Text

Proposed Tech Specs Re Max Extended Operating Domain
	ML20154A203
Person / Time
Site:	Clinton
Issue date:	09/06/1988
From:	; ILLINOIS POWER CO.
To:	;
Shared Package
ML19297G893	List: ML20154A203; ML20154A214; U-601239, Application for Amend to License NPF-62,revising Tech Spec to Support Cycle 2 in Max Extended Operating Domain & W/ Reduced Feedwater Temps.Supporting Documentation Encl.Fee paid.NEDC-31546P Withheld;
References
	NUDOCS 8809120108
	Download: ML20154A203 (66)
	v • d • e

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. Attech=nt 2 to U-60l?39 page 3 of 58 Technical Technical Sp;cification Specification Reason ffar Page No. Section Description of Change ,,g.3 nge_,

B3/4 2-7 Fig. B3/4.2.3-1 Replace the current operating map with FTVD a new operating map.

B3/4 2-8 Fig. B3/4.2.3-2 Add an operating map for single (~.arititar.on (new) recirculation loop operation.

B3/4 6-5 B3/4.6.2.5 Revise the value for drywell peak MEOD calculated pressure from 18.9 psig to 19.7 psig.

GSL/bjc RFP3/GSL10 I

t I

b$k $$$$$f9 pdc

to U-601239 Pcgo 6 of $8

. ]NDEX 1.0 DEFINITIONS PAGE SECTION 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............................l............... 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 CORE ALTERATION................................................ 1-2 1.9 CRITICAL POWER RATI0........................................... 1-2 1.10 DOSE EQUIVALENT I-131..........................................

1-2 1.11 DRYWELL INTEGRITY.............................................. 1-2 -

1.12 E - AV E RAG E D I S I NT E G RAT I O N E N E RGY. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 1.13 EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME............. 1-3 1.14 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME...... 1-3 1.15 fmTION Of LIMITINC SIR DENSITd'M.'ATF.9. . . . . . . . . . . . . . . . . 1-3 1.16 FR',CTION Of RATED '9E RMAL P0'CifI. f? .E.4 F75 P. . . . . . . . . . . . . . . . . . . 1-3 1.17 FREQUENCY N0TATION.............................................

1-4 1.18 GASEOUS RADWASTE TREATMENT SYSTEM.............................. 1-8 1.19 IDENTIFIED LEAKAGE............................................. 1-4 I

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 MAM MUH-FRAC 410N-OF--L-lMITI NG-POWEMENd. 9.U.Ey,g, , , , , , , ,1-4 ,,

l CLINTON - UNIT 1 i 1

1

Attcchment 2 to U-601239 Paga 7 of 58 INDEX DEFINITIONS SECTION PAGE 1.48 UNRESTRICTED AREA.............................................. 1-9 1.49 VENTILATION EXHAUST TREATMENT SYSTEM........................... 1-9 1.50 VENTING........................................................ 1-9 Table 1.1 Surveillance Frequency Notation.......................... 1-10 Table 1.2 Operational Conditions................................... 1-11 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS THERMAL POWER, Low Pre ssure or Low F1ow. . . . . . . . . . . . . . . . . . . . . . . . 2-1 THERMAL POWER, High Pressure and High Flow..................... 2-1 Reactor Coolant System Pressure................................ 2-1 Reactor Vessel Water Leve1..................................... 2-1 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Protection System Instrumentation Setpoints............ 2-2 1

' Table 2.2.1-1 Reactor Protection System Instrumentation Setpoints.. 2-3 BASES 2.1 SAFETY LIMITS Introduction................................................... B 2-1 THERMAL POWER, Low Pres sure or Low F1ow. . . . . . . . . . . . . . . . . . . . . . . . B 2-1 THERMAL POWER High Pressure and High F1ow..................... B 2-2 Reactor Coolant System Pressure................................ B 2-2 0.. . T;bi; O 2.1.4-A un;e r 6a int :;; L';;f in 'h; 0:te- ' st t;r, ef'# ~

th: E;l Ciaa ms 5aissy L i rui tt . 9.h .L4 T.E,0, , , , , , , , , , , , , , , B 2-3 CLINTON - UNIT 1 iii

to U-601239 !

Pago 8 of 58 INDEX BASES 9

SECTION PAGE SAFETY LIMITS (Continued) 1;;;; T:b' 8 91 S-? 'e i :1 n V:1ue: ef Par ete= L'ted 4- th: *"

St.ti Toel Cle A:nv Inty it, 0;f;;., ~

41,l.qjjetAr.ai,i,:,vi.....QF4CC4A...................................... B 2-4 ReactorVesselWaterLevei..................................... B 2-5

2. 2 LIMITING SAFETY SYSTEM SETTINGS Reactor Protection System Instrumentation Setpoints............ B 2-6 3.0/4.0 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.0 APPLICABILITY................. .............................. 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4,1.1 SHUTDOWN MARGIN........................................... 3/4 1-1 3/4.1.2 REACTIVITY ANOMALIES. .................................... 3/4 1-2 3/4.1.3 CONTROL RODS Control Rod Operability................................... 3/4 1-3 Control Rod Maximum Scram Insertion Times................. 3/4 1-6 Control Rod Scram Accumulators............................ 3/4 1-9 Control Rod Drive Coup 1fng................................ 3/4 1-11 Control Rrf Posi tion Indication. . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-13 Control Rod Drive Housing Support......................... 3/4 1-15 3/4.1.4 CC/JROL ROD PROGRAM CONTROLS .

Control Rod Withdrawal.................................... 3/4 1-16 Rod Pattern Control System................................ 3/4 1-17 CLINTON - UNIT 1 iv

ateccQcont 8

to U-601239 Pogo 9 of 58

~

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIRENENTS 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.h POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE.............. . 3/4 2-1 Figure 3.2.1-1 Maximum Average Planar Linear Heat Generation Rate 7 (MAPLHGR) Versus Average Planar Exposure Initial Core Fuel Types - High Enrichment............................... 3/4 2-2

A" Figure 3.2.1-2 Maximum Average Planar Linear Heat Generation Rate

m a .J (MAPLHGR) Versus Average Planar Exposure Initial Core Fuel lypes - Medium Enrichment....................... 3/4 2-3 Figure 3.2.1-3 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus Average Planar Exposure Initial

( Core Fuel Types - Natural Enrichment...................... 3/4 2-4 y 3/4 2.2 #^e SCT"0!NTS.T.N.E.W.5 9.............................. 3/4 2-5 1 3/4.2.3 MINIMUM CRITICAL POWER RATI0.............................. 3/4 2-7 j Figure 3.2.3-1 Clinton MCPRf Versus Core F1ow...................... 3/4 2-8 Figure 3.2.3-2 Clinton MCPR p Versus Power.ter. 4T S.59. f. .%4 k.t.8.W.f 3/4 2-9 101 */*

3/4.2.4 i.INEAR HEAT G2NERATION RATE............................... 3/4 2-10 i

3/4.3 INSTkuMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION................. 3/4 3-1 Table 3.3.1-1 Reactor Protection Syste.li Instrumentation............ 3/4 3-3 Table 3.3.1-2 Reactor Protection System Response Times............. 3/4 3-7 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 i i

Table 3.3.2-1 CRVICS I ns trumentati o n. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

3/4 3-13 CLINTON - UNIT 1 v

Att chment 2

. to U-601239 Pcg3 10 of 58 Insert "A" to Index Figure 3.2.1-1 Flow-Dependent MAPLHGR Factors (MAPFACg ) ................... 3/4 2-2 Figure 3.2.1-2 Power-Dependent MAPLHGR Factors (MAPFAC ) ................... 3/4 2-3 Figure 3.2.1-3 Maxt.um Average Planar Linear Heat Generation Rate (MAPLHGR) Versus Average Planar Exposure Initial Core Fuel Types - High Enrichment . . . . . . . . . .. 3/4 2-4 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 Figure 3.2.1-5 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus Average Planar Exposure Initial Core Fuel Types - Natural Enrichment . . . . . . . .. 3/4 2-4B Figure 3.2.1-6 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus Average Planar Exposure-Reload 1 Fuel Type BP8 SRB 284L . . . . . . . . . . . . . . .. 3/4 2-4C Figure 3.2.1-7 Maximum Average Planar Linear Heat Generation Race (MAPLHGR) Versus Average Planar Exposure-Reload 1 Fuel Type BP85RB284LC . . . ... . . . . . . . . .. 3/4 2-4D L

i Editorial Note l

Addition of the above figures will cause the page numbers of the index to change, i

l 6

W I

l

gggy -g g- - - ,

to U-601239 Pcgo 11 of 58 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS .--

PAGE SECTION BASES 3/4.0 APPLICABILITY................................ u.............. B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS .

3/4.1.1 SHUTOOWN MARGIN........................................... B 3/4 1-1 3/4.1.2 REACTIVITY AN0MALIES...................................... B 3/4 1-1 3/4.1.3 CONTROL R005.............................................. B 3/4 1-1 3/4.1.4 CONTROL ROD PROGRAM CONTR0LS.............................. B 3/,4 1-3 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM.............................

B 3/4 1-4 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE................ B 3/4 2-1 ir 3/4.2.2 AFRM SETP0!NTS............................................

B 3/4 2-2 i

Bases Table B 3.2.1-1 Significant Input Parameters to the Loss- B 3/4 2-3 of-Coolant Accident Analysis..............................

MINIMUM CRITICAL POWER RATI0..............................

B 3/4 2-4 3/4.2.3 LINF.AR HEAT GENERATION RATE...............................

B 3/4 2-5 3/4.2.4 Bases Figt.*e B 3/4.2.3-1 Reactor Operating Map 8pr. L9. .%.eP.vM'.eA B 3/4 2-7 Loop Optes.& ten 6aus F85ure. O s/4.2.3-2. Reader C W ia3 Map for 9

% (t. Redewkfico 63/42.-8 Lcop opwhoo CLINTON - UNIT 1- xvi

. to U-601239 l Page 12 of 58 DEFINITIONS CONTAINMENT AND REACTOR VESSEL ISOLATION CONTROL SYSTEM RESPONSE TIME 1.7 The CONTAINMENT AND REACTOR VESSEL ISOLATION AND CONTROL SYSTEM (CRVICS)

RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its isolation actuation setpoint at the channel sensor until the 1

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 t 1.8 CORE ALTERATION shall be the addition, removal, relocation or movement of ;

fuel, sources, incore instruments or reactivity controls within the reactor l

1 pressure vessel with the vessel head removed and fuel in the vessel. Normal l l 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 '

{

completion of the movement of a component to a safe conservative position. L i

CRITICAL POWER RATIO g yp rend Gened Elen. Cnha I E*W'r r

l 1.9 The CRITICAL POWER RATIO (CPR) shall be tr e ratio of that power in the assembly which is calculated by application ofF t h: C;ne al Ciectric Critica+---

j > ^ ;iity iii % Length (GH tt correlation to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating 3

power.

DOSE EQUIVALENT I-131 1.10 DOSE EQUIVALENT I-131 shall be that concentration of I-131, microcuries per gram, which alone would produce the same thyroid dose as the quantity and (

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

j The thyroid dose conversion factors used for this calculation shall be those l l listed in Table III of TID-14844, "Calculation of Distance Factors for Power -

i and Test Reactor Sites."

i DRYWELL INTEGRITY

, 1.11 ORWELL INTEGRITY shall exist when:

4

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

} l

1. Capable of being closed by an OPERABLE drywell automatic isolation system or ,

- t j 2. Closed by at least one manual valve, blind flange, or de' activated j automatic valve secured in its closed position, except as provided l

in Table 3.6.4-1 of Specification 3.6.4. ;

)

b. The drywell equipment hatch is closed and sealed.

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

CLINTON - UNIT 1 1-2 l

.i t I i

) .

Attcchment 2 to U-601239 Pegs 13 of 58 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.

I - AVERAGE DISINTEGRATION ENERGY 1.12 I shall be the average, weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling, of the sum of the average beta and gamma energies per disintegration, in Mev, for isotopes,

> with half lives greater than 15 minutes, making up at least 95's of the total non-iodine activity in the coolant.

EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME 1.13 The EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME shall be that time interval from when the monitored paranieter exceeds its ECCS actuation setpoint at the channel sensor until the ECCS equipment is capable of performing its safety function; i.e., the valves travel to their required positions, pump dis-charge pressures reach their required values, etc. 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.

. END-0F-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME i 1.14 The END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME shall be that time interval to complete suppression of the electric arc between the fully open contacts of the recirculation pump circuit breaker from initial

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 g (total steps such that the entire response time is measured, FRACTION OF LIMIilNG POWER DENSITY j

1 l

$ 3 he FRACTION OF LIMITING POWER DENSITY (FLPD) shall be the LHGR existing (itagivenlocationdividedbythespecifiedLHGRlimitforthatbundletype.g t

(FRACTION OF RATED THERMAL P3WER he FRACTION OF RATED THERMAL POWER (FRTP) shall be the measured THERM J

l POWER divided by the RATED THERMAL POWER.

[0ELEJE D]

1 g CLINTON - UNIT I 1-3 l

y to U-601239 Pega 14 of $8 DEFINITIONS FREQUENCY NOTATION 1.17 The FREQUENCY NOTATION specified for the performance of surveillance requirements shall correspond to the intervals defined in Table 1.1.

GASEOUS RA0 WASTE TREATMENT SYSTEM 1.18 A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the main condenser evacuation system and providing for delay or holdup for the purpose of reducing the~ total radioactivity prior to release to the environment.

IDENTIFIED LEAKAGE 1.19 IDENTIFIED LEAKAGE shall be:

a. Leakage into collection systems, such as pump seal or valve packing leaks, that is captured and conducted to a sump or collecting tank or

b. Leakage into the drywell atmosphere from sources that are both specifically located and known either not to interfere with the operation of the. leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE. -

LIMITING CONTROL ROD PATTERN 1.20 A LIMITING CONTROL ROD PATTERN shall be a pattern which results in the core bei g on a thermal hydraulic limit, i.e., operating on a limiting value for APLHGR, LHGR, or MCPR.

LINEAR HEAT GENERATION RATE 1.21 LINEAR HEAT GENERATION RATE (LHGR) shall be the heat generation per unit length of fuel rod. It is the integral of the heat flux over the heat transfer area associated with the unit length.

LOGIC SYSTEM FUNCTIONAL TEST 1.22 A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all logic components, i.e., all relays and contacts, all trip units, solid state logic elements, etc., of a logic circuit from sensor through and including the actuated device to verify OPERABILITY, The LOGIC SYSTEM FUNCTIONAL TEST may be per-formed by any series of sequential, overlapping or total system steps such that the entire logic system is tested.

. (EAXIMUM FRACTION OF LIMITING POWER DENSITY [

d heMAXIMUMFRACTIONOFLIMITINGPOWERDENSITY(MFLPD)sh -

j hiohest value of the FLPD which exists in the core. J I

[ MEMBER (5) 0F THE PUBLIC 1.24 MEMBER (S) 0F T'HE PUBLIC shall include all persons who are not occupational}y

, associated with the plant. This category does not include employees of the Also excluded from this category are licensee, its contractors or vendors.

1-4 CLINTON - UNIT I

to U-601239 Peas 15 of 58 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS THERMAL POWER, low Pressure or Low Flew 2.1.1 THERMAL POWER shall not exceed 25% of RATED THERMAL POWER with the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION: ,

With THERMAL POWER exceeding 25% of RATED THERMAL 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 HOT SHUTDOWN 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. . .

THERMAL POWER, High Pressure and High Flow t . 01 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than ke6 with two recirculation loop operation and shall not be less than L47Jith g. o S single recirculation loop operation with the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10% of rated flow.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION:

1.01 1 With MCPR less than 1Att with two recirculation loop operation or less than g.06 with single loop operation and the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10% of rated flow, be in at least HOT SHUTOOWN 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.

REACTOR COOLANT SYSTEM PRESSURE 2.1.3 The reactor coolant system pressure, as measured in the reactor vessel 1, team dome, shall not exceed 1325 psig.

APPLICABILITY: ODERATIONAL CONDITIONS 1, 2, 3, and 4.

ACTION: ,

With the reactor coolant system pressure, as measured in the reactor vessel steam dome, above 1325 psig, be in at least HOT SHUTDOWN 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.

REACTOR VESSEL WATER LEVEL 2.1.4 The reactor vessel water level shall be above the top of the active irradiated fuel.

CLINTON - UNIT 1 2-1

O TABLE 2.2.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS

?

~

25 FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE E

. 1. Intermediate Range Monitor C

g a. Neutron Flux-High < 120/125 divisions < 122/125 divisions if full scale of full scale

b. Inoperative NA NA

! 2. Average Power Range Monitor:

a. Neutron Flux-High, Setdown < 15% of RATED < 20% of RATED THERMAL POWER THERMAL POWER l 67

b. Flow Biased Simulated Thermal Power-High M t

1) Ducie3 % rec 8r W e leap *P u=Jsen - (*)

< 0.66 ( A (a) < 0.66 ( A

-+30J Flow Blased , ,

'? Gith a maximum of Gith a maximum of

-- ]b.,d High Flow Clamped < 111.0% of RATEC < 113.0% of RATED THERMAL P0wrn THERMAL POWER Twet. W

c. Neutron Flux-High < 118% ef RATED < 120% of RATED THERMAL POWER THERMAL POWER

d. , Inoperative NA NA

3. Reactor Vessel Steam Dome Pressure - High < 1065 psig < 1080 psig

4. Reactor Vessel Water Level - Low, Level 3 > 8.9 inches above > 8.3 inches Instrument zero* above instrument zero

l S. Reactor Vessel Water Level-High, Level 8 < 52.0 inches above < 52.6 inches above Instrument zero* Instrument zero 2gR

%c0 *a

6. Main Steam Line Isolation Valve - Llosure -< 8% closed -< 12% closed eoft', ?,-

7. Main Steam Line Radiation - High < 3.0 x full power < 3.6 x full power Eackground background o. ' co o>

Attachment 2 to U-60123) '

Page 17 of 58 Insert W for Page 2-3 -

2) During single recirculation loop operation:

a) Flow Biased i 0.66(W-AW)+48%

i 0.66(W-aW)+51%(*

b) liigh Flow Clamped Not Required Not Required OPERABLE OPERABLE f

e 4

~~~

Attcchment 2 to U-601239 Pcgo 18 of 58 2.1 SAFETY LIMITS BASES 2.

1.0 INTRODUCTION

The fuel cladding, reactor pressure vessel and primary system piping are the principal barriers 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 ess than ,

v. v . iwi . - . . . . . w . u o n s o u p v v . . . . . . . . . .. . . . , . . . . . . .. . . . . . . . . . . . . . . . . i v v y

' ;= t hr.. MCPRs greater than these safety limits represent a conservative margin relative to the conditions required to maintain fuel cladding integrity.

The fuel cladding is one 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 life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding perforations,'however, can result from thermal stresses which occur from reactor operation significantly above design-condi-tions and the Limiting Safety System Settings. While fission product migration from cladding perforation is just as measurable as that from use related crack-

. ing, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration. Therefore, the fuel cladding Safety Limit is defined with a mar-MCPR of gin to These the conditions conditions which wouldaproduce represent significant onset of transition departure from boiling,ition the cond 1.0.

intended by design for planned operation. ,

2.1.1 THERMAL POWER, low Pressure or Low Flow an appewed Guual tis 4He. CriM P w c.ersMien (Aabe=* h The use of th S L ~ ~= h " P is not valid for all critical power calcula-tions at pressures below 785 psig or core flows less than 10% of rated flow.

Therefore, the fuel cladding integrity Safety Limit is established by other means. This is done by establishing a limiting condition on core THERMAL 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 always be greater than 4.5 psi. Analyses show that with a bundle flow of 28 x 108 lbs/hr, bundle pressure drop is nearly independent of bundle power

and has a value of 3.5 psi. Thus, the bundle flow with a 4.5 psi driving head will be greater than 28 x 103 lbs/hr. Full-scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt. With the design peaking factors, this corresponds to a THERMAL POWER of more than 50% of RATED THERMAL POWER.

Thus, a THERMAL POWER limit of 25% of RATED THERMAL POWER for reactor pressure below 785 psig is conservative. ,

ne. Vaha. giVta li\ Tred h d rors M.

CLINTON - l! NIT 1 B 2-1

to U-601239 Pego 19 of 58 SAFETY LIMITS BASES 2.1.2 THERMAL POWER, High Pressure and High Flow The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated to occur if the limit is not violated. Since the parameters which result in fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions resulting in a departure from nucleate boiling have been used to mark the beginning of the region where fuel damage could occur. Although it is recognized that a departure from nucleate boiling would not necessarily result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit. However, the uncertainties in monitoring the core operating state and in the procedures used to calculate the critical power result in an uncertainty in the value of the critical p'ower. Therefore, the fuel cladding integrity Safety Limit is defined as the CPR in the limiting ft.1 assembly for which more than 99.9% of the fuel rods in the core are expected to avoid boiling transition considering the power distribution within the core and all uncertainties, hetA fTheSafetyLjmitMCPRisdetermineauniowtoeueneralElectricThermalAnalysis[

Basis, GETAB , which is a statistical model that combines all of the uncertain-ties in operating parameters and the procedures used to calculate critical power. The probability of the occurrence of boiling transition is determined using the General Electric Critical Quality (X) Boiling Length (L), GEXL, correlation. The GEXL correlation is valid over the range of conditions used in the tests of the data used to develop the correlation.

The required input to the statistical model are the uncertainties listed in Bases Table B2.1.2-1 and the nominal values of the core parameters listed in Bases Table 82.1.2-2.

The basis for the uncertainty in the GEXL correlation is given in NED0-10958 A*

and the bases for the uncertainties in the core parameters are given in 4E00-20340.** The power distribution is based on a typical 764 assembly core in which the rod pattern was arbitrarily chosen to produce a skewed power dis-tribution having the greatest number of assemblies at the highest power levels.

The worst distribution during any fuel cycle would not be as severe as the dis- J (tribution used in the analysis.

2.1.3 REACTOR COOLANT SYSTEM PRESSURE The Safety Limit for the reactor coolant system pressure has been selected such that it is at a pressure below which it can be shown that the integrity of the

"General Electric BWR Thermal Analysis Bases (GETAB) Data, Correlation and Design Application," NE00-10958-A.

- General Electric "Process Computer Perfor'mance Evaluation Accuracy,"

NE00-20340 and Amendment 1, NE00-20340-1, dated June 1974 and December 1974, respectively. _

CLINTON - UNIT 1 B 2-2

Attcchment 2

. to U-601239 l Pago 20 of 58 "A" to page B 2-2 Insert The Safety Limit MCPR is determined using a statistical model that combines all of the uncertainties in the operating parameters and in the procedures used to calculate critical powe.r. The probability of the occurrence of boiling transition is determined using the approved i General Electric Critical Power correlation. Details of the fuel cladding integrity safety limit calculation are given in Reference 1.

Reference 1 includes the tabulation of the uncertainties used in the determination of the Safety Limit MCPR and of the nominal values of

] parameters used in the Safety Limit MCPR statistical analysis.

4 Reference .

1. "General Electric Standard Application for Reactor Fuel (GESTAR)."

NEDE-24011-P-A-8 as amended.

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. to U-601239 Pcge 21 of 58 BASES TABLE B 2.1.2-1 A

UNCERTAINTIES USED IN THE DETERMINATION OF THE FUEL CLADDING SAFETY LIMIT STANDARD i DEVIATION

QUANTITY (% of Point)

. Feedwater Flow 1.76 ,

Feedwater Temperature 0.76 Reactor Pressure 0.5 Core Inlet Temperature 0.2 Core Total Flow Two Recirculation Loop Operation 2.5 !

Single Recirculation Loop Operation 6.0 . j Channel Flow Area 3.0 ;

Friction Factor Multiplier 10.0 .

Channel Friction Factor 5.0

{ Multiplier j i TIP Readings !

Two Recirculation Loop Operation 6.3 l Single Recirculation Loop Operation 6.8 l

R Factor 1.5 !

Critical Power 3.6 t

I Note: The uncertainty analysis used to establish the !

corewide Safety Limit MCPR is based on the l assumption of quadrant power symetry for the reactor core. [

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'i

. to U-601239 1 Page 22 of 52 l

{

f BASES TABLE B 2.1.2-2 7 ,

(

NOMINAL VALUES OF PARAMETERS USED IN '

f THE STATISTICAL ANALYSIS OF FUEL CLADDING INTEGRITY SAFETY LIMIT l i

A h PARAMETER VALUE THERMAL POWER 3323 MW !

I

! Core Flow 108.5 M1b/hr '

i Dome Pressure ' 1010.4 psig Chcnnel Flow Area 0.1089 ft2 i R-Factor High enrichment - 1.043 I

) Mediu:.1 enrichment - 1.039 f l k Low enrichment - 1.030 f f

j [ DEL.cT'ED3 l

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r

. Attachment 2 l

to U-601239 l . Pego 23 of $8 l LIMIf!NG SAFETY SYSTEM SETTINGS BASES l

2.2.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued)

Average Power Range Monitor (Continued) l The APRH trip system is calibrated using heat balance data taken during steady-l state coriditions. Fission chambers provide the basic input to the system and therefore, the monitors resoond directly and quickly to changes due to transient operation for the case of the Neutron Flux-High setpoint; i.e; for a power -

increase, the THERFAL POWER of the fuel will be less than that indicated by the neutron flux due to the time constants of the heat transfer associated with the i fuel. For the Flow-Biased Sirrulated Thermal Power-High setpoint, a time con-l stant of 6 2 0.6 seconds is introduced into the floicbiased APRM in order to simulate the fuel thermal transient characteristics. A more conservative inaximum value is used for the flow biased setpoint as shown in Table 2.2.1-1.

The APRM setpoints were selected to provide adequate margin for the Safety l Limits and yet Lilow operating margin that reduces the possibility of unneces-sary shutdownJ The flow referenced trip setpoint must se adjusted by the i (specified formula in Specification 3.2.2 in order to maintain these margins l

Q hen MFLPD is > FRTP.

3. Reactor Vessel Steam Dome Pressure-High l High pressure in the nuclear system could cau:e a rupture to the nuclear system I process barrier resulting in the release of fission products. A pressure i l increase during operation will also tend to increase the power of the reactor by [

compressing voids, thus adding uactivity. The trip will quickly reduce the neutron flux, counteracting the pressure increase. The trip setting is slightly higher than the operating pressure to permit normal operation without spurious l

trips. The setting provides for a wide margin to the maximum allowable design pressure and takes into account the location of the pressure measurement com-pared to the highest pressure that occurs in the system during a transient.

This trip setpoint is effective at low power / flow conditions when the turbine stop valve closure and turbine control valve fast closure trips are bypassed.

For a turbine trip or load rejection under these conditions, the transient analysis indicated an adequate margin to the thermal hydraulic limit.

1 1

4. Reactor Vessel Water Level-Low ,

The reactor vessel water level trip setpoint has been used in transient analy-ses dealing with coolant inventory decrease. The scram setting was chosen far enough below the normal operating level to avoid spurious trips but high enough above the fuel to assure that there is adequate protection for the, fuel and pressure limits. j CLINTON - LMT 1 B 2-7 l

ggg gg to U-601239 Pcg3 24 of 58 REACTIVITY CONTROL SYSTEMS CONTROL R00 MAXIMUM SCRAM INSERTION TIMES LIMITING CONDITION FOR OPERATION 3.1.3.2 The maximum scram insertion time of each control robrom the fully withdraan position, based on deenergization of the scram pilot valve solenoids as time zero, shall not exceed the fo.llowing limits:

Maximum Insertion Times to Notch Position (Seconds)

Reactor Vessel Dome Pressure (psig)* F 43 29 13 950 G G I"T4 1050 0.32 0.86 1.57 APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION:

a. With the maximum scram insertion time of one or more control rods exceeding the maximum scram insertion time limits of Specification 3.1.3.2 as deter-mined by Surveillance Requirement 4,1.3.2.a or b, operation may c'ontinue provided that:

1. For all "slow" control rods, i.e., those which exceed the limits of Specification 3.1.3.2, the individual scram insertion times do not exceed the following limits:

Maximum Insertion Tires to Noten Position (Seconds)

Reactor Vessel Dome Pressure (psig)* F 43 29 13 950 G M M 1050 0.39 1.1A 2.22

2. For "fast" control rods, i.e., those which satisfy the limits of Specification 3.1.3 2, the average scram insertion times do not exceed the following limits:

Maximum Average Insertion Times toNotchPosition(Seconds}

Reactor Vessel 0 me Pressure (psic)" 43 29 13 950 G G M 1050 0.31 0.84 1,53 f or the initial f uel cycle only, up to 32 control rods, not occupying adjacentp locations in any direction, including the diagonal, may be exented f rom scram tire testing under hot, pressurized conditions, provided they meet all other surveillance requirements, and that all scram reactivity requirements of the plant safety analysis are ret with no scram contribution from these rods.

For intercediate reactor vessel dome pressure, the scram time criteria are determined by linear interpolation at each notch position.

CLINTON - UNIT 1 3/4 1-6

Attcchment 2 ;

to U-601239 i Pego 25 of 58 i REACTIVITYCONTR0L,,QSjj!_

CONTROL R00 MAXIMLW M I d ATION TIMES i

SURVEILLANCE REQUIk % N'.5 ;

4.1.3.2 ThemaximumscraminsertiontimeofthecontrolrodNhallbedemon-strated through measurement with reactor coolant pressure greater than or equal ;

to 950 psig and, during single control rod scram time tests, the control rod j drive pumps isolated from the accumulators: >

i a. For all control rods prior to THERMAL POWER exceeding 40% of RATED THERMAL

, POWER following CORE ALTERATIONS or after a reactor shutdown that is

. greater than 120 days,

b. Forspecificallyaffectedindividualcontrolrods*Io11owingmaintenance f l l

on or modification to the control red or control rod drive system which could affect the scram insertion time of those specific control rods, and .

i j c. For at least 10% of the control rods, on a rotating basis, at least once per 120 days of POWER OPERATION. ,

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_[*For the initial fuel cycle only, up to 32 control rods, not occupying adjacent 11ocations in any direction, including the diagonal, may be exempted from scram time testing under hot, pressurized conditions, provided they meet all other j surveillance test requirements, and that all scram reactivity requirements of j i theplantsafetyanalysisaremetwithnoscramcontributionfromtheserods.f j

The provisions of Specification 4.0.4 are not applicable for entry into OPERATIONAL CONDITION 2 provided this surveillance requirement is co pleted l

j prior to entry into OPERATIONAL CONDITION 1.

] CLINTON - L' NIT 1 3/4 1-8 i

. Attachment 2 to U-601239 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

_o_f fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the limits "shown in Figures 3.2.1-1, 3.2.1-2, and 3.2.1-3.

3,2.1-2, and 3.2.1-3 shall be reduced to a value of ThelimitsofFigures3.2.1-1,]A-0.85 times the two-(recirculation loop operation limit when in single recirculation loop operation.

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

't, .s s multipt;. 4 by As a r rte pet *+ t mul4(pu t* *'* 4 f*O* r, ACTION:

(3.2.1- 3 nr..$ f.t..l With an APLHGR exceedin the limits of Figures &3.1-1, 3. 2<!-2, or 3.2-1-+t- l initiate corrective act on 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 . .

b SURVEILLANCE REQUIREMENTS ceprd 4.2.1 All APLHGRs shall be verified to be equal to or less than thea limits:

s terairad-4c- Figuree- 3. 2.1 1,-3,2.-1-2, .4 3. 2 <!- 3 A

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 THERML 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 far APLHGR, and

d. The provisions of Specification 4.0.4 are not applicable.

a. s d t. k r . % . / b I. v :

o.. Durin3 +wo cuarcu.\ wen I..p .p.eme.n. w,. \;m;4 s sh.wn in ngure.s s.z..t 3 Regg 3 'A l*l mulk \ led t by At. S me,\tt y- of e.t hte iht(luw-dtH6 dad peer dtp.ad u,+

mA9t.%R bder QuetAQ)of F gut.

gApt % & f. dor (A APrAQ of Figure 1 1.l- 2, , 5.1.\-1 er As

b. Dwrirs3 sI% l c. ettitcutdren loep operdien. Ac ItWf s shoe in Figures 3,*2,.1-5 &hr ewgh 3.2..l '1 muikp\t td b9 ut siwdut of MMFACg )

MAPFA(p oc c. 8 5 .

CLINTON - UNIT 1 3/4 2-1

Attachm2nt 2 to U-601239

, N c.Q bSuf 6 Page 27 of 58 1.1 I I

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CUNTON - UNIT 1 ,

3/4 2 2 l

MGaoew reg N --

to U-601239 Page 28 of 58 New Fipr e 1.1 -

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CLINTON - UNIT 1 3/4 2 3 I

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"CGG17 Paga 34 of 58 POWER DISTRIBUTION LIMITS 3/4.2.2 APRM SETPOINTS [Det.ctrb) rg;. p age. in+otheally id + bWs k .

fLIMITINGCONDITIONFOROPERATION' -

3.2.2 The APRM flow-biased simulated thermal power-high scram trip setpoint l (S) and flow-biased neutron flux-upscale control rod block trip setpoint (SRB) shall be established according to the following relationships:

TRIP SETPOINT ALLOWABLE VALUE S 5 (0.66(W-ow) i 48%)T 5 5 (0.66(W-aW) + 51%)T t

S RB 1 (0.66(W-aW) + 42%)T S RB 5 (0.66(W-aW) + 45%)T where: S and S are in percent of RATED THERMAL POWER, RB W = Loop recirculation flow as a percentage of the loop recirculation flow which produces a rated core flow of 84.5 million 1bs/hr. ;

f AW = Difference in indicated drive flow (in percent of drive flow which produces the same core flow) between two loop and single loop operation at the same core flow. See note (a) to Table 2.2.1-1.

T = The ratio of FRACTION OF RATED THERMAL POWER (FRTP) divided by the MAXIMUM FRACTION OF LIMITING POWER DENSITY (MFLPD). T is applied only if less than or equal to 1.0, APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.

ACTION:

With the APRM flow-biased simulated thermal power-high scram trip setpoint and/or the flow biased neutron flux-upscale control rod blof.k trip setpoint less con-servative than the value shown in the Allowable Value column for S or SRB, as determined above, initiate corrective action within 15 minutes and adjust S and/or S

RB to be consistent with the Trip Setpoint value* within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 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 4.2.2 The FRTP and MFLPD shall be determined, the value of T calculated, and the most recent actual APRM flow-biased simulated thermal power-high scram and flow-biased neutron flux-upscale control rod block trip setpoints verified to I be within the above limits or adjusted, as required:

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

With MFLPD greater than the FRTP rather than adjusting the AFRM setpoints, the APRM gain may be adjusted such that the APRM readings are greater than or equal to 100%~ times MFLPD, provided that the adjusted APRM reading does not exceed 100% of RATED THERMAL POWER and a notice of the adjustment is posted on

( the reactor control panel.

CLINTON - UNIT I 3/4 2-5

to U-601239 Pega 35 of 58 o '

POWER DISTRIBUTION LIMITS APRM SETPOINT'S [ DELETE D3 r SURVEILLANCE REQUIREMENTS (Continued) 3 4 .2.2 (Continued) f Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completien of a THERMAL POWER increase of at.

b.

least 15% of RATED THERMAL POVtR, 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 with MFLPD greater than or equal to FRTP.

The provisions of Specification 4.0.4 are not applicable.

b o

This pag. intenhenally idt blanti.

4 CLINTON - UNIT 1 3/4 2-6

acteuteter a

, to U-601239 Page 36 of 58 1

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20 40 60 80 100 o

CORE F LOW (% of Ratedi u _ , - - - - A Reptge, with reol5td Ftgu.3 313-l 0 Hdtd Figure 3.2.3-1 Clinton MCPR f Versus Core Flow CLINTON - UNIT 1 3/4 2-8

= . . . _ . . . ._ .. .

Attachmant 2

. s' to U-601239 Page 37 of 58 1.7 i

\ '

1.6 ', ,

T L

T L 4

\T 'ss j FOR MAX RUNOUT FLOW SETTING = 109.0% --

MCPRf = MAX (1.18,1.841 -0.00707F) j e ~_~

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t 1.3 MCPRf = MAX (1,18,1.746- 0.00659F) i V t i 16 L T i

( \

t 'k i i T N __

k A 1.2 \ \'T __

1.1 1.0 0 20 40 60 80 100 120 CORE FLOW (% rated). F Figure 3.2.3-1 Clinton MCPRf Versus Core Flow CLINTON - UNIT 1 3/4 2 8

Attschmint 2 to U-601239 Pags 38 of 58 1.8 A. A ' -

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CLINTON UNIT 1 3/42-9

Attechmsnt 2 to U-601239 Page 39 of 58

\

1 l

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2.4

-THERMAL POWER 25% s P s 40%

'/ CORE FLOW > 50%

2.2 j

MCPRp =,.2.10 + 0.0033 (40 - P) g'%

\ f

,, THERMAL POWER 25% s P s 40%

CORE FLOW s 50%

2.0 1, < r MCPRp = 1.85 + 0.0133 (40 - P) .__

m /

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x A, __

1.8 c.

2 THERMAL POWER 40% < P s 70%

1.6 [ MCPRp = 1.43 + 0.0045 (70 - P)

%, / ,----

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1 %, s 1.2 ei THERMAL POWER P > 70%

MCPRp = 1,18 + 0.0059 (100 - P) 1.0 100 120 0 20 40 60 80 CORE THERMAL POWER (% rated). P Figure 3.2.3-2 Clinton MCPRp Versus Power for 67 $ 500F and Core Flow 1 107%

CONTON - UNIT 1 3/4 2 9

/.ttschmenc 2 to U-601239 Pega 40 of 58 TABLE 4.3.1.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS TABLE NOTATIONS (a) Neutron detectors may be excluded from CHANNEL CALIBRATION.

(b) The IRM and SRM channels shall be determined to overlap for at least 1/2 de-cade during each startup after entering OPERATIONAL CONDITION 2 and the IRM and APRM channels shall be determined to overlap for at least i decade dar-ing each contro11eo shutdown, if not performed within the previous 7 days.

(c) Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to startup, if not performed within the previous 7 days.

(d) This calibration shall consist of the adjustment of the APRM channel to conform to the power values calculated by a heat balance during OPERATIONAL CONDITION 1 when THERMAL POWER > 25% of RATED THERMAL ~ POWER. Adjust the APRM channel if the absolute difference is greater than 2% of RATED THERMAL POWER, JAny APRM cnannei gain adjustment made in compliance with spectri '

(cation 3<2.2 shall not be included in determining the absolute difference.

(e) This calibration shall consist of a setpoint verification of the Neutron Flux-High and the Flow Biased Simulated Thermal Power-High trip functions.

The Flow Biased Simulated Thermal-High trip function is verified using a calibrated flow signal.

(f) The LPRMs shall be calibrated at least once per 1000 effective full power hours (EFPH) using the TIP system.

(g) Calibrate the analog trip module at least once per 31 days.

(h) Verify measured core (total core flow) flow to be greater than or equal to established core flow at the existing loop tiow control (APRM % flow).

(i) This calibration shall consist of verifying the 610.6 second simulated thermal power time constant.

(j) This function is not required to be OPERABLF when the reactor pressure vessel head is removed per Specification 3.10.1.

(k) With any control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

(1) This function is not required to be OPERABLE when DRYWELL INTEGRITY is not required to be OPERABLE per Special Test Exception 3.10.1.

(m) The CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION shall include the turbine first stage pressure instruments. -

CLINT0h - UNIT 1 3/4 3-10

TABLE 3.3.6-2 b CONTROL ROD BLOCK INSTRUMENTATION SETPOINTS

'l TRIP FUNCTION TRIP SETPOINT ALLOWABLE VALUE

1. ROD PATTERN CONTP.0L SYSTEM

a. Low Power Setpoint (*)% of RATED THERMAL POWER (*)% of RATED THERMAL POWER

,_, b. RWL High Power Setpoint (*)% of RATED THERMAL POWER (*)% of RATED THERMAL POWER

2. APRM

a. Flow Biased Neutron Flux

- Upscale < 0.00 (W-aW) * '2%ata. < e,55 (p au) a 45y** -

Tasact.X Inoperative NA

'b. RA

> 5% of RATED THERMAL POWER > 3% of RATED THERMAL POWER

c. Downscale -

~

)

d. Neutron Flux - Upscale l Stsrtup < 12% of RATED THERMAL POWER < 14% of RATED THERMAL POWER 1

l

$ 3." SOURCE RANGE MONITORS Detector not full in NA NA T a.

< 1 x 105 cps < 1.6 x 105 cps g b. Upscale NA

c. Inoperative RA

d. Downscale 1 3 cps 1 1.8 cps 4 INTERMEDIATE RANGE MONITORS NA

a. Detector not full in NA

b. Upscale < 108/125 division of full scale < 110/125 division of full scale Inoperative RA

c. NA

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

5. SCR*M DISCHARGE VOLUME

a. Water Level-High, C11-N602A < 12" # < 19 7/8" #

b. Water Level-High, C11-N6028 512"## 3197/8"## ,,,

=an oo n

6. REACTOR COOLANT SYSTEM RECIRCULATION FLOW

a. Upscale 4-108bf rated flow < 11-1%Yf rated flow Ii t,*/o -a It 3*/.

7. REACTOR MODE SWITCH EyN NA NA U

a. Shutdown Mode NA

b. Refuel Mode NA

Attcchmelt 2 to U-601239 .

Page 42 of 58 ,,

Insert I to page 3/4 3-66

1) During two recirculation loop operation:

a) Flow Biased 60.66W + 58%** with a maximum 40.66W + 61%** with a maximum of of b) High Flow Clamped 4108.0% of RATED THERMAL 6110.0% of RATED THERMAL POWER POWER

2) During single recirculation loop operation:

a) Flow Biased 40.66(W-AW) + 42%** {0.66(W-Si) + 45%**

b) High Flow Clamped Not required OPERABLE Not required OPEkAB!E e

6 f

Attcchnint 2 to U-601239 TABLE 3.3.6-2 (Continued)

CONTROL ROD BLOCK INSTRUMENTATION SETPOINTS 4

TABLE NOTATIONS To be determined during startup test program. The actual setpoints are the corresponding values of the turbine first stage pressure for these power levels.

- The Average Power Range Monitor rod block function is varied as a function of recirculation loop flow (W). The trip setting of this function must be maintained in accordance with Sp cificetier. 3.2.?, 2nd? note (a) of Table 2.2.1-1. ~

Instrument zero is 758' 5" ms1.

1. Instrument zero is 758' 4^1/2" ms1.

b l

CLINTON - UNIT 1 3/4 3-67 i

, Attschment 2 to U-601239 INSTRUMENTATION s

TRAVERSING IN-CORE PROBE SYSTEM LIMITING CONDITION FOR OPERATION 3.3.7.7 The traversing in-core probe system shall be OPERABLE with:

a. Four movable detectors, drives and readout equipment to map the core and

b. Indexing equipment to allow all four detectors to be calibrated in a common location. ,

APPLICABILITY: When the traversing in-core probe is used for:

a. Recalibration of the LPRM detectors and or 4'

b. Monitoring the APLHGR, LHGR,4MCPR[.{rMFLPDi* l ACTION:

With the traversing in-core probe system inoperable, do not use the system for the above applicable monitoring or calibration functions. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.7.7 The traversing in-core probe system shall be demonstrated OPERABLE by normalizing each of the above required detector outputs within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months prior to use when required for the LPRM or calibration functions.

0nly the detactor(s) in the location (s) of interest are required to be OPERABLE.

CLINTON - UNIT 1 3/4 3-91

Attcchment 2 to U-601239 Pcg2 45 of 58 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM RECIRCULATION LOOPS LtMITING CONDITION FOR OPERATION 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. 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 lavels.

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 W per Specification 2.1.2, and l

t .06 d) Reduce the' Maximum Ave' age Planar Linear Heat Generation Rate (MAPLHGR) lim"-te c ch: of-0.05 ti es the twc--recirculat4e#

-leepcparaticNimitperSpecification3.2.1,and 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, W and 3.3.6, and i

See Special Test Exception 3.10.4.

tThe operating region for which monitoring is required. See Surveillance Requirement 4.4.1.1.2. ,

ftDatector levels A and C of one LPRM s'tring per core octant plus detectors

~ A and C of one LPRM string in the center of the core should be monitored.

t i CLINTON - UNIT 1 3/4 4-1

Attechn:nt 2 to U-601239

, 3/4.2 POWER DISTRIBUTION LIMITS Pega 46 of 58 8ASES The specifications of this section assure that the peak cladding temperature followingthepostulateddesignbasisloss-of-coolantaccidentwillnotexceed 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 GENERATION RATE (APLHGR) is this LHGR of the highest powered rod divided by its local peaking factor. The limiting value for APLHGR is : hewn in Figurce 1 ? 1-1, 3. 2.1-2 &nd 3. 2.1-3.' 96c f4AFL H62.

Invr't Y S.I.1-5 4ro h 341 1 The calculational procedure used to establish the APLHGR shown on Figures 33r2t1-41--

= E3.2.1-2 cad 3.2.1-0 is based on a loss-of-coolant accident analysis. The analy-sis was performed using Ceneral Electric (GE) calculational models which ar.e consistent with the requirements of Appendix K to 10 CFR 50. A complete discus-sion of each code employed in the analysis is presented in Reference 1. Differ-ences in this analysis compared to previous analyses can 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. Incorporated more accurate bypass areas - The bypass areas in the top guide were recalculated using a more accurate technique.

3. Corrected guide tube thermal resistance.

4. Correct heat capacity of reactor internals heat nodes.

b

~

CLINTON - UNIT 1 8 3/4 2-1

Attachm2nt 2 to U-601239 Page 47 of 58 Insert Y to page B 3/4 2-1 The MAPLHGR limits of Figures 3.2.1-3 through 3.2.1-7 are multiplied by the smaller of the flow-dependent MAPLHGR factor (MAPFAC f

) or the power-dependent MAPLHGR factor (MAPFAC ) corresponding to existing core flow and power conditions to assure tee adherence to fuel mechanical design bases during the most limiting transient (Reference 2). The MAPFAC factorsaredeterminedusingthethree-dimensionalBWRsimulatorcoketo 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 factors are generated using the same data base as the MCPR to protIct the core from plant transients other than core flow runout P h

=

e

y _ _ _ _ _ _ _ _

. Attcchment 2 to U-601239

POWER DISTRIBUTION LIMITS Paga 48 of 58 BASES 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (Continued)

b. Model Change

1. Core CCFL pressure differential - 1 psi - Incorporate the assumption that flow from the bypass ter 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 below.

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. the smallest et MAPFACp, 3.2.1-3 ihrough 3.2.l-7 MAPFAC, or For t operation plf0 2.1-1,witK a singleand 3.2.1-2 re girculation 3.2._1-3loop,rare the MAPLHGR limitsby multiplied of 0.84 Figure The constant factor, 0.85, is derived from LOCA analyses initiated from Tingle loop operation to account for earlier boiling transition at the limiting fuel node compared to standard LOCA evaluations.

(Reference 2) 3/4.2.2 APRM SETPOINTS [ del.ETED]

hhe fuel cladding integrity Safety Limits of Specific '!on 2.1 were based on a power distribution which would yield the design LHGR .s "\TED THERMAL POWER.

The flow biased simulated thermal power-high scram setting and the flow biased neutron flux-upscale control rod block functions of the APRM instruments must be adjusted to ensure that the MCPR does not become less than the fuel cladding safety limit or that > 1% plastic strain does not occur in the degraded situa-tion. The scram and rod block setpoints are adjusted in accordance with the formula in this specification when the combination of THERMAL POWER and MFLPD indicates a peak power distribution to ensure than a LHGR transieM. would not L be increased in degraded conditions.

% _- a CLINTON - UNIT 1 B 3/4 2-2

, . Attcchment 2 to U-601239 C 6 Paga 49 of 58 BASES TABLE B 3.2.1-1 SIGNIFICANT INPUT PARAMETERS TO THE LOSS-OF-COOLANT ACCIDENT ANALYSIS

Plant Parameters:

Core THERMAL POWER .................... 3015 MWt** which corresponds 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 fts,

b. Small Breaks 0.09 ft 2, .

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 and Rdead )

Initta1 4Cores 8x8 13.4 1.4 1.17*** l

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 requirement 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 operation, loss of nucleate boiling is ,

assumed at 0.1 seconds after a LOCA p.< ore. St.ws less h e5'/. of red ,regardless At innial of McPt.initialisMCPR.

k h <n Ec.e

% tac P Rg beva.. . \

l I

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

, Att chment 2 to U-601239 POWER DISTRIBUTION LIMITS Pcga 50 of 58 BASES 3/4.2.3 MINIMUM CRITICAL POWER RATIO

n Specifee.Jaen 2.1.2.

The required operating limit MCP at steady state operating conditions as specified in Specification . 3 are derived from the established fuel cladding integrity Safety Limit MCPR .f 1.007 and an analysis of abnormal operational l transients. For any abnormal operating transient analysis evaluation with the initial condition 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. f M or ge. rechkHen lo.P epa

hen t u pu" A To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting transients have been analyzed to determine which result in the largest reduction in CRITICAL POWER RATIO (CPR). . The type of tre.nsients evaluated were loss of flow, increase in pressureThe decrease. and power, positive power-flow map reactivity insertion, an[doperational j of FiguresB 3/4.2.3-1 give coolant temperature limits.

AMehn Ret <. 3 %ferem 3 or 8 %.t.3 21 and neiprowe NIM The evaluation of a gi'on transient begins with the system initial parameters E ,. in f$AR Teb1.5 1p.02er.d15.6.5-ithatareinputtoaGE-coredynamic behavior trinsient conputer program The code $used to evaluate pressurization eventsNdescribedi[NEDO- 54 3) an the program used in non pressurizatioA Eeventsis'describedinNED0-10802(2) The outputs of this program along with Q the initial MCPR form the input for further analyses of the thermally limiting j bundle with the sing!e channel transient thermal hydraulic TASC code describedi (in NEDE-25149(4) J The principal result of this evaluation is the reduction in MCPR~ caused by the transient.

The purpose of the MCPR and MCPR of Figures 3.2'.3-1 and 3.2.3-2 is to define f p operating limits at other than rated core flow and power conditions. At less than 100% of ratad flow and power the required MCPR is the larger value of the MCPR and MCPR at the existing core flow and power state. The MCPRf s are f p established to protect the core from inadvertent core flow increases such that the 99.9% MCPR limit requirement can be assured.

The MCPR f s were calculated such that for the maximum core flow rate and the mott upwu , pow-corresponding THC.RMAL POWER along the 10% m .nw .te.J flow control line, the limiting bundle's relative power was adjusted until the MCPR was slightly above the Safety Limit,o Using this relative bundle power, the MCPRs were cal-culated at different poi nts along the 10 Z ef reted A em -flow control line l corresponding to differe ntcoreflows.fThecalculatedMCPRatag_ivenpoint of core flow is defined as MCPRf . \mp. lim;Hng par-Insed. k CLINTON - UNIT 1 B 3/4 2 4

Attschment 2 to U-601239 Page 51 of 58 Insert "A" to page B 3/4 2-4 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.

0 4

e

Attacha nt 2 to U-601239 POWER DISTRIBUTION LIMITS BASES 3/4.2.3 MINIMUM CRITICAL POWER RATIO (Continued)

The MCPR s are established to protect the core from plant transients other than p

core flow increases, including the localized event such as rod withdrawal error.

The MCPR s were calculated based upon the most limiting transient at the given p

core power levelg inclM 3 mar owuv W 61 rejeAn b o mb.. :Insu t Z.

At THERMAL POWER levels less than or equal to 25% of RATED THERMAL POWER, the reactor will be operating at minimum recirculation pump speed and the moderator void content will be very small. For all designated control rod patterns which may be employed at this point, operating plant experience indicates that the resulting MCPR value is in excess of requirements by a considerable margin.

During initial start-up testing of the plant, a MCPR evaluation will be made at 25% of RATED THERMAL POWER level with minimum recirculation pump speed. The MCPR margin will thus be demonstrated such that future MCPR evaluation below this power level will be shown to be unnecessary. The daily requirement for calculating MCPR when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes. The requirement for calculating MCPR within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after the completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER ensures thermal limits are met after power distribution shifts while still allotting time for the power dis-tribution to stabilize. The requirement for calculating MCPR after initially determining a LIMITING CONTROL ROD PATTERN exists ensures that MCPR will be known following a change in THERMAL POWER or power shape that could place operation exceeding a thermal limit.

3/4.2.4 LINEAR HEAT GENERATION RATE This specification assures that the Linear Heat Generation Rate (LHGR) in any rod is less than the design linear heat generation even if fuel pellet densifi-cation is postulated.

The daily requirement for calculating LHGR when THERMAL POWER is greater than or equal te 25% of RATED THERMAL POWER is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes. The requirement to calculate LHGR within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after the completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER ensures ther-mal limits are met after power distribution shift. Calculating LHGR af ter initially determining a LIMITING CONTROL P.0D PATTERN exists ensures that LHGR will be known following a change in THERMAL POWER or power shape that could place operation exceeding a thermal lii.:it. ,

REFERENCES:

1. General Electric Company Analytical Model for loss-of-Coolant Analysis in Accordance with 10 CFR 50, Appendix K, NEDE-20566, November 1975.

i CLINTON - UNIT 1 B 3/4 2-5

4 Attachm2nt 2 to U-601239 Page 53 of 58 Insert Z co page B 3/4 2-5 For core power below 40% of RATED THERMAL POWER where the EOC-RPT and reactor scram on turbine stop valve closure and turbine control valve fast closure sre bypassed, separate sets of MCPR limito are provided for high and low core flows to account for the sInsitivity to initial core flows. For core power above 40% of RATED THERMAL POWER, bounding MCPR limits were developed.

b 0

l

~

, Attachment 2 to U-601239 POWER DISTRIBUTION LIMITS cas M of M BASES

' REFERENCES (Continued):

f2. R. B. Linford, Analytical Methods of Plant Transient Evaluations for the A.

GE BWR, February 1973 (NEDO-10802).

3. Qualification of the One Dimensional Core Transient Model for Boiling Water Reactors, NEDO-24154, October 1978.

f

4. TASC 01-A Computer Program for The Transient Analysis of a Single Channel, Technical Description, NEDE-25149, January 1980.

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150 A. NATURAL CIRCULATION

8. LOW RECIRCULATION PUMP SPEED VALVE MINIMUM POSITION IN ~

C. LOW RECIRCULATION PUMP SPEED VALVE MAXIMUM POSITION D. RATED RECIRCULATION PUMP SPEED VALVE MINIMUM POSITION 130 -

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. Attachm2nt 2 to U-601<39 Page 57 of 58 00 -

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Attachment 2

to U-601239 e

Pag 3 $8 ci $8.

CONTAINMENT SYSTEMS BASES . _

3/4.6.2 4 DRWELL STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the d ywell will be maintained comparable to the original design specification for the life of the unit. A visual inspection in conjunction with Type A leakage tests is suffi-cient to demonstrate this capability.

3/4.6.2.5 DRWELL INTERNAL PRESSURE IM The limitations on drywell-to-containmen differential pressure ensure that the drywell peak calculated pressure of MWpsig does not exceed the design pressure of 30.0 psig and that the containment peak pressure of 9.0 psig does not exceed the design pressure of 15.0 psig during steam line break conditions. The maxi-mum external drywell pressure differential is limited to 0.2 psid, well below the pressure at which suppression pool water will be forced over the wier wall and into the drywell. The limit of 1.0 psid for initial positive drywell to i containment pressure will limit the drywell pressure to I g psid which'is less I than the design pressure and is consistent with the safety (analysis to limit drywell internal pressure. \ p.]

3/4.6.2.6 DRWELL AVERAGE AIR TEMPERATURE The limitation on drywell average air temperature ensures that peak drywell temperature does not exceed the design temperature of 330*F during LOCA condi-tions and is consistent with the safety analysis.

3/4.6.2.7 DRWELL VENT AND PURGE The drywell purge system must be normally maintained closed to eliminate a potential challenge to containment structural integrity due to a steam bypass of the suppression pool Intermittent venting of the drywell is allowed for pressure control during OPERATIONAL CONDITIONS 1, 2, and 3, but the cumulative time of venting is limited to 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months per 365 days. Venting of the drywell is prohibited when the 12-inch continuous containment purge system or the 36-inch containment building ventilation system supply or exhaust valves are open. This eliminates any resultant direct leakage path from the drywell to the environment.

In OPERATIONAL CONDITIONS 1, 2 and 3, the drywell isolation valves (IVQOO2, IVQ003) can be opened only if they are blocked so as not to open more than 50'.

This assures that the valve would be able to close against drywell pressure buildup resulting from a LOCA. .

Operation of the drywell vent and purge 24-inch supply and exhaust valves during plant operational conditions 4 and 5 is unrestricted; the 50' blocks may be removed to allow full opening of the valves, and the cumulative time for vent and purge operation is unlimited.

CLIHTON - UNIT 1 8 3/4 6-5

Attcchm:nt 3 to U-601239 Page 1 of 2 Technical Specification Changes Required to Support Refueling The reload analysis provided in Attachment 6 "SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR CLINTON POWER STATION UNIT 1. RELOAD 1, CYCLE 2, 23A5921, Rev. 0" provit es a licensing basis and the essential analyses for allowing CPS to perform its first reactor refueling (in which new types of nuclear fuel will be utilized) and to proceed with sub_ sequent reactor operation with the reloaded core.

Two new G.E. BWR fuel types will be utilized for the CPS reload. These fuel types have specified MAPLHGR-vs-Core Exposure requirements.

Additionally, as noted in th9 reload analysis, General Electric Standard Application for Reactor Fuel (GES. TAR), NEDE-24011-P-A-8 which established the MCPR Safety Limit of 1.06 (1.07 for single recirculation loop operation) for the initial fuel cycle for CPS, requires this safety limit to be increased by 0.01 for reload cores. As the reload analysis includes an evaluation of plant operation (including postulated responses to design basis accidents or transients), the Technical Specifications must be changed to reflect the revised safety or power distribution limits.

Each of the proposed Technical Specification changes related to t'he reload analysis is briefly described 'in Attachment 2. Detailed justification for each of these changes is provided below. Some of the changes identified as reload-related are not unique to the firat reload 3

(for Cycle 2) as they apply to reloads in general. They should, however, minimize the number of Technical Specification changes that would otherwise be required for future roloads.

Technical Specification 1.91 This change replaced the reference to the GEXL correlation with reference to "an approved GE critical power correlation". This will allow future reloads to be analyzed vaing other approved critical power correlations, without requiring a change to the definition of CRITICAL POWER RATIO. For cycle 2 the approved GE critical power correlation is the GEXL correlation.

Technical Specification 2.1.2: The change to the MCPR Safety Limit from 1.06 to 1.07 for two recirculation loop operation (and from 1.07 r,o 1.08 for single recirculation loop operation) is a typical change made at the first refueling outage. The increase is due to increased uncertainties in power distribution, and local fuel rod power.

Technical Specification Bases 2.1.0: This change replaces the specific value for MCPR with a referenc e ') Technical Specification

' This change is being 2.1.2 where the value for MCPR is spet :s d .

proposed to allow the vaJue of MCPR t . : 1 changed in future reload applications without requiring a chane so this section.

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Att:chment 3

to U-601239 Page 2 of 2 Technical Specification Bases 2.1.1: This change replaces reference to the GEXL correlation with roference to "an approved GE critical power correlation". This will allow future reloads to be analyzed using other approved critical power correlations, without requiring a change to this Technical Specification section, Technical Specification Bases 2.1.2: This change revises the description of the determination of the MCPR safety limit by referencing the standard GE reload licensing document (GESTAR).

GESTAR contains all of the assumptions, analysis methods, and other information used to perform this analysis for GE fuel at CPS.

Technical Specification Bases Tables B2.1.2-1 and B2.1.2-2: This change deletes the cycle specific reload information f rom the Technical Specification Bases. This information is contained in ,

the GESTAR document. This change allows changes to the nominal values and uncertainties used to perform MCPR calculations to be updated without requiring a change to the Technical Specificatf ons.

Technical Specification 3/4.1.3.2: The footnote which grant's relief from scram time testing during the initial fuel cycle is being deleted because the initial fuel cycle will be completed at the beginning of the first refueling outage, therefore, the note is no longer applicable.

Technical Specification Bases 3/4.2.3 (paragraph 1): This change replaces the specific value for MCPR with a reference to Technical Specification 2.1.2 where the value for MCPR is specified. This change is being proposed to allow the value of MCPR to be changed in the future without requiring this section to be changed.

Technical Specification Bases 3/4.2.3 (paragraph 3): This change replaces references to specific analysis methods used to analyze pressurization and non-pressurization transients with a reference to the standard GE reload licensing document (GESTAR), because GESTAR contains these specific analysis methods and other information used to perform analyses for GE fuel at CPS.

i Technical Specification Bases section 3/4.2 (References): lhis change revises the reference list to remove documents that have been removed from Technical Specification Bases section 3/4.2. The documents listed are contained within GESTAR which is now referenced in this section.

l

Attcchm:nt 4

to U-601239 Pcgo 1 of 9 Technical Specification Changes Required to Support Operation Consistent with the "MAXI}UM EXTENDED OPERATING DOMAIN AND FEEDWATER HEATER OUT-OF-SERVICE ANALYSIS FOR CLINTON POWER STATION" (HEOD/FWHOS)

Purpose of MEOD/FWHOS Analysis The "MAXI}RTM EXTENDED OPERATING DOMAIN AND FEEDWATER HEATER OUT-0F-SERV (CE . i_ ' S FOR CLINTON POWER STATION" provides the licensing basis and essential analyses for permitting reactor operation in an expanded domain. The current power flow operating domain is depicted in Technical Specification Bases Section 3/4.2.3 (Bases Figure B3/4.2.3-1, "Reactor Operating Map"). This operating map was developed '

based on restrictions such as recirculation pump NPSH, plant control characteristics and core thermal power and flow limits. Safe operation in this region is justified by the accident and transient analyses described in Final Safety Analysis Report (FSAR) Chapters 6 and 15. In order to improve the operating flexibility and the capacity factors for CPS, IP contracted General Electric to evaluate the accident and transient scenarios for the modified operating map in the regions of the Maximum Extended Oparating Domain (MEOD).

The MEOD consists of two regions which supplement the current power flow map. One region expands the map to permit flows up to 107% of rated core flow; this region is termed the increased core flow region (ICFR).

The second region, known as the extended load-line region (ELLR),

permits operation at rated power levels with core flows less than 100%.

By expanding the operating domain allowed on the power flow map, significant benefits can result leading to greater operational flexibility and to improved unit capacity factor. From a core operations and fuel management standpoint the chief benefits are: 1) better power shaping and fuel preconditioning, 2) xenon compensation, and 3) compensation for reactivity reduction due to exposure. -

4 The ability to increase power into the extended load-line region at low core flows allows the withdrawal of more control rod notches. As a result, the plant can attain, or closely approach, the full power target rod pattern. This enhances the capability to obtain optimum axial power shapes prior to encountering fuel preconditioning limitations. The net effect of this capability is an improvement in capacity f actor brought about by optimized preconditioning ramps and the elimination of subsequent power reductions to attain the target control rod pittern.

The MEOD additionally provides a fuel performance improvement through the reduction in thermal duty cycling on the fuel-cladding interface.

If the rated load line control rod pattern is maintained as core flow is increased, changing equilibrium xenon' concentrations will result in less than rated power at rated core flow. Additional operating margin above the rated rod line on the power flow map (as permitted under the MEOD) allows for compensation for power reductions during plant startups due to transient xenon. The gross power reduction due to the reestablishment of equilibrium xenon conditions at rated power have been observed to be as great as 10%-12% during startups with peak xenon and 8%-10% during xenon-free startups. Excess flow capability will ensure.

subsequent to attainment of equilibrium xenon, that the plant is capable of maintaining rated power.

u Attachm:nt 4 to U-601239 Page 2 of 9 In order to maintain a high capacity factor, continued operation at rated conditions is necessary. The effects of xenon ouildup and fuel burnup reduce core thermal power and decrease the plant capacity factor.

A significant benefit that MEOD offers during rated power operation lies in the fact that rated power conditions can be maintained for a longer period of time withcut maneuvering rods. This is made possible because rated power can be achieved at less than rated core flow. In the extended load-line region, 100% power can be achieved at 75% flow.

Reactivity changes due to fuel burnup, burnable poison depletion, and increased xenon inventory can be countered with variations in core flow.

Increased core flow above 100% is an additional aid provided by the increased core flow region of the MEOD. The ability to stay at full power can be extended by increases in core flow above rated core flow.

Included with the MEOD analysis is an evaluation which justifies eliminating the APRM flow-biased simulated thermal power-high scram se'epoint adjustment (currently required under Technical Specification 3.2.2). More mea:1,s tful power and flow-dependent MAPLHGR limits, together with the naw MCPR limits, supersede the need to manually adjust chis setpoint. Additional details are provided in the next section of this attachment.

Also provided with the MEOD analysis in Attachment 7, is an analy' sis performed to evaluate the impact of reduced feedwater temperature on reactor operation at rated (worst case) conditions. It provides justification for operating the reactor with a feedwater temperature from 420*F (rated) down to 370'F at rated conditions with no required changes to the limits or setpoints established or assumed by the MEOD analysis (including the consideration given to eliminating the APRM flow biased simulated thermal power-high scram setpoint adjustment).

O m

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i Attachm:nt 4 l to U-601239 Page 3 of 9 1

Evaluation of MEOD/FWHOS Analysis As discussed previously, the MEOD analysis provides a basis for permitting reactor operation in an expanded domain. Operation within the new domain was therefore evaluated for impact on the accident and transient analyses. Although the details of this evaluation are a part of the analysis provided in Attachment 7, the key results obtained are summarized below.

1) LOCA analysis - A bounding BWR-6 analysis determined that the current MAPLHGR and MCPR limits (FSAR Chapter 6) are adequate f

for the MEOD.

2) Containment response - A conservative containment analysis produced a peak dryvell pressure of 19.7 psig. This pressure is greater then the drywell pressure determined in the FSAR Chapter 6 analysis but otill well below the design pressure of 30 psig.

3) Abnormal Transients - A bounding BWR-6 analysis concluded that the delta-CPR results for all cases analyzed in the ME0D are enveloped by the current MCPR limits. The MCPR f curve is revisedbasedonthenewanalfsisoftheslowrecircula' tion flow runour transient event to accommodate operation in the ICFR. In addition, the following limiting transients were analyzed in detail for CPS:

a) Generator Load Rejection with Bypses Failure - As discussed in FSAR Section 15.2.2 this limiting vessel pressurization transient produced a peak vessel pressure of 1225 psig in the FSAR results. tihen evaluated for the MEOD, the peak pressure increased only slightly to 1226 psig.

b) Feedwater Flow Controller Failure - Based on the MEOD cvaluation, the existing MCPR operating limits are adequate to ensure this transient will not violate the MCPR safety limit. (MCPR has been revised, however, based on elimination of tee APRM flow-biased simulated thermal power-high scram setpoint adjustment as discussed below.)

c) As described in the attached report, the results of the FSAR Chapter 15 evaluations of both the 100'F Loss of Feedwater Hester Transient and the Rod Withdrawal Error Transient were found to provide adequate protection in the FEOD.

d) llow Runout Transient - The evaluation of this transient in the ME0D stablished the flow-dependent MCPR limits.

This event, analyzed at two flow limiter settings, resulted in MCPR values which were found to bound other flowdependentafnormaltransientevents, i

Attcchment 4 to U-601239 Page 4 of 9

4) Stability Evaluation - CPS Technical Specifications liave implemented the recommendations of GE SIL-380. The stability compliance of all licensed GE BWR fuel designs, including those fuels described in General Electric Standard Application - ;

for Reactor Fuel (CESTAR) is demonstrated in NEDE-22277-P-1.

The CPS cycle 2 reload contains GE BWR core fuel and therefore complies with 10CFR50, Appendix A. GDC 12.

In addition, evaluation demonstrates that substantial thermal-mechanical margin is available for the GE BWR fuel i designs even in the unlikely event of very large power oscillations.

5) The effects of increased Reactor Internal Pressure !

Differences, acoustic loads, flow induced loads, and fuel !

bundle lift forces have been evaluated and shown to not cause [

design limits to be axceeded.

Based on actual flow-induced vibration testing at a valid I prototype plant (Kusheng 1) to 105% of rated flow, and

! extrepolations of this data to 107% of rated flow, it is predicted that the maximum alternating stresses on vessel internals will be approximately 70% of tha acceptance criteria (10,000 psi).

6) Overpressure Protection - The MSIV c16sure transient was analyzed in the MEOD. The peak vessel pressure is 1245 psig, 4

which is well within the design limit of 1375 psig.

The MEOD analysis also addresses the elimination of the APRM flow-biased

- simulated thermal power-high scram setpoint adjustmenc. The transient ,.

analyses performed for MEOD defines the appropriate thermal / operating !

limits which will assure that the criteria associated with thermal-mechanical fuel integrity and LOCA considerations are satisfied I without requiring a setpoint adjustment.

With respect to the Feedwater-Heater-Out-of-Service (or Reduced ,

Feedwater Temperature) analysis, which was included with the MEOD !

analysis in Attachment 7, an evsluation was performed which considered i

the followingt {

l 1

1) FSAR Chapter 15 abnormal operating transients

2) fuel mechanical design limits f t

3) LOCA and c:ontainment response as described in FSAR Chapter 6 l

4) fuel integrity thermal-hydraulic stability, and

5) effects of acoustic and flow induced loads. t t

__l

Attachment 4

. to U-601239 Page 5 of 9 The results of the evaluation indicate that operation with reduced feedwater temperatures (as low as 370'F at rated power) is acceptable because such operation is supported by the transient analyses performed for MEOD in which it was demonstrated that the acceptance criteria related to fuel integrity and LOCA considerations continue to be satisfied.

e 9

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, i

Attcchacnt 4 to U-601239 Page 6 of 9 Justification for Individual Technical Specification ch'anges Technical Specification 1.15: The definition of FRACTION OF LIMITING POWER DENSITY (FLPD) is being deleted because the only section of the CPS Technical Specifications that uses this parameter is Technical Specification 3/4.2.2 which is being deleted as part of this submittal.

Technical Specification 1.16: The definition of FRACTION OF RATED THERMAL POWER (FRTP) is being deleted because the only section of the CPS Technical Specifications that uses this parameter is Technical Specification 3/4.2.2 which is being deleted as part of this submittal.

Technical Specification 1.23: The definition of MAXIMUM FRACTION OF LIMITING POWER DENSITY (MFLPD) is being deleted because the only section of the CPS Technical Specifications that uses this parameter is Technical Specification 3/4.2.2 which is being deleted as part of this submittal. .

Technical Specification Table 2.2.1-1: This change increases the APRM flow biased scram setpoint and allowable value by 16%. This

- change is made to accommodate operation in the MEOD. Operation in the MEOD vas analyzed as described in the attached MEOD analysis.

It was found that operationi in this region would not exceed design limits. This proposed change provides access to the extended load line region of the MEOD. The revised setpoints maintain the same slope, the same clamped setpoint and the same margin between the scram and rod block setpoints as the current technical specifications.

Technical Specification Bases 2.2.1: This change to the APRM scram function is an administrative change to delete the reference to Specification 3.2.2 which has been proposed for deletion in this change request. Justification for this caletion is provided below.

Technical Specification 3/4.2.1: As discussed in Section 2.6.3 of the atta'ched MEOD analysis, the MAPLHCR reduction factors (MAPFAC g and MAPFAC ) are derived from bounding BWR/6 analysis and the Clinton spFcific analysis where needed. Using these MAPLHGR reduction factors to reduce the rated MAPLHGR limits will ensure that the fuel thermal-mechanical limits will not be exceeded when the APRM flow-biased simulated thermal power-high scram setpoint adjustment (Technical Specification 3/4.2.2) is no longer required. ,

Attcchment 4 to U-601239 Pcge 7 of 9 Technical Specification 3/4.2.2: The current CPS Technical Specifications require that the flow biased scram and rod block setpoints be lowered when the ratio of the Fraction of Rated Power to the Maximum Fraction of Limiting Power Density is less than 1.0.

This requirement originated from a now obsolete Minimum Critical Heat Flux Ratio (MCHFR) criterion. The change to the General Electric BWR Thermal Analysis Basis (GETAB), NEDO-10958-A, as a licensing basis and a secondary reliance of flux scram for transient evaluations (for those transients terminated by a scram) ncy provides a more effective alternative to this requirement.

With a revision in the power dependent MCPR limit and new flow and power dependent MAPLHCR reduction factors, it has been demonstrated that operation remains within design and regulatory' limits.

Technical Specification Figures 3.2.3-1 and 3.2.3-2: As a result of the MEOD analyseg of the slow recirculation flow run out transient a new flow dependent MCPR (MCPR f ) limit was established.

The proposed curve is slightly greater than the existing curve but is not expected to unduly regrict normal operation.

A new set of power dependent mci'R (MCPR ) limits has been developed based on the evaluation of the elimination of the APRM flow biased simulated thermal power-high scram setpoint adjustment. The new limits are derived from the results of both CPS-specific and bounding BWR/6 analyses. These limits have been generated considering reduced feedwater temperature, and are therefore applicable to operation with reduced feedwater temperatures.

The operating limit MCPR at any power / flow condition is the larger of the new MCPRg and the MCPR . The new values are presented in the revised Figures 3.2.3-1 aEd 3.2.3-2.

p ehnical Specification Bases 3/4.2.1: A paragraph is being added to this section to discuss how MAPFAC and MAPFAC are to be used, and how MAPFAC, and MAPFAC are determined. The Iource of this information is~section 2.6 3Pof the MEOD analysis.

The MAPLHGR figures in Technical Specification section 3.2.3 are referred to in two paragraphs of this section. These references are being revised to correctly reference the renumbered MAPLHGR figures.

The last paragraph in this section discusses MAPLHGR requirements while in single recirculation loop operation (SLO). This paragraph is being revised te require consideration of the MAPLHGR reduction factors (MAPFAC multiplier (0.8b)and MAPFAC ), as well as the SLO MAPLH';Rwhen d The MAPLHGR reduction factors curt be considered because the justificatien for de.letion of Technical Specification 3, 2.2 was based in part on the conservatisms gained by use of the.e MAPLHCR reduction factors.

Attcchment 4 to U-601239 '

Page 8 of 9 Technical Specification Bases 3/4.2.2 This section is 'oeing deleted because Technical Specification 3/4.2.2 was deleted as discussed previously.

Technical Specification Bases Table B3.2.1-1: This change incorporates additional information into this section of the l Technical Specification Bases regarding assumptions used to determine MCPR for core flows less than 85%. This information is from the MEOD analysis. l i

Technical Specification Baus 3/4.2.3 (paragraph 5): This change i defines the new control 7od line to be used when determining values ,

for MCPR,. The change is required because the MEOD region allcws ,

contrcl tod lines (and core flow rates) in excess of the current limits.

i t

Technical Specification Bases 3/4.2.3: This changc' adds a - !

discussion of how the MCPR limits are determined whe core power :

is less than 40% of RATED fHERMAL POWER. The new MCPR limits are '

flow dependent at core power below 40% of RATED THERMAE POWER.

Below 40% of rated power, the end of cycle-recirculation pump trip and the turbine stop valve closure and turbine control valve fast closure scrams are bypassed. Because of the bypass there is a significant MCPR sensitivity to initial core flows. At high core l flows (i.e., greater than 50% of rated) the MCPR is increased in I order to maintain the margin of safety. P I Technical Specification Bases Figure B 3/4.2.3-1: This change revises the CPS operating map. The map is being revised to show the boundary of operation allowed by the MEOD. f L

, Technical Specification Bases Figure B 3/4.2/3-2 (new): This '

c Onge adds a new operating map for single recirculation loop operaMon, to clarify what operating regions are acceptable for single recirculation loop operation. ;

i Technical Specification Table 4.3.1.1-1 (note d): This change j removes reference to Technical Specification 3.2.2 which is being t i

deleted, t l

l

Attechnent 4 ,

to U-601239 Page 9 of 9 Technical Specification Table 3.3.6-2 (APRM flow biased rod blockt) The increase in the APRM flow-biased rod block setpoint is similar to that for the flow-biased simulated thermal

power scram justified above. The high flow clamp is added to maintain the same clamp setpoint at rated power / flow conditions as is currently available. That is, the maximum setpoint currently available is 108% (at W=100%). In the MEOD, the new setpoint (which could be calculated to be as high as 128%) is still to be clamped at this value (108%).

Technical Specification Table 3.3.6-2 (Reactor Coolant System Recirculation Flow): The Reactor Coolant System Recirculation Flow

- High rod block setpoint is increased from 108% to 113%. Operacian in the increased core flow region of the operating map (i.e., with core r flow up to 107% of rated) has been evaluated as discussed in the l attached MEOD analysis. Raising the rod block setpoint will minimize i unnecessary rod block alarms when operating in the increased core flow region. While the allowable flow range has been extended from 100% to 107%, the rod block setpoint has been conservatively raised by only 5%.

Technical Specification Table 3.3.6-2 (** note): This change removes ref erence to Technical Specification 3.2.2 which is being .

deleted.

Technical Specification 3.3.7.7 This change removes the requirement to monitor MFLPD. The only reason for monitoring MFLPD is to ensure that Technical Specification 3/4.2.2 can be met.

Since Technical Specification 3/4.2.2 is being deleted as part of this submittal MFLPD no longer needs to be monitored.

l Technical Specification 3.4.1.1 ACTION a.1.cl This change revises i i

the value for the MCPR Safety Limit to 1.08 as discussed in the justification for the similar change to Technical Specification 2.1.2.

t Technical Specification 3.4.1.1 ACTION a.1.dt This change removes the j j requirement to multiply the MAPLHGR limits by 0.83, and replaces it with a .

j reference to Technical Specification 3.2.1 which contains a requirement to multiply the MAPLHGR limit by the smallest of MAPFAC , MAPFAC or 0.85.

1 I i

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Technical Specification 3.4.1.1 ACTION a.1.vt This change removes reference to Technical Specification 3.2.2 which is being deleted [,

as discussed previously.

l Technical Specification Bases 3/4.6.2.5: The dryvell peak !

l calculated pressure is being changed from 18.9 psig to 19.7 psig.

l For a reactor recirculation piping bresk at the most limiting condition j in the MEOD and with a reduced feedvater temperature, the predicted peak >

drywell pressure increases slightly. This increase is predoninately due to i I

the increased mass flow rate out of the break. The increasedThe mass flow new peak rate results from increased density of the reactor coolant. l predicted value is well within the design limit of 30 psig.

1

< Attcchment 5 to U-601239 Page 1 of 3 Basis For No Significant Hazards Consideration (for the Technical Specification Changes Proposed as a Result of Operation with New' Fuel Types and in the MEOD Region)

According to 10CFR50.92, a proposee change to the license (Technical Specifications) involves no significant hazards consideration if operation of the facility in accordance with the proposed change would not (1) involve a significant increase in the probability or consequences of any accident previously evaluated, (2) create the possibility of a new or different kind of accident from any accident previously evaluated, or (3) involve a significant reduction in the margin of safety. ,

The changes proposed in this submittal will allow continued operation during cycle 2 at CPS, provide for operation in the Maximum Extended Operating Domain (MEOD) and allow for elimination of the APRM flow-biased simulated thermal power-high scram setpoint adjustment requirements. All of these propoted changes have been evaluated and found to be appropriate for operation at RATED THERMAL POWER with feedwater temperature as low as 370*F (and equivalent operation at lower thermal power).

These changes have been evaluated as discussed in the attached analyses.

Based on these analyses, the proposed changes do not involve a ,

significant hazards considerr. tion. Details ot' the basis for this conclusion are provided below by addressing the three concerns outlined ,

in 10CFR50.92.

1) The proposed changes do not involve a significant increase in the ,

probability or consequences of any accident previously evaluated. !

The changes to support the MEOD satisfy this concern because plant equipment and systems will continue to operate within their design limits. Changes to support operation in the MEOD involve a revision of the MCPR g limit and higher limits for the APRM scram j

and rod block setpoints. The increased Operating Limit for MCPR is needed to maintain the same margin of safety in the increased core flow region that was established for the current operating domain with respect to the recirculation flow run out transient. The revised limit ensures the consequences of this event are not increased. The revised APRM setpoints maintain the same scram / rod ,

block-to-power margin in the MEOD as is currently provided. These conclusions are based on an evaluation (see attached MEOD analysis) which considered the following:

- A bounding BVR/6 LOCA analysis was performed for the MEOD. It was determined that current MAPLHGR and MCPR limits and the revised MCPR limits are adequate to ensure EOCA consequences f

are not increased.

- The containment response for a design basis accident in the MEOD, considering a,feedvater temperature reduction due to feedwater heater (s) out of service, is slightly more severe than the analysis provided in Final Safety Analysis Report '

(FSAR) Section 6.2. As presented in the FE0D analysis, the f

Attcchment 5 to U-601239 Page 2 of 3 differential peak drywell pressure of 19.7 psig is 0.8 psi above the current CPS FSAR Chapter 6 value, but it is still well below the design limit of 30 psig.

Fuel thermal and mechanical performance for transients initiated in the KEOD is bounded by the fuel design bases.

- The effects of acoustic, flow induced, and reactor internal pressure differential induced load and of increased flow on the fuel bundle and reactor internals '

were found to be well within allowable design limits.

The elimination of the APRM flow-biased simulated thermal power-high scram setpoint-adjustment requirement involves revised g MCPR and new MAPLHCR reduction factors. These limits are imposed 5

to eEsure that margins to fuel integrity limits are equal to or larger than those currently in existence. The criteria by which these changes were judged include the following:

- The MCPR safety limit shall not be violated.

Fuel performance shall remain within design and licensing bases, and

- PCT and maximum cladding oxidation fractions shall remain within regulatory limits.

Based on the above criteria, elimination of the APRM flow-biased simulated thermal power-high scram setpoint-adjustment requirement is judged to meet the first concern for significant hazards consideration.

Operation at RATED THERMAL POWER with feedwate temperatures as low as 370*F (and ' equivalent operation at lower thermal power) was evaluated as described in the attached MIOD analysis. The evaluation considered the Chapter 15 transient evaluations, the Chapter 6 LOCA evaluation, fuel mechanical limits and thermal-hydraulic stability and the effects of flow-induced and acoustic loads on the vessel internals. All results remain within design and regulatory limits. Rated power operation with feedvater temperatures down to 370'F therefore does not inco1ve a significant hazards consideration.

The reload analysis allows operation with two new fuel types.

MAPLHCR curves for these new fuels are being added to the Technical Specifications to ensure that ECCS peak clad temperature (PCT) and LHGR limits are not exceeded as established by applicable analyses.

Analysis shows that the PCT for the two new fuel types (for the postulated DBA LOCA) is 2078'T which is well below the 10CFR50.46 limit of 2200*F. Therefore, this change does not increase the probability or consequences of any accident previously analyzed.

Attech :nt 5 to U-601239 Page 3 of 3

2) The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

The MEOD analysis effectively provides for normal plant operation in an increased area of the power-flow operating map. While the events previously analyzed may be initiated from a new operating point, these events were addressed in item 1 above. There are no new or different accidents, created by the MEOD related changes.

The elimination of the APRM flow-biased simulated thermal power-high scram setpoint-adjustment requirement itself involves no physical design changes. With the incorporation of the new MCPR and MAPLHGR limits, plant operation does not change. Therefore, no new or different accident is created by these changes.

Operation with reduced feedvater temperatures involves normal plant operating practice, and no new or different accidents are created in this mode of operation.

Operation under the provisions of the reload analysis does not change any mode of plant operation and therefore does not create the possibility of a new or different kind of accident. ,

3) The proposed change does not involve a significant reduction in the margin of safety.

The provisions for reloading the reactor and operation within'the MIOD (including operation with reduced feedwater temperature and without the APRM setpoint adjustment) have been completely evaluated. Revised limits and setpoints have been established which maintain or increase the margin of safety provided by the current values. As noted above, the containment response to a DBA initiated from the MEOD with reduced feedvater temperatures resulted in a slightly higher drywell differential pressure than was determined in the original FSAR evaluation. However, this does not constitute a significant reduction in the margin of safety.

_ ___ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

ML20154A203

Text

1.0 INTRODUCTION

REFERENCES:

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