Text

Proposed Tech Specs Eliminating Majority of License Amend Requests Due to Changes in Value of cycle-specific Parameters,Resulting in Resource Savings for Util & NRC
	ML20244D997
Person / Time
Site:	Clinton
Issue date:	06/12/1989
From:	; ILLINOIS POWER CO.
To:	;
Shared Package
ML20244D990	List: ML20244D997; U-601465, Application for Amend to License NPF-62,in Response to Generic Ltr 88-16, Removal of Cycle-Specific Parameter Limits from Tech Specs. Amend Would Result in Resource Savings by Eliminating Majority of License Amend Requests;
References
	NUDOCS 8906200098
	Download: ML20244D997 (43)
	v • d • e

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U-601465-l LS-88-039 Page S of 47 INDEX

1. 0 DEFINITIONS SECTION PAGE 1.1 ACTI0N......................................................... 1-1 1.2' AVERAGE PLANAR EXPOSURE........................................ 1-1 1.3 AVERAGE PLANAR LINEAR HEAT GENERATION RATE..................... 1-1 1.4 CHANNEL CALIBRATION............................................ 1-1 1.5 CHANNEL CHECK.................................................. 1-1

1. 6 CHANNEL FUNCTIONAL TEST........................................ 1-1 -

1. 7 CONTAINMENT AND REACTOR VESSEL ISOLATION CONTROL SYSTEM RESPONSE TIME.................................................. 1-2

1. 8 CO R E A LT E RAT I O N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-2 1.9 cott on.%Tnsk LtMITS Rt.po4r 1 -2.

1.10 lef CRITICAL POWER RATI0........................................... 1-2

%.(t IA6 DOSE EQUIVALENT I-131.................................. ....... 1-2

1. turtT O RYWE LL I NTEG R I TY. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-2 1131df E - AVERAGE DISINTEGRATION ENERGY.............................. 1-3 1 14 M 3 EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME............. 1-3 1651d f END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME....... 1-3 3

Q.150ELETED.............................. . ..................... 1-3) 1.16 DELETE 0........................................................ 1-3 1.17 FREQUENCY N0TATION............................................. 1-4 1.18 GASEOUS RA0 WASTE TREATMENT SYSTEM.............................. 1-4 1.19 IDENTIFIED LEAKAGE............................................. 1-4 1.20 LIMITING CONTROL. ROD PATTERN................................... 1-4 1.21 LINEAR HEAT GENERATION RATE.................................... 1-4 1.22 LOGIC SYSTEM FUNCTIONAL TEST................................... 1-4 1.23 0ELETED........................................................ 1-4 g62OOO99990612

p ADOCK 05000461 PNu CLINTON . UNIT 1 i

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U-601465

LS-88-039 Pags 6 of 47 -

INDEX. 1 i

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS !

SECTION PAGE' -

REACTIVITY CONTROL SYSTEMS (Continued) ;

3/4.1.5 STANDBY LIQUID CONTROL SYSTEM..........................'... 3/4 1-19 ,

Figure 3.1.5-1 Weight Percent Sodium Pentaborate Solution as a Function of Net Tank Vo1ume.............................. 3/4 1-21

]

3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE.... p ......... 3/4 2-1 Figure 3.2.1-1 fFlow-Oependent MAPLHGR Factors (MAPFAC ) .W4(TM.. 3/4 2-2 f

Figure 3.2.1-2 Power-Dependent MAPLHGR Factors (MAPFAC, . M .N ?. 3/4 2-3 , i o.

Figure 3.2. - Vlaximum Average Planar Linear Heat Generation Rate) >

i MAPLHGR)

Core Versus Fuel Types - High Average Enrichments. Planar .Exposure ........

Initiat@El:E@. 3/4'2-4

f. Figure 3.2.1-4 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus Average Planar Exposure Initial Core Fuel Types - Medium Enrichment....................... 3/4 2-4A A '

Figure 3.2.1-5 Maximum Average Planar Linear Heat Generation Rate (

(MAPLHGR) Versus Average Planar Exposure Initial l !

Core Fuel Types - Natural Enrichment....................... 3/4 2-48 Figure 3.2.1-6 Maximum Average Planar Linear' Heat Generation Rcte (MAPLHGR) Versus Average Planar Exposure-Reload 1 Fuel Type 8P85RB284L........................................... 3/4 2-4C-Figure 3.2.1-7 Maximum Average Planar Linear Heat Generation Rate ,

I (MAPLHGR) Versus Average Planar Exposure-Reload 1 Fuel Type BP85RB284LC.......................................... 3/4 2-4D 3/4 2.2 DELETED................................................... 3/4 2-5 p 3/4.2.3 MINIMUM CRITICAL POWER RATIO. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 2-7 Figure 3.2.3-1 (C_linton MCPR, Versus Core Flo .9f.L.F.Tf. p. . . . . . . . . . 3/4 2-8 .

Fi.gure 3.2.3- Clinton MCPR Versus Power for AT< 50 F and CoreI 3/4 2-9 ow < 107% . 3 g ..................................

3/4.2.4 LINEAR HEAT GENERATION RATE............................... 3/4 2-10 3/4.3 INSTRUMENTATION 3/4.3.1' REACTOR PROTECTION SYSTEM INSTRUMENTATION................. 3/4 3-1 Table 3.3.1-1 Reacto'r Protection System Instrumentation. . . . . . . . . . . 3/4 3-3 Table 3.3.1-2 Reactor Protection System Response Times............. 3/4 3-7 I Table 4.3.1.1-1 Reactor Protection System Instrumentation Surveillance Requirements.......................... 3/4 3-8 3/4.3.2 CONTAINMENT AND REACTOR VESSEL ISOLATION CONTROL SYSTEM... 3/4 3-11 Table 3.3.2-1 CRVICS Instrumentation............................... 3/4 3-13 I

CLINTON - UNIT 1 v

4. ~ . .

U-601465 LS-88-039 Page 7 of 47 INDEX ADMINISTRATIVE CONTROLS SECTION PAGE L REVIEW AND AUDIT (Continued)

Meeting Frequency............................................ 6-10 Quorum........................................................ 6-11 Review....................................................... 6-11 Audits....................................................... 6-11 Authority..................................................,. 6-13 Records...................................................... 6-13 6.5.3 TECHNICAL REVIEW AND CONTR0L................................. 6-13 Activities................................................... 6-13 6.6 REPORTABLE EVENT ACTI0N........................................ 6-14 6.7 S AFETY LIMIT VIO LATI ON, . . . . . . . . . . . . . . . . . . . . . . . . . . .' . . . . . . . . . . . . . 6-14 6.8 PROCEDURES AND PR0 GRAMS........................................ 6-15 6.8.1 PROCEDURES................................................... 6-15 6.8.2 REVIEW AND APPR0 VAL.......................................... 6-15 6.8.3 TEMPORARY CHANGES............................................ 6-15 6.8.4 PR0 GRAMS.....,............................................... 6-16 6.9' REPORTING REQUIREMENTS......................................... 6-17 6.9.1 ROUTINE REPORT 5.......................... ................... 6-17 Startup Report............................................... 6-17 Annual Reports............................................... 6-17 Annual Radiological Environmental Operating Report........... 6-18 Semiannual Radioactive Effluent Release Report............... 6-19 Monthly Operating Reports.................................... 6-21 Coll OftRhT*Ma LsMSTs RlPOR3 (a -2,1.

CLINTON - UNIT 1 xxv

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U-601465 LS-88-039 Page 8 of 47 DEFINITIONS CONTAINMENT AND REACTOR VESSEL ISOLATION CONTROL SYSTEM RESPONSE TIME

1. 7 The CONTAINMENT AND REACTOR VESSEL ISOLATION AND CONTROL SYSTEM (

RESPONSE TIME shall be that time interval from-when the monitored parameter exceeds its isolation actuation setpoint at the channel sensor until the isolation valves travel to their required positions. Times shall include diesel generator starting and sequence loading delays where applicable. The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time.is measured.

CORE ALTERATION

1. 8 CORE ALTERATION shall be the addition, removal, r.elocation or movement of fuel, sources, incore instruments or reactivity controls within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Normal .

movement of the SRMs, IRMs, or TIPS, or special movable detectors, is not con-sidered a CORE ALTERATION. Suspension of CORE ALTERATIONS shall not preclude g,,e completion -w of the movement of a componen,t to a safe conservative position. .

CRITICAL POWER RATIO l

l.%Ok9" The CRITICAL POWER RATIO (CPR) shall be the ratio of that power in the l assembly which is calculated by application of an approved General Electric Critical Power correlation to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

DOSE EQUIVALENT I-131 t

1.11 .MD DOSE EQUIVALENT I-131 shall be that concentration of I-131, microcuries l l per gram, which alone would produce the same thyroid dose as the quantity and 1

isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present.

The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites."

i DRYWELL INTEGRITY 1017. JMORWELL INTEGRITY shall exist when: (

l

a. All drywell penetrations required to be closed during accident conditions are either: 3

1. Capable of being closed by an OPERABLE drywell automatic isolation i

system or

2. Closed by at least one manual valve, blind flange, or deactivated automatic valve secured in its closed position, except as provided in Table 3.6.4-1 of Specification 3.6.4.

b. The drywell equipment hatch is closed and sealed.

c. The drywell airlock is OPERABLE pursuant to Specification 3.6.2.3.

CLINTON - UNIT 1 1-2

U-601465 LS-88-039 Pcg2 9 cf 47 Insert for Page 1-2 CORE OPERATING LIMITS REPORT 1.9 The CORE OPERATING LIMITS REPORT is the' Clinton-specific document that provides core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be determined for each. reload cycle in accordance with Specification 6.9.1.9. Plant operation within these operating limits is addressed in individual Specifications.

U-601465 LS-88-039 Page 10 of 47 DEFINITIONS DRYWELL INTEGRITY (Continued) d.-

The drywell leakage rates are within the limits of Specification 3.6.2.2.

e. The suppression pool is OPERABLE pursuant to Specification 3.6.3.1.

f.

The sealing mechanism associated with each drywell penetration, e.g.,

welds, bellows or 0-rings, is OPERABLE.

E - AVERAGE DISINTEGRATION ENERGY 1.13 ME shall be the' average, weighted in proportion to the concentration of l each radionuclides in the reactor coolant at the time of sampling, of the sum of the average beta and gamma energies per disintegration, in Mey, for isotopes,.

with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in +.he coolant.

EMERGENCY CORE COOLING SYSTEM (ECCS)' RESPONSE TIME 1 14 W The EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME shall interval from when the monitored parameter exceeds its ECCS actuation'setpoint at the channel safety functionsensor until the ECCS equipment is capable of performing its charge pressures; reach their required values,Times etc.i.e., the valves trzel to their re shall include diesel generator starting and sequence loading delays where applicable. The response time may be measured by any series of sequential ~, overlapping or total steps such that the entire response time is measured.

I END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME 1 15 M T The END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONS l that time interval to complete suppression of the electric arc between the l fully open contacts of the recirculation pump circuit breaker from initial l movement of the associated:

a. Turbine stop valves and

b. Turbine control valves.

The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.

1.15 ((:LETC] t 1.16 [ DELETED] .

t CLIN 70N - UNIT 1 1-3 L__l____-------------

U-601465

- LS-88-039 Page 11 of 47 3/4.2 POWER DISTRIBUTION LIMITS .

3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) for each type of fuel as a function of AV E PLANAR EXPOSURE shall not exceed the limits ras determined below:(spe.cl in thc. C, ORE. OPER AT"pJ4 Limin Re.PonT.

a. During two recirculation loop operation - the limits shown in Figures 3.2.1-3 through 3.2.1-7 multiplied by the smaller of either]

the flow-dependent MAPLHGR factor (MAPFAC ) of Figure 3.2.1-1 or the -

power-dependent MAPLHGR factor (MAPFAC )

p f Figure 3.2.1-2.

b. During single recirculation loop operation - the limits shown in .

Figures 3.2.1-3 through 3.2.1-7 multiplied by the smallest of

( MAPFAC f , MAPFAC, or 0.85.

APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.

spec.6fle) in Me. CORE OPERAruJ4: Lamr5RE%e'$

With an APLHGR exceedino the limitsfof Figures 3.2.1-3 throuch 3.2.1-7. asY (multiplied by the appropriate multiplication factor,finitiate corrective action within 15 minutes and restore APLHGR to within the required limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS 4.2.1 All APLHGRs shall be verified to be equal to or less than the required limits:

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 after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER,

c. Initially and at 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 CONTROL ROD PATTERN for APLHGR, and

d. The provisions of Specification 4.0.4 are not applicable.

I CLINTON - UNIT 1 3/4 2-1 l

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U-601465 LS-88-039 Page 12 of 4'7' ,

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0.6 0.5 0.4 0 20 40 60 80 100 120 CORE FLOW (% reted). F Figure 3.2.1-1 Flow-Dependent MAPLHGR Factors (MAPFACf)

DELE.TE D CLINTON - UNIT 1 3/4 2-2

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r U-601465 LS-88-039 Pags 19 of 47 POWER DISTRIBUTION LIMITS' 3/4.2.3 MINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION 3.2.3 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be equal to or greater s than both MCPR and f MCPR p limits at indicated core flow and THERMAL POWER and AT* as shown in Ficures 3.2.3-1 and 3.2.3-2.[%c limih seusGal in Mc CORE OPE RAT M(a t.iMars REPORT *.

APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.

^ "

limits spaded in we coRL oPERAT:M6 L.sMirs Ef

a. With MCPR less than the[a'pplicable MCPR limit shown in F1aures 3.2.3-1 ) -

(snd 3.2.3-23 initiate corrective action within 15 minutes and restore MCPR to within the required limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS 4.2.3 MCPR shall be determined to be equal to or greater than the limits (determined from Ficura< 3 7. 3-1 and 3. 2. 3-ZD ver.1fic.d in the CORE OPER ATsNG LIMir:5 RE96RT :

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 after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER,

c. Initially and at 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 CONTROL R0D PATTERN for MCPR, and

d. The provisions of Specification 4.0.4 are not applicable.

1

^This-AT refers to any reduction of rated feedwater temperature 420*F, such as prolonged removal of feedwater heater (s) from service.

1 CLINTON - UNIT 1 3/4 2-7 i

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U-601465 LS-88-039 -

Page 20 of 47' 7-~

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1.0 0 20 40 60 80 100 120 CORE FLOW (% ratedt, F Figure 3.2.3-11ClintonHCPRf Versus Core Flow DELETED l CLINTON - l NIT 1 3/4 9-8

1

..= . U-601465 LS-88-039 Page 21 of 47

'Gi.5 Past inhohonally left blank, f 2.4 ,

~~

- THERMAL POWER'25% s P s 40%

/ CORE FLOyV > 50% ;

l 2.2 j MCPRp = 2.10 + 0.0033_(40, , , . . - P) 7

% [_,,

\( ,,, THERMAL POWER 25% s P s 40%

CORE FLOW s 50%

2.0 f 6 s e MCPRp = 1.85 + 0.0133 (40 - Pi i -

i e '

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t i T

i 1.8 W

THERMAL POWER 40% < P 5 70%

1.6 , ' MCPRp = 1.43 + 0.0045 (70 - P1 v

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/ ,%

MCPRp = 1.18 + 0.00591100 - P) 1.0

O 20 40 60 80 100 120 CORE THERMAL POWER (% rated), P Figure 3.2.3-21Clinton HCPRp Versus Power for AT $ 500F cod Core Tiow 1107%

D ELE.TE D CUNTON - UNIT 1 3/4 2-9

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Page 22 of 47

. s. -

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.d . ,

POWER DISTR!!UTION LIMITS 2/4.2.4 ' LINEAR HEAT GENERATION RAT ~ J LIMITING CONDITION FOR OPERATION a

3.2.4 The LINEAR HEAT GENERATION RATE (LHGR) shall net exceed (13.4 kW/f t.I 4.hs, l;m;f .3pdsed i in t4w CoE.E OPERATn APPLICA!!LITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or W M RE NET equai to 25a ,of RATED THERMAL POWER. .

ACTION: .

With the LHGR of any fuel rod exceeding the limit, initiate corrective acticn -

within 15 minu es and restore the LHGR to within the limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or recu:e THERXAL POWER to less than 25% of RATED THERMAL POWER within the next l 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

i SURVEILLANCE REQUIREMENTS 4.2.4 LHGRs shall be determined to be equal to or less than the limit:

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

- a. 4

b. Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after co ole' tion of a THERhAL -

POWER increase of at least 15% of RATED THERMAL POWER, and

c. Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating on a LIMITING CONTROL RCD PATTERN for LHGR.

c. The provisions of Specification 4.0.4 are not applicable. _

. I

\

l

. 1 i.

O CLINTON - UNIT 1 3/4 2*10 -

p .- U-601465. -

LS-88-039 Page 23 of 47 7.-

l 3/4.4 REACTOR COOLANT SYSTEM = I y'

3/4.4.l' RECIRCULATION SYSTEM RECIRCULATION LOOPS LINITING CONDITION FOR OPERATION '

3.4.1.1 with: Two-reactor coolant system recirculation loops'shall be in ' opera, tion -

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

b. THERMAL POWER within the unrestricted zone of Figure 3.4.1.1-1, or

c. THERMAL POWER within the restricted zonet of Figure 3.4.1.1-1 and APRM -

or.LPRMtT noise levels not larger than three times their established baseline noise levels.

APPLICABILITY OPERATIONAL CONDITIONS 1" and 2*.

' ACTION:

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

1. Within 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months s:

a) Place the recirculation flow control system in the Local Manual (Position Control) mode, and -

b) Reduce THERMAL POWER TO $ 70% of RATED THERMAL POWER, and c) Increase the MINIMUM CRITICAL POWER RATIO (MCPR) Safety Limit by 0.01 to 1.08 per Specification 2.1.2, and d) Reduce the Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Ifmit per Specification 3.2 3, and cand W ColtCOPEthnMG Lamars g e) Reduce the Average Power Range Monitor (APRM) Scram and Rod-Block Trip Setpoints and Allowable Values to those applicable for single-recirculation-loop operation per Specifications 2.2.1 and 3.3.6, and

See Sp'ecial Test Exception 3.10.4.

tThe operating regio'n for which monitoring is required. See Surveillance Requirement 4.4.1.1.2.

ttDatector levels A and C of one LPRM string per core octant plus detectors A and C of one LPRM string in the center of the core should be monitored. ,

CLINTON - UNIT 1 3/4 4-1

U-601465 LS-88-039

. Pega 24 of 47 3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section assure that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the 2200 F limit specified in 10 CFR 50.46.

3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE The peak cladding temperature (PCT) following a postulated loss-of-coolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is dependent only secondarily on the rod to rod power distribution within an assembly. The peak clad tempera-ture is calculated assuming a LHGR for the highest powered rod which is equal to or less than the design LHGR corrected for densification. This LHGR times 1.02 is used in the heatup code along with the exposure dependent steady state gap conductance and rod-to-rod local peaking factor. The Technical Specifica- -

tion AVERAGE PLANAR LINEAR HEAT GENEP.ATION RATE (APLHGR) is this LHGR of the highest powered rod divided by its local peaking factor. The limiting value for APLHGR is the MAPLHGR. apeaf4ed in As CORE OPERAD4 LIMITS REPcAT The MAPLHGR limits h Figures 3.2.1-3 throuch 3.2.1-7 re multiplied by the smaller of the flow-dependent MAPLHGR factor (MAPFAC f

) or the power-dependent MAPLHGR factor (MAPFAC p

) corresponding to existing core flow and power conditions to assure the adherence to fuel mechanical design bases during the most limiting transient (Reference 2). The MAPFAC p factors are determined using the three-dimensional BWR simulator code to analyze slow flow runout transients. The maximum runout flow settings of 102.5% and 109% include design allowances for recirculation flow instrument uncertainties (2.5% and 2.0% respectively) to ensure that the rated flow conditions of 100% and 107%

can be achieved. The MAPFAC p factors are generated using the same data base i as the MCPR p to protect the core from plant transients other than core flow l runout. -

pfnJ m M CME oPERATnMe La Mars Re.PcR.T Thecalculationalprocedu'eusedtoestablishthe'APLHGRYshownonFioures[

r (3.2.1-3 throuch 3.2.1-8 based on. a loss-of-coolant accident analysis. The analysis was performed using General Electric (GE) calculational models which i are consistent with the requirements of Appendix K to 10 CFR 50. A complete discussion of each code employed in the analysis is presented in Reference 1.

l Differences in this analysis compared to previous analyses tan be broken down as follows.

a. Input Changes

1. Corrected Vaporization Calculation - Coefficients in the vaporization correlation used in the REFLOOD code were corrected.

2. In;;rporated more accurate bypass areas - The bypass areas in the top l guide were recalculated using a more accurate technique.

3. Corrected guide tube themal resistance.

4. Correct heat capacity of reactor internals heat nodes.

CLINTON - UNIT 1 B 3/4 2-1

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[. .. U-601465 ,

LS-88-039 .

l Page 25'of 47-POWER'0 DISTRIBUTION' LIMITS

' BASES n.

13/4.?.1' AVERAGE PLANAR LINEAR HEAT GENERATION RATE (Continued)- i l

b .~ Model Chance .

1. Core CCFL pressure. differential - l' psi - Incorporate, the assumption.

that flow from the bypass to lower plenum must overcome a.1. psi pressure drop in core.

2. Incorporate NRC pressure transfer assumption - The assumption used in the SAFE-REFLOOD pressure transfer when the pressure is increasing' was changed.

A few of the changes. affect the accident calculation ' irrespective of CCFL. .

These changes are. listed beTow.

a. Input Change

1. Break ' Areas - The ~ DBA break area was calculated more accurately.

b. Model' Change

1. -Improved Radiation and Conduction' Calculation - Incorporation of CHASTE 05 for_ heatup calculation.

A list of the significant plant input parameters to the loss-of-coolant accident analysis is' presented in Bases Table B 3.2.1-1. '

spes.bsd in k. CDRL dtRRTg@ UMITS

- tor plant operation with a single recirculation loop, the MAPLHGR' limit Rt W (Figures 3.z.1-4 rnrouah 3.2.1-77fft multiplied by the smallest of MAPFA MAPFAC p or 0.85 (Reference.2). Theconstant-factor,0.85,isderivedfr$m,LOCA-analyses,. initiated from single loop operation to account for earlier boiling transition at tha-limiting fuel. node compared to standard LOCA evaluations. -

3/4.2.2 APRM SETPOINTS '[ DELETED] , ,

S. O g ,.

e .

4

~

CLINTON - UNIT 1 8 3/4 2-2

U-661465 LS-88-039 Page 26 of 47

~ BASES TABLE'8 3.2.1-1.

SIGNIFICANT INPUT PARAMETERS TO THE LOSS-OF-COOLANT ACCIDENT ANALYSIS

Plant Parameters:

Core THERMAL POWER . . . . . . . . . . . . . . . . . . . . 3015 MWt** which correspor ds to 105% of rated steam flow Vessel Steam Output ...................

13.08 x 108 lb,/hr which corresponds to 105% of rated steam flow Vessel Steam Dome Pressure. . . . . . . . . . . . . 1060 psia Design Basis Recirculation Line ..

Break Area for: .

~

a. Large Breaks 2.2 ft 2, b.' Small Breaks 0.09 ft.

Fuel Parameters:

PEAK TECHNICAL INITIAL SPECIFICATION DESIGN MINIMUM LINEAR HEAT ~

AXIAL CRITICAL FUEL BUNDLE GENERATION RATE PEAKING POWER FUEL TYPE GEOMETRY (kW/ft) FACTOR RATIO Initial and 8x8 W 1.4 1.17***

Reload Cores

A more detailed listing of input of each model and its source is presented in Section II of Reference 1 and Section 6.3 of the FSAR. *

- This power level meets the Appendix.K reoutrement of 102%. The core heatup calculation assumes a bundle power consistent with operation of the highest powered rod at 102% of its Technical Specification LINEAR HEAT GENERATION RATE limit.

- - For single recirculation loop oparation, loss of nucleate boiling is assumed at 0.1 seconds after a LOCA, regardless of initial MCPR. For core i flows less than 85% of rated, the initial MCPR is taken from the MCPR Curv l 7

speobed in ut CORE. OPERATidG Limits Refostr

This VAkt 15 SPeoged in Hu CORE 09ERArtN6 Limits, ggpoq7, t

CLINTON ' UNIT 1 - B 3/4 2-3

U-601465 LS-88-039 .

Page 27 of 47 POWER DISTR 1eUTION LIMITS BASES .

3/4.2.3 HINIMUM CRITICAL POWER RATIO TherequiredoperatinglimitMCPpsatsteadystateoperatingconditionsas specified in Specification 3.2.3 are derived from the established fuel cladding l integrity Safety Limit MCPR in Specification 2.1.2, and an analysis of abnormal operational transients. For any abnormal operating transient analysis evalua-tion with the initial conditi6n of the reactor being at the steady state operating limit it is required that the resulting MCPR does not. decrease below the Safety Limit MCPR at any time during the transient assuming instrument trip setting given in Specification 2.2.

To assure that the fuel cladding integrity Safety , Limit is not exceeded during .

any anticipated abnormal operational transient, the' most 1,imiting transients -

have been analyzed to detemine which result in the largest reduction in CRITICAL POWER RATIO (CPR). The type of transients evaluated were loss of flow, increase in pt essure and power, positive reactivity insertion, and coolant temperature decrease. The power-flow maps of Figures. 8 3/4.2.3-1 or 8 3/4.2.3-2 give operational limits for double or single recirculation loop operation, respectively.

The evaluation of a given transient begins with the system initial parameters identified in Reference 3 that are input to a GE-core dynamic behavior transient computer program. The codes used to evaluate pressurization a'nd non pressurization events are described in Reference 3. The principal result of this evaluation is the reduction in MCPR caused b the transient.

spe.c.@ad a +he. CORE OPN" L N n Rf>o $

The purpose of the MCPR f and MCPR Q Figures 3.2.3-1 and 3.2.3-2)is to define operating limits at other than rated core flow and power conditions. At less than 100% of rated flow and power the required MCPR is the larger value of the NCPR 7 and MCPR p

at the existing core flow and power state. The MCPR s are 7

established to protect the core from inadvertent core flow increases such that the 99.9% MCPR limit requirement can be arsured.

The MCPR f s were calculated such that for the maximum core tiow rate and the

~

corresponding' THERMAL POWER along the most limiting" power flow control line, the limiting bundle's relative power was adjusted until the MCPR was slightly above the Safety Limit. The maximum' runout flow settings (109% and 102.5%)

include design allowances for recirculation flow instrument uncertainties (2%

and 2.5% respectively) to ensure that the rated flow conditions (107% and 100%) can be achieved. Using this relative bundle power, the MCPRs were cal-culated at different ' points along the most limiting power flow control line corresponding to different core flows. The calculated MCPR at a given point j of core flow is defined as MCPR7.

ik %c. vale assouM se&h +kis bmi& art- 4mbed M A'COREdEOD L twun Rtfo*I l

CLINTON - UNIT 1 8 3/4 2-4 1

U-601465 LS-88-039 Page 28 of 47

^

3/4.4 REACTOR COOLANT SYSTEM l l

BASES 3/4.4.1 RECIRCULATION S STEM-The impact of single recirculation loop operation upon plant safety is assessed 1 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 a $d in Tables 2.2.1-1 and 3.3.6-2, respectively MAPLHGR limits are decrease v the_fe tne civen in Saecifica- h Rion 3.2.1,Nnd MCPR operating limits are adjustedTer Section 3/4.2.3.)n--

Additionally, surveillance on the volumetric flow rate of the operating recir-culation loop is imposed to exclude the possibility of excessive core internals vibration. The surveillance on differential temperatures below (30%)* 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 re-circulation loop inoperat ie, but 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 pre-scribed schedule for significant degradation. Significant degradation is indicated if more than one of three specified surveillance performed confirms unacceptable deviations from established patterns or relationships. The surveillance, including the associated acceptance criteria, are in accordance with General Electric Service Information Letter No. 330, the recommendations of which are considered acceptable for verifying jet pump operability according to NUREG/CR-3052, "Closecut of IE Bulletin 80-07: BWR Jet Pump Assembly. Failure." Performance of the specified surveillance, however, is not required when thermal power is less than 25% RATED THERMAL POWER because flow esci11ations and jet noise precludes the collection of. repeatable I meaningful data during low flow conditions approaching the threshold response of the associated flow instrumentation.

i Recirculation loop flow mismatch limits are in compliance, with 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 ifaits cannot be maintained during two loop operation, continued operation 'is permitted in a single-recirculation loop mode.

In order to prevent undue stress on the ve'ssel nozzles and bottom head region, the recirculation loop temperatures shall be within 50'F of each other prior to startup of an idle loop. The loop temperature must also be within 50*F of the reactor pressure vessel coolant temperature to prevent thermal shock to the l recirculation pump and recirculation nozzles. . Sudden equalization of a tempera- I ture difference > 100*F between the reactor vessel bottom head coolant and the coolant in the upper region of the reactor vessel by increasing core flow rate would cause undue stress in the reactor vessel bottom head.

" Initial Values. Final values to be determined during Startup Testing based on the threshold THERMAL POWER and recirculation loop flow which will sweep the cold water from the vessel bottom head preventing stratification.

f In occonkru. w% % vaW, 94uhed in A. Cort. ofu rd Limrs RE90RT' CLIKTON - UNIT 1 8 3/4 4-1

U-601465

- LS-88-039 _.

Page 29 of 47 ADMINISTRATIVE CONTROLS SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (Continued)

The Semiannual Radioactive Effluent Release Reports shall include a list and description of unplanned releases from the site to UNRESTRICTED AREAS of radio-active materials in gaseous and liquid effluents made during the reporting period.

The Semiannual Radioactive Effluent Release Reports shall include any chang'es made during the reporting period to the PROCESS CONTROL PROGRAM (PCP) and to the OFFSITE DOSE CALCULATION MANUAL (ODCM), pursuant to Specifications 6.13 and 6.14, respectively, as well as any major changes to liquid,' gaseous, or Solid Radwaste-Treatment Systems pursuant to Specification 6.15. It will also include a list-ing of new locations for dose calculations and/or environmental monitoring iden- -

tified by the land use census pursuant to Specification 3.12.2.

MONTHLY OPERATING REPORTS

6. 9.1. 8 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the main steam system safety /

relief valves, shall be submitted on a monthly basis to the Director, Office of Resource Management, U.S. Nuclear Regulatory Commission, Washington, D.C.

20555, with a copy to the Regional Administrator of the Regional Office of the NRC, no later than the 15th of each month following the calendar month covered by the report.

-nSut ----- >

SPECIAL REPORTS 6.9.2 Special rgorts shall be submitted to the Regional Administrator of the Regional Office 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 of Title 10, Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated.

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

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

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

c. All REPORTABLE EVENTS.

~

d. Records of surveillance activities, inspections, and calibrations required by these Technical Specifications and the Fire Protection Program.

CLINTON - UNIT 1 6-21

U. '4 .

U-601465 LS-88-039 Paga 30 of 47 Insert for Page 6-21 CORE OPERATING LIMITS REPORT 6.9.1.9 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC'in General Electric Standard Application for Reactor Fuel (GESTAR), NEDE-24011-P-A-8, as ,

amended and Maximum Extended Operating' Domain and Feedwater Heater l Out-of-Service Analysis for Clinton Power Station, NEDC-31546P, August 1988.

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

. REPORT, including any mid-cycle revisions or supplements thereto, shall be provided to the NRC Document Control Desk with copies to the Regional .

Administrator and Resident Inspector within 30 days after the report (including revisions and supplements) is issued.

-4: .

I i U-601465 LS-88-039 i Page 31 of 47 i.

4 ILLINOIS POWER COMPANY CLINTON POWER STATION O

CORE OPERATING LIMITS REPORT FOR RELOAD 1 CYCLE 2 REVISION O

c -_. . -__,--- _ . - _ _ - _ - - - _ _ _ - _-- _-- - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . - _ _ _ - _ _ _ _ _ _ _ - _ _ - _ _ _

f:

, U-601465 LS-88-039 Page 32 of 47'

, -INDEX 1.0 DEFINITIONS 2.0 POWER DISTRIBUTION LIMITS-2.1 AVERAGE PIANAR LINEAR ' HEAT GENERATION RATE (Technical Specification 3/4.2.1) 2.2 MINIMUM-CRITICAL POWER RATIO (Technical Specification 3/4.2.3) ,

2.3 LINEAR HEAT GENERATION RATE (Technical, Specification 3/4.2.4) 4

n. ,

U-601465 LS-88-039 Page 33 of 47 1.0 DEFINITIONS O

1 l-p i

L . . _ . . _ _ _ . _ . - - - _ . . _ . _ _ _ _ - . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___

U-601465 1.S-88-039 Page 34 of 47 l

1.1 The CORE OPERATING LIMITS REPORT is the Clinton-specific document that provides core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6,9.1.9. Plant operation within these operating limits is addressed in individual specific'ations, f

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b. U-601465 LS-88-039 i

Page 35 of 47 2.0 POWER DISTRIBUTION LIMITS t

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H 2.1 ' AVERAGE PIANAR LINEAR HEAT GENERATION' RATE All AVERAGE PIANAR LINEAR HEAT GENERATION RATES (APLHGRs) for each type.of fuel as a function of AVERAGE Planar Exposure shall not exceed the limits as determined below;

a. During two recirculation loop operation - the limits shown in figures 2.1-3.through 2.1-7. multiplied by the smaller of either the flow-dependent MAPLHGR factor (MAPFAC f ) of Figure

. 2.1-1 or the. power-dependent MAPLHCR factor- (MAPFAC p ), of Figure:2.1-2.

b. During single recirculation loop operation - the limits shpw n in Figures 2.1-3 through 2.1-7 multiplied by the smallest of MAPFACg ,'MAPFAC or 0.85.

p This' limit applies to Technical Specification-3/4.2.1 and associated. Bases.

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0.4 0 20 40 60 80 100 120 CORE FLOW (% retee. F .

Figure 2.1-1 Flow-Dependent MAPLEGR Factors (MAPFACf )

CUNTON - UNIT 1

U-601465-

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LS-88-039

, Page 44 of 47 -i 2.2- MINIMUM CRITICAL POWER RATIO n e. MINIMUM CRITICAL POWER RATIO (MCPR) shall be equal to or-greater than both MCPRf and MCPP, limi'ts at indicated core flow and THERMAL POWER and AT as'shown ih rigures 2.2-1 and 2.2-2.

i These limits aPfly to j Technical Specification 3/4.2.3 i and associated Bases.

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This .4T refers to any reduction of rated feedwater temperature l (420*F), such as prolonged removal of feedwater heater (s).from service.

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L L i i t \ j FOR MAX RUNOUT FLOW SETTING = 109.0% --

T s _< MCPRf a MAX (1.18.1.841 -0.00707F) .

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I'3 FOR MAX RUNOUT FLOW SETTING = 102.5% T '

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MCPRg e MAX (1.18,1.746-0.00669F)

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O 20 40 90 80 100 120 CORE FLOW (% rated). F

. Figura 2.2-1 Clinton ICPRg Versus Core Flow CLINTON - UNIT 1

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< LS-88-039-Page 46 of'47.

77 2.4 4 ,

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-THERMAL POWER 25% s P s 40%

'f CORE FLOP > 50%

2.2 j MCPRp = 2.10 + 0.0033 (40 - Pt 7 , ,

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,,, THERMAL POWER 25% s P s 40%

CORE FLOW s 50%

2.0 y r

" MCPRp = 1.85 + 0.0133 (40 - PI -

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' THERMAL POWER 40% < P s 70% .--..-_-

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i Figurs 2.2-2 Clinton HCP Versus Power for 5T $ 500F and Core Flow < 107%

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23 . LINEAR HEAT GENERATION. RATE- j i

The LINEAR HEAT GENERATION RATE (LHCR) shal1~not exceed ~13.4 i KW/ft. l

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