Text

Proposed Tech Specs Changes to Permit Operation of Reactor W/One of Two Reactor Circulation Loops in Svc Under Certain Specified Conditions
	ML20205Q974
Person / Time
Site:	Limerick
Issue date:	11/04/1988
From:	; PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	;
Shared Package
ML19297H235	List: ML20205Q926; ML20205Q947; ML20205Q974;
References
	NUDOCS 8811090413
	Download: ML20205Q974 (22)
	v • d • e

_ _ - _ _ _ _ _ _ _ _ _ _ _ _ . _

2.0 SAFETY LIMITS AHO LIMITING SAFELY SYSTEM SETTINGS

?_.1 SAFELY LIM 11S lifLRMAL POWER. Low Pressure or low flow 2.1.1 TiiERMAL POWER shall not exceed 25% of RATED TilERMAL POWER with the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow.

M't.lCAulLITY: OPERATIONAL CON 0lil0HS I and 2.

ACTION:

With illERMAL POWER exe.?eding Pfr% of RAT [0 TilERMAL POWER and the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow, be in at least 1101 SilVT00WH within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and comply with the requirements of Specification 6.7.1.

IllERMAL POWER t fligh Prtssure and liigh flow 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than 1.07 for two recirculation loop operation and shall not be less than 1.08 for single recirculation loop operation with the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10% of rated flow.

APPLICABIL11Y: OPERATIONAL CONDITIONS 1 and 2.

AC T 10ti:

With MCPR less than 1.0/ for two recirculation loop operation or less than 1.08 for single recirculation loop operation and the reactor vessel steam dome pressure greater than 185 psig and core flow greater than 10% of rated flow, be in at least iloi Silul00WN within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and comply with the requirements of Specification 6.7.1.

i<t ACIOR C00l Atli SYSTEM PRESSURE 2.1.3 1he reactor coolant system pressure, as measured in the reactor vessel steam dome, shall not exceed 1325 psig. ;

APPIICABitIIY: OPli<All0tlAl. CONDI fl0liS 1, 2, 3, and 4. ,

AC110ll:

With the reactor coolant system pressure, as measured in the reactor vessel steam '

dome, above 1325 psig, be in at least 1101 SifuiOOWN with reactor coolant system pressure less than or equal to 1325 psig within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and comply with the requirements of Specification 6.7.1.

9V/109oit/3 !?tte tyg n ooc /t Qv3yL IIMIRlCF - UN!! l 2-1 0

TAEtt 2.2.1-1 .

REACTOP PPOTECTICN SYSTEM INSTRUMENTATION SETPolnis -

~

ALLOWABLE FUNCTIOGL UNIT TRIP SETFOINT VALUES ,

1. Intecediate Range Monitor, Neutron Flux-Hign i 20/125 1 divisions ,1 122/125 divisions of full scale of full scale -

2. Average Pc.er Range Monitor:

a. Neutron Flu >-U: scale, Setdown ~< 15% of FATED THERMAL POWER s 20% of RATED L. heutron Flu >-U scale -THERMAL POWER

1) Curing tao re:irculation loop operation:

a) Flow Biasec 1 0.5E W + 59%, -ith 2 0.58 W + 62%, with a maximum of a maximum of b) Hign Fic. Clamced i 116.5% of RATED ; 118.5% of RATED THERMAL POWER THERMAL POWER

2) During single recirculation locp operation:

a) Flow Btased < 0.58 W + 54% < 0.58 W + 57%

b) riigh Flc. Cla~yed hot Required hot Recuired OPERABLE OPERABLE

c. Incperative N.A. N.A.

d. Oc.nscale ~> 4% of RATED ~> 3% of RATED THERMAL POWER THERMAL POWER

3. Peactor Vessel Steam Dome Pressure - Hign 5 1037 psig 1 1057 psig

4. Peactor vessel Water Level - Low, Level 3 3 12.5 inches abw e instrument 3 11.0 inches above zero* instrument zero

5. Main Steam Line Isolation Valve - Closure 2 8% closed 2 12% closed

6. Main Ste w Line Radiation - High 1 3.0 x full cower background 3 3.6 y full power background 7 Drywell Pressure - High 3 1.68 psig $ 1.88 psig

8. Screm Discharge Volumo Water Level - High

a. Level Transmitter : 260' 9 5/8" elevation ** < W 5 5/8" elevation

b. Float Switch 3260'95/8" elevation ** 5 2 k 4 5/8" elevation

9. Turbine Stop Valve - Closure 1 5% clcsed 1 7% closed

10. Turbine Control Valve Fast Closure, Trip 011 Pressure - Low 1 00 5 psig 3 465 psig

11. Reactor Mode Switch Shutdown Position N.A. N.A.

12. Manual Scram N.A. N.A.

See Bases Figure B 3/4.3-1.

- Eq41 valent to 25.45 gallons / scram discharge volume.

LIMERICK - UNIT 1 2-4

3/4.2 POWER DISIRIBUTION tiAllS 3/4.2.1 AVI. RAGE PLANAR l!NEAR HEAT GENIRAll0N RATE LIMITING CONDil10N FOR OPLRAll0N L

3.2.1 All AVERAGE 'LANAR LlHIAR llEAT GENERATION RATES (APlllGRs) for each type of fuel as a function f axla) location and AVERAGE PLANAR EXPOSURE shall be within limits based on r, 1 cable APilIGR limit values which have been approved for the respective fuel an 'a.ttice types for two recirculation loop operation. When hand calculations are .. ired, the APlllGR for each type of fuel as a function of AVfRAGE PLANAR EXPOS ME shall not ex(.ced the limiting value for the most limiting lattice (excluding natural urunlun) as shown in the applicable figures for BP/P8X8R and GE8X8EB fuel types. The limits shall be reduced to a value of 0.09 times the two recirculation loop operation limit when in single rei.irculation loop operation.

APPL ICABil liY : OPERAil0NAL CONDITION 1, when IllERMAL POUER is grea:er than or equal to 25% of RATED THLRRAL POWER.

ACT 10ll:

With an APLHGR exceeding limiting value, initiate corrective accion within 15 minutes and restore APlllGR to within the required limits with'.n 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce

,:ERMAL POWER to less than 25% of RAIED TilERHAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE AlQUIREMENTS 4.2.1 All APil!GRs shall be vertfled tc be equal to or less than the limiting value:

a. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,

b. Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months aften completion of a TilERMAL POWER increaar of at least 15% of RAIED TillRMAL POWER, and

c. Initlally and a* least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating with a LIMITING CONIROL R00 PAllERN for APlilGR.

d. lhe provisions of Specification 4.0.4 are not applicable.

LIMIRlCK - UNil 1 3/4 2-1

-l ~,

POWER PISTRIBulldN IIMliS 3/4.2.2 APRM SEifCINTS 1,lMITING CONDlil0N LOR OPERA 110H 3.2.2 The APRM flow biased neutron flux-upscale scram trip setpoint (S) and flow biased neutron flu oupscale control rod block trip setpoint (Spg) shall be established according to the following relationships:

1 RIP SETP0lHT ALLOWA8LC VALUE OurIng two retirculatlon S 3 (0.58W + 59%)T 5 $ (0.58W e 62%)T loop operation Spg 3 (0.58W + 50%)T SRB $ (0.58W + 53%)T During single recirculation S S (0.58W + 54%)I S $ (0.58W + 57%)T loop operation Spn 1 (0.58W + 4b%)1 SRB 1 (0.58W + 48%)T where: S and Spa are in percent of RATED TilERMAL POWER, W= Loop recirculation flow as a percentage of the loop recirculation flow which produces a rated core flow of 100 million Ibs/hr.

I= Lowest value of the ratio of FRACTION OF RATED TI:ERMAL POWER divided by the CORE MAXlMUM lRACil0N Of LlHITING POWER DENSITY. 1 is applied only if less than or equal to 1.0.

APPLICABitilY: OPERATIONAL CONulTION 1, when THERMAL POWER is greater than or equal to M % of R4Thi'1HERftAL POWER.

ACTION:

With the APRM flow biased neutron flux-upscale scram trip setpoint and/or the flow biased neutron flur-apscale control rod block trip setpoint less conservative than the value shown in the Allowable Value column for S or SRB, as above determined, initiate corrective action within 15 minutes and adjust S and/or Spa to be consistent with the Trip Setpoint values

within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or reduce TilERMAL POWER to less than 25% of RAll0 illERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILtANCE REQUIREMENTS 4.2.2 lhe IRIP and the MflPD shall be determined, the value of T calculated, and the most rnent actual APRM flow biased neutron flux-upscale scram and flow biased neutron Ilus-upscale cont rol rod block t rip :.etpoints verified to be within the above limits or adjusted, as required:

d. At least ance per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,

b. Within l? hours af ter t empletion of a TilEPMAL POWER increase of at least 15% of RAll0 IH[RMAL POW (R, and C. Initidlly and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reac* is operating with MflPD greater than or equal to FRIP.

J. lhe provisions of 5pec t f (cation 4.0.4 are not applict',

With MftPD gre. iter than the IRIP, rather than adjusting the APRM setpoints, the APRM qain may be adjusted such that tN APRM readings are greater than or equal to 100% times Mi t Pit . provided that the adjusted APPM reading does not esceed 100% of RATED lht R$1At POWIR and a notice of adjustment is posted on th reactor control pc lel, tIMERifr tmli 1 3/4 2 7

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NOTE: THESE LIMITS APPLY TO BOTH TWO RECIRCUL ATION LOOP AND SINGLE RECIRCUL ATION LOOP OPERATION Dt t W'T IONS gf. sw:mt&sto Coat rLow 19* t0 Tese.mattet LEtt) FitDwattt TEMP.htbuCT40N. Mt& TING ACMitvic OUTtr08 atMOVSitvict AL TMe0VCWCVT 07 FitDWAlth ME&TER(SI) CT*Lt (U* TD 8 t*F F'WTP FM&L tttDW&Tth TEM *tRATULt ht0UCTs0N$1AT Act ENp* Mt&TEF08+dVCLE S) IV' TO 8 0'F itMF, htSVCT@H,4CMit vtD ST RIMOv&L OF ALL 8 i

MINIMUM CRITIC AL POWER R ATIO (MCPR) VERSUS T (PBXER/B FIGURE 3.2.3-1a LIMERICK - UNIT 1 3/4 2-10

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NOTE: THESE LIMITS APPLY TO BOTH TWO RECIRCULATION LOOP AND SINGLE RECIRCULATION LOOP OPER ATION Pt t twivt0NS G'.* INCEt&sto Coat FL0w tur TO telt. m&TtD) f MD0s t tEDw&Ttt NE ATINC Cut ce simV1Ct twoo,powout Cv0tt (W10 t t*F TEMP. ALDutit0N. AtwitvtD tv aspov&L 0 F g towatt a wtattats))

F# Wyt F sNAL F ttDwat ta igupta&tyst Af DUCTION at EWy'*0f stact *CvCLt Ntattas) tu* YO S t'F true.mtDvCvioN.atuitvte gr atwov&L 0r att s MINIMUM CRITIO AL POWER RATIO (MOPR) VERSUS T (GE8 X8EC FUE FIGURE 3.2.3-1b LIMERICK - UNIT 1 3/4 2-10a e

TABLE 3.3.6-2 '

CONTPOL P00 BLOCK INSTRUMENTATION SETPOINTS ._

FR8P FUNCTION TRIP SETPOINT ALLOWABLE VALUE * ' -

1. POD BLOCr MONITOP

a. Upscale

1) During two recirculation loop operation a) Flee Biased * $ 0.65 W + 41%. with a maximum ef, 1 0.66 W 4 44%, with a maximum of, b) High Flow Clamped 2 107% I 110r

2) During single recirculation loop operation a) Flow Biased

1 0.66 W + 35%, with a maximum of, 1 0.66 W + 38%. with a maximum of, b) High Flow Clamped $107*, 2 110%

b. Inoperative N.A. N.A.

c. Downscale 3 5% c' DATED THEDFAL POVFD ; 3" c' DATED THEDMa' POWED .

2. APRM

a. Flow Biased Neutron Flux - Upscale

1) During two recirculation loop operation 1 0.58 W + 50%* 1 0.58 W + 53%* : c-

2) During single recirculation loop operation 1 0.53 W + 45%* 2 0.58 W + 48%*

b. Inoperative N.A. N.A.

c. Downscale > 4% of RATED THERMAL POWER > 3% of RATED THERMAL POWER

d. Neutron Flux - upscale, Startup i12%ofRATEDTHERMALPOWER 514%ofRATEDTHERMALPOWER

3. SOURCE RANGE MONITORS d

a. Detector not fuli in N.A. N.A.

b. Upscale < 1 X 105 cps < 1.6 X 105 cp3

c. Inoperative N.A. N.A.-

d. Downscale 1 3 cps ** 1 1.8 cps ** s ,

d. INTERMEDIATE RANGE MONITORS i a. 02tector not full in N.A. N.A.

b. Upscale 1 108/125 divisions of '1 110/125 divisions of l full scale full scale

c. Inoperative N.A. N.A. ~

d. Downstale 1 5/125 divisions of full scale 1 3/125 divisions of full scale

i. SCRAM DII, CHARGE VOLUME

a. W:ter Level-Hign 1 257' 5 9/16" elevation *** $ 257' 7 9/16" eievation

a. Float Switch

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, LIMERICK - UNIT 1 3/4 3-60 ,

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

o TAB'.E 3.3.6-2 (Continued) ,e.

C0!1 TROL R00 BLOCK Irl5TRUttEf1TATIO?! SETroINTS TRIP FU!!CTI0ft TRIP SETPolttT ALLOWABLE VALUE

6. REACTOR C00LAf1T SYST[li RECIRCUL ATIO!!

Ti~UW

a. Upscale < 111% of rated flow < 114% of rated flow

b. Inoperative N.A. N.A.

c. Comparator 1 0%1 flow deviattor. I 11 % flow deviation

7. REACTOR !10DE SWITCil SIIUIDOWii it. A. N.A.

TOSITIDil ine rod block function varies as a function of recirculation loop drive flow (W). The trip setting of the Average Power Range f1onitor rod block function must be maintained in accordance with Specification 3.2.2.

- liay be reduced to 0.7 cps provided the signal,-to-noise ratio is > 2.

- - Equivalent to 13 gallons / scram discharge vohme.

3/4 3-60a LiftERICK - Uilli 1 m

3/4.4 REP.CTOR COOLANT SYSifM r 3/4.4.1 RECIRCULA110N SYSTEM RECIRCULAI10N l00PS LlHiflNG C0itDlil0N FOR OPERAll0N 3.4.1.1 Two reactor coolant system recirculation loops shall be in operation with:

a. Total core flow greater than or equal to 45% of rated core flow, or

b. TitERMAL POWER within the unrestricted :one of Figure 3.4.1.1-1. l APPLICADillFv: OPERATIONAL CONDjil0NS 1*'and 2*.

ACTION:

a. With one reactor coolant system recirculation loop not in operation:

-1. Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; a) Place tha recirculation flow control system in the local Manual mode, ar d b) Reduce illERMAL POWER to < 70% of RATED THERMAL POWER, and, c) Reduce the Maximum Average Planar Linear Heat Generation Rate (MAPlilGR) limit to a value of 0.83 times the two recirculation loop operation limit per Specification 3.2.1, and, d) Limit the speed of the operating recirculation pump to less than or equal to 90% of rated pump speed, and c) Verify that the differential temperature requirements of Survelliance Requirement 4.4.1.1.5 are met if TliERMAL POWER is 2 30% of RAIED TitERMAL POWER or the recirculation loop flow in the operating loop is 3 50% of rated loop flow, or suspend the illERMAL POWLR or recirculation loop flow increase.

See Spetlal Test Exception 3.10.4.

L ittt Rif t - Ottil I 3/4 4-1

4, ".'

REACTOR C00LANI 5) STEM-LIMIIINGlCONDil10N FOR OPERAI10H (Continued) 1 ACTION: (Continued)

2. Within 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months s:

Reduce the Average Power Range Monitor (APRM) Scram and Rod Block, dr;l Rod Block Monitor Trip Setpoints and Allowable Values, to those applicable for single recirculation loop operation per SI'ecfications P.2.1, 3.2.2, and 3.3.6, or declare the associated channel (s) Inoperable and take the actions required by the referenced specifications, and,

3. lhe provisions of Specif.ication 3.0.4 are not applicable.

4. Otherwise be in at least HOT SHUIDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. With no reactor coolant system recirculation loops in operation, immediately initiate action to reduce THERMAL POWER such that it is not within the restricted zone of figure 3.4.1.1-1 within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , and initiate measures to place the unit in at least STARTUP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN Within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,

c. With one or'two reactor coolant system recirculation loops in operation and total core flow less than 45% but greater than 39% of rated core flow and THERMAL POWER within the restricted zone of figure 3.4.1.1-1:

1. Determine the APRM and LPRM** noise levels (Surveillance 4.4.1.1.3):

I a) At least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , and h) Within 30 minutes after the completion of a THERMAL POWER increase of at least 5% of RATED THERMAL POWER.

1 4 2. With the APRM or LPRM** neutron flux noise levels greater than three times their established basellne noise levels, within 15 1 minutes initiate corrective action te restore the noise levels within the required limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months by increasing core flow or by reducing IHERMAL POWER.

l I d. With one or two reac tor coolant system recirculation loops in operation And total core flow less than or equal to 39% and THERMAL POWER within i

the restricted zone of figure 3.4.1.1-1, withir 15 minutes initiate 1 correct.tve action to reduce THERMAL POWER to within the unrestricted

' zone of figure 3.4.1.1-1 or increase core flow to greater than 39%

within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

- Detector levels A and C of one (PRM string per core octant plus detectors A and C of one tPRM string in the center of the core should be monitored.

!MiRitI - IMii i 3/4 4-la

x o a o' RE ACIOR ( t)Gt. ANI S)SIEM Sl1RVElt LAllCL R!QUIREMENTS 1.4.1.1.1 Lath pump discharge valve shall be demonstrated OPERABLE by cycling each valve thiough at least one complete cycle of full travel during each startup* l prior to lilERMAt POWER exceeding 25% of RATED lilERHAL POWER.

l 1.4.1.1.2 Each pump HG set scoop tube mechanical and electrical stop shall be demcnstrated Oft 4AblE with overspeed setpoints i 109) and 107% respectively, of rated core flow, at least nnce per 18 months.

4.4.1.1.3 tstabitsh a baseline APRM and I.PRH** neutrun flux noise value within the regions for which monitoring is required (Specification 3.4.1.1. ACTION c) within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of entering the region for which monitoring is required unless baselining has previously been performed in the region since the last refueling oute.ge.

4.4.1.1.4 With one reactor coolant system recirculation loop not in operation, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months verify that:

4. Reactor lilEPHAL POWER is 3 /0% of RAIED lilERMAL F0WER,

b. lhe recirculation flow ccatrol system is in the local Manual mode, and

c. The speed of the operating recirculation pump is 5 90% of rated pump speed.

d. Core flow is greater than 39% when THERHAL POWER is within the restricted inne of figure 3.4.1.1-1.

4.4.1.1.5 With one reactor toolant system retirculatton loop not in operation, witt in 15 minutes prior to either litERHAL POWER increase or recirculation loop ficw increase, ver ify that the following dif ferential ten ~' iture requirements are met if Il!! RMAl. I'0WER is 1 30% of RATED lilERHAL POWER or tn recirculation loop (Inw in the operating recirculation loop is 5 50% of rated loop flow:

a. _ 1450 1 between reactor vessel steam space toolant and bottom head drain line ccolant,

b. ; 50% between the reactor coolant within the loop not in operation and the coolant in the reactor pressure vessel, and

c. ; 50 i between the reactor coolant within the loop not in operallon and the operating loop, lhe illt f erent ini ten,perature requirements of Specifi(at ion 4.4.1.1.5b. and c. do not apply when the loop not in operat ion is isolated f rom the reactor pressure vessel.

If not perf ormed within tht- previous 31 days.

- Detc1 tar levels A and C of one IPRM string per core octant plus detectors A anitt' or' one tPRM string in the center of the tore should tie monitored, l IMt k lf t UN!! 1 J/4 4-2

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REACidR C00LANI SYSTEM JEI PUMPS LlHITING CONDITION FOR OPERAil0N

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3.4.1.2 All jet pumps shall be OPERABLE.

APPL ICABit 11Y: OPERATIONAL 00Nulil0NS 1 and 2.

ACTION:

With one or nere jet pumps inoperable, be in at 1 cast H0T SliUT00WN within 12 hnurs.

50RVE li t ANCE REQd!REHENTS 4.4.1.2 All jet pumps shall be demonstrated OPERABLE as follows:

a. During two recirculation loop operation, each of the above required Jet pumps shall be demonstrated OPERABLE prior to TilERMAL POWER c<ceeding 25% of RATED TilERMAL POWER and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months while greater than 25% of RATED THERMAL POWER by determining recirculation loop flow, total core flow and diffuser-to-lower plenum differential pressure for each jet pump and verifying that no two of the following conditions occur when both recirculation loop indicated flows are in compliance with Specification 3.4.1.3.

1. The indicated recirculation loop flow differs by more than 10%

from the established

pump speed-loop flow characteristics.

2. The indicated total core flow differs by more than 10% from the established

total core flow value derived from recirculation loop flow measurements.

3. Ilic indicated dif f user-to-lower plenum dif ferential pressure of dny individudl jcl pump differs from the established

patterns by more than 10%.

To be determined from the startup test program data, tIMIRICL - IINII I 3/4 4-4

1 s, .,

Rf ACIOR COOLANI SYST[H SURVEILLA U REQUIREMEHIS (Continued)

b. During single recirculation loop operation, each of the above required jet pumps shall be demonstrated OPERABLE at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying that no two of the following conditions occur:

1. The indicated recirculation loop flow in the operating loop differs by more than 101 from the established

pump speed-loop flow characteristics.

2. The indicated total core flow differs by more than 10% from the established

total core flow value derived from single recirculation loop flow measurements.

3. The indicated diffuser-to-lower plenum differential pressure of any individual jet pump differs from established

single recirculation loop patterns by more than 10%.

c. The provisions of Specification 4.0.4 are not applicable provided that this surveillance is performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding 25% of RATED THERMAL POWER and upon entering single recirculation loop operation.

Io be determined from the startup tast program data, tIPtPIC) - UNil i 3/4 4-4a

REACIOR COOLANT SfSTEM RECIRCULATION PUMPS tlMillNG CONUlil0N FOR OPERAi!ON 3.4.1.3 Recirculation loop flow mismatch shall be maintained within:

a. 5% of each other with core flow greater than or equal to /0% of rated core flow.

b. 10% of each other with core flow less than 70% of rated core finw.

APPL lC ABil l iY : OPERATIONAL CONDITIONS 1* and 2* during two recirculation loop operation.

AClION:

With the recirculation loop flows dif ferent by more than '.he specified limits, either:

a. Restore the recirculation loop flows to within the specified limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or

b. Shutdown one of the recirtulation loops within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and take the ACTION required by Specification 3.4.1.1.

SURVElLLANCE RLyulREHENTS 4.4.1.3 Recirc.ulation loop flow mismatch shall be verified to be within the limits at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

$ec Special lest Exception 3.10.4 LIMERIO - UN11 1 3/4 4-5

e, .

2.1_ SAILlY LlHllS BASLS 2.0 (HIR000CiloH lhe fuel (.udding, reactor pressure vessel and primary system piping are the principle hartiets to the release of radioactive materials to the environs.

Safety Limits are established to protect the integrity of these barriers during normal plant operations and anticipated transients. The fuel cladding integrity Safety limit is set such that no fuel damage is calculated to occur if the limit is not violated. Because fuel damage is not directly observable, a step-back approach is used to establish a Safety Limit such that the MCPR is not less than 1.0/ f or two retirculation loop operation and 1.08 for single recirculation loop operation. MCPR greater than 1.07 for two recirculation loop operation and 1.08 for single recirculation loop operation represents a conservative margin relative to the conditions required to maintain fuel cladding integrity. The fuel cladding is.cne of the physical barriers which separate the radioactive materials from the environs. The integrity of this cladding barrier is related to its relative ,

freedom from perforations or cracking. Although some corrosion or use related cracking may occur during the lif t of the cladding, fission product migration from this source is incrementally (umulative and continuously measurable, fuel cladding perforations, however, can result from thermal stresses which occur from reactor operation significantly above design conditions and the Limiting Safety !

System Settings. While fission produce migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration. l Therefore, the fuel cladding Safety limit is defined with a margin to the conditions ahlth would produce onset of transition holling MCPR of 1.0. These conditions represent a significant departure f rom the condition intended by design '

for planned operation.

?.l.1 IHIRMAL POWER. Low Pressure or low f low the use of the (GEXL) correlation is not valid for all critical power colsulations at pressures below 185 psig or core flows less than 10% of rated flow. Therefore, the fuel cladding integrity Safety Iimit is established by other means. This is done by establishing a limiting condition on core THERHAL POWER with the following basis. Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and flows will algoys be greater than 4.5 psi. Analyses show that with a bundle flow of 28 x 10 lb/h, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Igus,thebundleflowwitha4.5pstdrivingheadwillbe greater than 78 x 10 lb/n. Full scale AllAS test data taken at pressures from 14.7 psia to 8J0 psia Indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt. With the design peaking f actors, this corresponds to a 1HERMAt POWER of more than 50X of RAIED THERMAL POWER. Thus, a THERMAL POWER limit of 25% of RAIED THERMAL POWER for reactor pressure below 785 psig is conservative.

LlHIRiiK - UNil i B ?- 1

O, *-

3/4.1 RfACTIVlif CON 1ROL SYSifM5 BASES 3/4.1.3 CON 1 Rut. R005 The specification of this section ensure that (1) the minimum Silul00WN MARGIN is maintained, (?) the control rod insertion times are consistent with those used in tne accident analysis, and (3) the potential effects of the rod drop accident are limited. lhe ACT!ON statements permit variations from the basic requirements but at the same time impose more restrictive criteria for continued operation. A limitation on inoperable rods is set such that the resultant efrect on total rod worth and scram shape will be kept to a minimum.

The requirements for the various scram time measurements ensure that any indication of systematic problems with rod drives will be investigated on a timely basis.

Damage within the control rod drive mechanism could be a generic problem, therefore with a control rod immovable because of excessive friction or mechanical interference, operation of the reactor is limited to a time period which is reasonable to determine the cause of the inoperability and at the same time prevent operation with a large number of inoperable control rods.

Control rods that are inoperable for other reasons are permitted to be taken out of service provided that those in the nonfully-inserted position are consistent with the SilVIDOWN MARGIN requirements, lhe number of control rods permitted to be inoperable could be more than the eight allowed by the specification, but the occurrence of eight inoperable rods could be indicative of a generic problem and the reactor must be shutdown for investigation and resolution of the problem.

1 the control rod system is designed to bring the reactor subcritical at a rate fast enough to prevent the MCPR from becoming less than the fuel cladding safety limit during the limiting power transient analyzed in Section 15.2 of the FSAR. lhis analysis shows that the negative reactivity rates resulting from the scram with the average response of all the drives as given in the specifications, provide the required protection and MCPR remains greater than the fuel cladiiing safety limit. The occurrence of scram times longer than those specified should he viewed as an indication of a systemic problem with the rod drives and therefore the surveillance interval is reduced in order to prevent operation of the reactor for long periods of time with a potentially serious problem, lhe scram .lischarge volume is required to be OPlRABLE 50 that it will be available when needed to accept discharge water from the Control rods during a reactor scram and will i so l a t ta the reactor Coolant System from the containment hen required, j Control rods with inoperable accumulators are declared inoperable and Specif icat ion 3.1.3.1 then applies. This prevents a pattern of inoperable accumulators that would result in less reactivity insertion on a scram than has been analyzed even though control e ods with inoperable accumulators may still be inserted with normal drive water pressure. Operability of the accumulator ensures that there is a means available to insert the control rods even under the most unfavr,rable depressurization of the reactor.

! I IMW l',e 0411 i li i/4 1-?

L 3L4.? POWER DlhiklBUTION LlHils BA$f5 3p1 'e .1 AVLRAGt PLAHAR I IN[ AR _Ill'AT GENE RA110H RAIE lhis specift(allun assures that the peak cladding temperature (PCT) '

folinainq the postulated design basis loss-of-Coolant Accident (LOCA) will not etceed the limits specifico in 10 Cf R $0.46 and that the fuel design analysis limits specified in HE0f-240ll-P-A (Reference ?) will not be exceeded. <

Hechanical Design Analysis: HRC approved methods (specified in Ref erence ?) are used to demonstrate that all fuel rods in a lattice operating at the buunding power history, meet the fuel design limits specified in Rettrence J. No single fuel rod follows, or is capable of following, this boundi ng power history. This bounding power history 15 used as the basis for the fu0l design analysis MAPlilGR limit.

LUCA inalysis: A LOCA analysis is performed in accordance with 10 CfR i 50 Appendix K to demonstrate that the permissible planar power (MAPLHGR) limits (cmply with the !CCS limits specified in 10 CfR 50.46. The analysis is

performed for the most limiting break size, break location, and single failure combination for the plant.

lhe technical Specif ication MAPtitGR limit is the inost limiting n*posite of the fuel mechantral design analysis MAPlHGR and the ECCS HAPLliGR limit.

Only the most and least Itm ting MAPlHGR values are shown in the lecnnical Specitications fcr uwltiple lattice fuel. Compliance with the spetitir lattite MAPillGR operating limits, whirh are available in Reference 3, is ensured by use of the process computer, lht MAPtllGR limits shall be reduced to a salue of 0.89 times the two recirculation loop operaticn limit when in single recirculation loop operation.

The constant factor 0.89 is derived from LOCA analyses initiated from single loop operation to account fnr earlier boiling transition at the limiting fuel mode compared in the standar:1 LOCA evaluations.

L lhlk il l llN i l I li 3/4 2-1

4 *

]/ l.2 POWE R lilSIRIBUTION t lHilS BASES 3 / -1. P . !- tPhe)*RTPolNTS I

ibe f uel lattiling integrit r Saf ety Limits of Specification 2.1 were based on i a puer distribut ion which would yield the design LHGR at RATER, TilERMAL POWER. i lhe flow hiasen neutron flux-upscale scram trip setpoint and flow biased neutron flux-upscale control rod blotk functions of the APRM instruments must be adjusted to ensure that the MCPR does not become less than the Safety Limit HCPR or that . If plastic strain tioes not occur in the degraded situation. The scram and rod block setpoints are adjusted in accordance with the formula in this i spot ific ation s. hen the combination of 1HERMAL POWER and CHTLP0 indicates a higher ;>eare.1 poner distribution to ensure that an tilGR transient would not be inc reased in t he degraded (cndit icn. )

l l

l I

1 1

i ;;b i iJi ini I : it 1/i /-?

s . '.

POWL R DISIRIBui10H LIPills BASES 1

3/4.2.3 HlHIMUM CRiflCAL POWlR RATIO lhe required operating limit MCPRs at steady-state operating conditions as spet ified tri Specification 3.?.3 are derived from the established fuel cladding integrity Saf ety Limit MCPR, and an analysis of abnormal operational transients.

for any abnormal cperating transient analysis evaluation with the initial (onditions ci the reactor being at the steady-state operating limit, it is required that the resulting HCPR dot-s not decrease below the Safety Limit MCPR at any time durinq the transient assuming instrument trip setting given in Specif ication 2.2 to assur( that the fuel (ladding integrity Safety Limit is not exceeded during dny anticipated abnormal operational transient, the most limiting transients have been analyzed to determine which result in the largest reduction in CRITICAL POWER Pall 0 (CPH). the type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease, lhe evaluation of a given transient begins with the system initial parameters shown in f5AP lable 15.0-2 that are input to a GE-core dynamic behavior transient computer program. ihe codes used to evaluate transients are discussed in Pef erence ?.

The purpose of the tr factor of figure 3.2.3-2 is to define operating limits at other than rated core flow conditions. At less than 100% of rated flow ttie required HLPR is the product of the MCPR and the Ky f actor. The Kr factors assurt t hat the Safety Limit MCPR will not be violated durir.g a flow increase transient resulting from a motor-generator speed control failure. The ti fattcrs may be applied to both manual and automatic ficw control modes.

the Ly ractors values sho.n in figure 3.2.3-2 were developed generically aton are applit able to all EWP/?, thJP/3, and BWR/4 reactors. The Kr factors were dertved using the ilow cnntrol line corresponding to RATED THERHAL POWER at rated tere flon.

f or the manual flow contrul mode, the Kr f actors were calculated such that f or the masimum f low rate, as limited by the pump scoop tube set point and the corresponding 1HERMAL POWER along the rated flow control line, the limiting bundle's relative power was adjusted until the MCPR changes with different core flows. lhe t at to of the MCI R cairulated at a qtven point of core flow, divided by the operating limit MCPR, determines the Kr.

l !9t h h K tw i ! ; b 3/4 P-4

- 'o 3/4.4 REAC10R COOLANT SYSIEH BA$l5 3/4.4.1 RECIRCUL All0N SYSTEM lhe impac" of single recirculation loop operation upon plant safety is assessed and shows that single-loop operation is permitted if the MCPR fuel cladding safety limit is increased as noted by Specification 2.1.2, APRM scram and control rod block setpoints are adjusted as noted in Tables 2.2.1-1 and 3.3.6-2, respectively, and MAPtitGR limits are decreased by the f actor given in Specification 3.2.1.

Additionally. surveillance on the pump speed of the operating recirculation loop is imposed to exclude the pussibility of excessive internals vibration. The surveillance on differential temperatures below 30% RAILD THERMAL POWER or 50% rated recirculation loop flow is to mitigate the undue thermal stress on vessel nozzles, recirculation pump and vessel bottom head during the extended operation of the single recirculation loop mode.

An inoperable jet pump is not, in itself, a sufficient reason to declare a recirculation loop inoperable, out it does, in case of a design-basis-accident, increase the blowdown area and reduce the capability of reflooding the core; thus, the requirement for shutdown of the facility with a jet pump inoperable. Jet pump failure can be detected by monitoring jet pump performance on a prescribed schedule for significant degradation.

Recirculation pump speed mismatt h limits are in compliance with the ECCS LOCA analysis design criteria for two recirculation loop operation. The limits will ensure an adequate core flow coastdown from either recirculation loop following a LOCA. In the case where the mismatch limits cannot be maintained during two loop operation, continued operation is permitted in a single recirculation loop mode.

Inordertopreventunduestressonthevesselnozglesandbottomheadregion,the recirculationlooptemperaturesshallbewithin50Fofeachotherprjorto startup of an idle loop. The loop temperature nost also be within 50 f of the reactor pressure vessel coolant temperature to prevent thermal shock to the recirculation pump and recirculation nozzles. Sudden equal 12ation of a temperature difference > 145 F between the reactor vessel bottom head coolant and the coolant in the upper regicn of the reactor vessel by increasing core flow rate would cause undue stress in the reactor vessel bottom head.

The objectives of GE BWR plant and fuel design is to provide stable operation with margin over the normal operating domain, llowever, at the high power / low flow corner of the operating domain, a small probability of limit cycle neutron flux oscillations exists depending on combinations of operating conditions (e.g., rod pattern, power shape). 10 provide assurance that neutron flux limit tycle oscillations are detected and suppressed, APRM and LPRH neutron flut noise levels should be monitored while operating in this region.

Stability tests at operating BWRs were reviewed to determine a generic region of the power / flow map in wnich surveillance of neutron flut noise levels should be performed. A conservative decay ratio of 0.6 was chosen as the bases for determining the generic region for surveillance to account for the plant to plant variability of decay rat to with tore and f uel designs. This generic region has been determined to correspond to a core flow of less than or equal to 45% of rated core flow and a lilE RHAt POWE R greater than that spec ified in figure 3.4.1.1-1.

Plant specific calculations can ne performed to determire an applicable region for monitoring neutron flua noise levels. In this case the degree of conservatism Con be reduced since plant to plant variability would be eliminated, in this case, adequate margin will be assured by monitoring the region which has a decay ratio greater than or equal to 0.8.

i t MU

e u. ; ! . B 2 4 4- 1

UNITED STATES OF AMERICA NUCLEAR REGULATORY CottilSSION In the Matter of : Docket No. 50-352 PillLADELPillA ELECIRIC CottPANY :

(Limerick Generating Station, :

Unit No,1)

CfRTIFICATE OF SIRVICE I hereby certify that copies of the foregoing Application for Amendment of Facility Operating License NPF-39 in the above captioned matter were served on the following by deposit in the United States Mail, first class postage prepaid, on the 4th day of November , 1988.

William T. Russell, Regional Administrator U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406 T. J. Kenny U.S. Nuclear Regulatory Coinissicn Senior Resident inspector P.O. Box 47 Sanatoga, PA 19464 Thomas Gerusky, Director Bureau of Radiological Protection Department of Environmental Resources P.O. liox 2003 liarrisburg, PA 17120

' 7-Eugene J 7 radley I Attorney for Philadelphia Electric Cocpany

ML20205Q974

Text

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