO 1. 0437 8.$ 0 1. 0477 Q.OO 1. 0525
$ .20 1. 0575 Q.40 1 . 0524 Q.do 1.0588
$ .80 1.070$
10.00 1.0745 10.20 1.0778
10. 40 1.0000
10. do 1.0000 10.80 ,1*0000 11.00 1. 0000 11.20 1. OOOO
11. 40 1. 0000 11.50 1. OOOO S 6 7 8 9 10 11 1 2
' 11.$ 0 12.00
1. 0000 1.0000 ORE HE lGH t (F get) Top and dottcm excluaed as per 1',
Tecnni ca1 Speci f ication 4.2.2.4.
This figure is referred to by Technical Specifications 4.2.2.3a 4.2.2.4c S3/4.2.2 Page P6 of 36 (REV 0)
1.2 1.0
~N 0.9 lZ o,0.8 O
0.7 (3
z 0.6 gCL Elevation Normalized Peaking Factor c 0.5 N 0.0 1.0 6.0 1.0 12.0 0.925 0.2 0.1 0.0 0 1 2 3 4 5 6 7 . 8 9 10 11 12 CORE ELEVATION (FEET)
This figure is referred by Technical Specification a/o.2.2
' s P4

ML18009A664

Text

1.62 for LOPAR fuel RTP

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