ML20151G123

Amend 142 to License DPR-50,revising Tech Specs for Cycle 7 Operation

ML20151G123
Person / Time
Site:	Crane
Issue date:	07/18/1988
From:	Stolz J Office of Nuclear Reactor Regulation
To:	Metropolitan Edison Co, Jersey Central Power & Light Co, Pennsylvania Electric Co, GPU Nuclear Corp
Shared Package
ML20151G127	List: ML20151G123 ML20151G129 ML20151G135
References
DPR-50-A-142 NUDOCS 8807280144
Download: ML20151G123 (42)

Text

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UNITED STATES

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NUCLEAR REGULATORY COMMISSION r.

-l WASHINGTON, D. C. 20$55

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METROPOLITAN _ED_I_S_0N_ _ COMP _A_NY JERSEY CENTRAL POWER & LIGHT COMPANY PENNSYLVANIA ELECTRIC COMPANY GPU NU_ CLEAR CORPORATI0N DOCKET NO. 5_0 _289 THREE MILE ISLAND NUCLEAR STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.142 License No. DPR-50 1.

The Nuclear Regulatory Comission (the Conn.ission) has found that:

A.

The application for amendment by GPU Nuclear Corporation, et al.

(the licensee) dated April 5,1988 corrplies with the standards and requirements of the Aton.ic Energy Act of 1954, as arnended (the Act),

and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in confonnity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Coennission's regulations; D.

The issuance of this ainenchnent will not be inimical to the comon defense and security or to the health and safety of the public; and E.

The issuance of this ainendment is in accordance with 10 CFR Part 51 of the Connission's regulations and all applicable requireirents have been satisfied.

!!O S0oN OS 9

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2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachtrent to this license amendment, and paragraph 2.c.(2) of facility Operating License No. DPR-50 is hereby amended to read as follows:

(2) Technic _al Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.142, are hereby incorporated in the license.

GPU Nuclear Corporation shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance.

FOR THE NUCLEAR REGULATORY COPRISSION Jh F. Stolz, Direc Pr ect Directorate I-4 vision of Reactor Projects I/II Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: July 18,1988

ATTACHMENT TO LICENSE AMENDMENT NO. 142 FACILITY OPERATING LICENSE NO. DPR-50 DOCKET NO. 50-289 Replace the following pages of the Appendix A Technical Specifications with the attached pages. The revised pages are identified by amendment number and contain vertical lines indicating the area of change.

Remove Insert i

i vi vi vii vii viii viii 1-5 1-5 2-1 2-1 2-2 2-2 2-3 2-3 Figure 2.1-1 Figure 2.1-1 Figure 2.1-2 Figure 2.1-2 Figure 2.1-3 Figure 2.1-3 2-5 2-5 2-6 2-6 2-7 2-7 2-8 2-8 2-9 2-9 2-10 (Table 2.3-1)

Figure 2.3-1 Figure 2.3-1 Figure 2.3-2 Figure 2.3-2 3-34 3-34 3-34a 3-34a 3-35 3-35 3-35a, 3-35h 3-35a 3-36 3-36 3-36a 3-36a Figure 3.5-2A Figure 3.5-2A thru Figure 3.5-8L thru Figure 3.5-2M 5-4 5-4

TABLE OF CONTENTS Section pg TECHNICAL SPECIFICATIONS 1

DEFINITIONS 1 -1 1.1

_ RATED POWER 1.2 REACTOR OPERATING CONDITIONS 1 -1 1.2.1 cold snutdown 11 1.2.2 Hot Shutdown 1 -1 1.2.3 Reactor Critical 1 -1 1.2.4 Not Standby 1 -1 1.2.5 Power Operation 1 -1 1.2.6 Refueling Shutdown 1 -1 1.2.7 Refueling Operatfor 1 -1 1.2.8 Refueling Interval 1-2 tartup 1-2 1.2.9

{ Avg 1-2 1.2.10 1.2.11 Heatup-Cooldown Mode 1-2 1-2 1.2.12 Station, Unit, Plant, and Facility 1-2 1.3 OPERABLE 1-2 1.4 PROTECTIVE INSTRUENTATION LOGIC 1.4.1 Instrument channel 1-2 1.4.2 Reactor Protection System 1-2 1.4.3 Protection Channel 1-2 1-3 1.4.4 Reactor Protection System Logic 1-3 1.4.5 Engineered Safety Features System 1.4.6 Degree of Redundancy 1-3 1.5 INBTRUENTATION SURVEILLANCE 1-3 1.5.1 Trip Test 1-3 1.5.2 Channel Test 1-3 1.5.3 Channel check 1-3 1-4 1.5.4 Channel Calibration 1-4 1.5.5 Heat Balance Check 1-4 1.5.6 Heat Balance Calibration 1-4 1.6 POWER O!$TRIBUT!0N 1-5 1.6.1 Quadrant Power T11t 1-5 1.6.2 Axial Power Imbalance 1-5 l

1.7 CONTAINMENT INTEGRITY 1-5 1.8 FIRE SUPRE5SION WATER SYSTEM 1-5 1.12 post EgulyAttm1 1 -1 Ji 1-6 1.13 sounct cntcx 1-6 1.14 sgLIoiricsTiON 1-6 1.15 WITTMRITTALCULATION MANUAL 1-6 1.16 run,t55 cunrn0L PRosRAM 1-6 1.17 sAstous RAnwAsit rRgaimMT SYSTEM 1-6 1.18 vtmTILATION EIMAUIT TREATMKT 5Y5 TEM l-6 1.19 runst-ruRsIns 1-7 1.20 vruTING 1 -7 1.21 WPIRTK8LE EVENT 1-7 1.22 mnsEAm DF THE PUBLIC 1-7 Am.hduent No. }(, Jtf, g, ))/, ~142

LIST OF TABLES Th6LE TITLE PAGE 1.2 Frequency Wotation 1-8 2.3-1 Reactor Protection System Trip Setting Limits 2-9 3.1.6.1 Pressure Isolation Check Yalves Between the Primary 3-15a Coolant System and LPtS

3. 5-1 Instruments Operating Conditions 3-29 3.5-1A Quadrant Tilt Limits 3-34a

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3.5-2 Accident Monitoring Instruments 3-40c 3.5 3 Post Accident Monitoring Instrumentation 3-40d 3.18-1 Fire Detection Instruments 3-87 3.21 -1 Radioactive Liquid Effluent Monitoring Instrumentation 3-97 3. 21 -2 Radioactive Gaseous Process and Effluent 3-101 Monitoring Instrumentation 3.23-1 Radiological Envirormental Monitoring Program 3-122 3.23-2 Reporting Levels for Radioactivity Concentration 3-126 in Environmental Samples 4.1 -1 Instrument Suneillances Requirements 4-3 4.1 -2 Minimum Equipment Test Frequency 4-8 4.1-3 Ninimum Sampling Frequency 4-9 4.1 -4 Post Accident Monitorfag Instrumentation 4-1 04 4.19-1 Minimum Naber of Steam Generators to be 4-84 Inspected During Inservice Inspection 4.19-2 Steam Generator Tube Inspection 4-85 4. 21 -1 Radioactive Liquid Effluent Monitoring 4-88 Instrumentation Survei11ance Requf resents 4. 21 -2 Radioactive Gaseous Effluent Monitoring 4-91 Instrumentation Surveillance Requirements 4.22-1 Radioactive Liquid Waste Sampling & Analysis Program 4-96 4.22-2 Radioectfwe Gaseous Waste Samp1thg 8 Analysis Program 4-102 4.23-1 Maxima Values for the Lower Limits of Detection (LLD) 4-118 vi Amendment No. M, X 186. 346 W. 47,142

1

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LIST OF FIGURES Figure Title 2.1 -1 TMI-1 Core Protection Safety Limit 2.1 -2 TMI-1 Core' Protection Safety Limits j

2.1-3 TNI-1 Core Protection Safety Bases

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2.3-1 TMI-1 Protection System Maximum Allowable Set Points 2.3-2 Protection System Maximum Allowable Set Points for Axial Power Imb11ance, TMI-1 3.1 -1 Reactor Coolant System Heatup/Cooldown Limitations (Applicable to 10 EFPY) 3.1-2 Peactor Coolant System, Inservice Leak and Hydrostatic Test Limitations (Applicable to 10 EFPY) 3.1 -3 Limiting Pressure vs. Temperature Curse for 100 STD cc/ Liter H O 2

3.5-2A Rod Position Setpoints for 4 Pump Operation from 0 to 40+10/-0 EFPD, TMI-1 3.5-2B Rod Position Setpoints for 4 Pump Operation from 40+10/-0 to 100+10/-0 EFPD, TMI-1 3.5-2C Rod Position Setpoints for 4 Pump Operation after 100+10/-0 EFPD, TMI-1 3.5-2D Rod Position Setpoints for 3 Pump Operation from 0 to 40+10/-0 EFPD,TMI-1 3.5-2E Rod Position Setpoints for 3 Pump Operation from 40+10/-0 to 100+10/-0 EFPD. TMI-1 3.5-2F Rod Position Setpoints for 3 Pitap Operation after 100+10/-0 EFPD, TMI-1 3.5-26 Rod Position Setpoints for 2 Pump Operation from 0 to 40+10/-0 EFPD, TMI-1 3.5-2H Rod Position Satpoints for 2 Pump Operation from 40+10/-0 to 100+10/-0 EFPD, TMI-1 3.5-2I Rod Position Setpoints for 2 Pump Operation after 100+10/-0 EFPD, TMI-1 3.5-2J Axial Power Zubalance Envelope for Operation from 0 to 40+10/-0 EFPD, TNI-1 vii Amendment Ibs. W, /W, W, W, (?, W, M, W, )Q$, )Qp, 7/g, T/S, YM, 142 J

LIST OF FIGURES Figure Title 3.5-2K Axial Power Imbalance Cnvelope for Operation from 40+10/-0 to 100+10/-0 EFPD, TMI-1 3.5-2L Axial Power Is6alance Envelope for Operation af ter 100+10/-0 EFPD, TMI-1 3.5-2M LOCA Limited Maximum Allowable Linear Heat Rate l

3. 5-1 Incore Instrumentation Specification Axial Imbalance Indication, TMI-1 3.5-2 Incore Instrumentation Specification Radial Flux Tilt Indication,T MI-1 1

3.5-3 Incore Instrumentation Specification 3.11-1 Transfer Path to and from Cask Londing Pit 4.17-1 Snubber Functional Test - Sample Plan 2 5-1 Extended Plot Plan TN!

5-2 Site Topography 5 Mile Radius 5-3 Site Boundary for Gaseous Effluents 5-4 Site Boundary for Liquid Effluents 6-1 GPU Nuclear Corporation Organization Chart 6-2 TMI-1 Onsite Organization l

i viii Amendment Nos. 72,77,128,142 l

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- - - - - -. ~.--- --- --

1.6 POWER DISTR!BUT!ON 1.6.1 QUADRANT POWER TfLT Quadrant power tilt is defined by the following equation and is expressed in percent.

~

100 Power in any core quadrant

-1

_ Average power of all quadrants j

The quadrant tilt limits are stated in Specification 3.5.2.4.

1.6.2 AXIAL POWER IMBALANCE l

Axial power 12alance is the power in the top half of the core minus l

the power in the bottom half of the core expressed as a percentage of rated power. Idalance is monitored continuously by the RPS using input from the power range channels. Imbalance limits are defined in Specification 2.1 and isbalance setpoints are defined in Specification 2.3.

i.7 CONTAINMENT INTEGRITY CAtainment integrity exists when the following conditions are satisfied:

a. The equipment hatch is closed and sealed and both doors of the personnel hatch and emergency hatch are closed and sealed except as in "b" or "f" below,

b. At least one door on cach of the personnel hatch and emergency hatch is closed and sealed during refueling or personnel passage through these hatches,

c. All nonautomatic containment isolation valves and blind flanges are closed as required by the "Containment Integrity Check List" attached to the operating procedure "Containment Integrity and Access Limits".

d. All automatic containment isolation valves are operable or locked closed,

e. The containment leakage determined at the last testing interval satisfies Specification 4.4.1.

f. One door of the personnel hatch or emergency hatch may be open for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for maintenance, repair or modification provided the other door of the hatch is maintained closed and has been leak tested and found to meet the local leak rate criteria for door seals within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to the maintenance, repair or undification.

1.8 FIRE SUPPRESSION WATER SYSTEM A FIRE SUPPRESSION WATER SYSTEM shall consist of: a water source, gravity tank or pump and distribution piping with associated sectionalizing control or isolation valves. Such valves include yard hydrant curb valves, and the first valve upstream of the water flow alarm device on each sprinkler, hose standpipe or spray system riser.

Amendment No. W.142 1-5

2.

SAFETY L2MITS AND LIMITING SAFETY SYSTEM SETTING _S 2.1 SAFETY LIMITS, REACTOR CORE A plicability J

Applies to reactor thermal power, axial power fealance, reactor l

coolant system pressure, coolant temperature, and coolant flow during power operation of the plant.

Objective To maintain the integrity of the fuel cladding.

Speci fication, 2.1.1 The combination of the reactor system pressure and coolant temperature shall not exceed the safety if mit as defined by i

the locus of points established in Figure 2.1-1. If the actual pressure / temperature point it below and to the right of the Ifne, the safety limit is exceeded.

I 2.1.2 The codination of reactor thermal power and axial power l

imbalance (power in the top half of core minus the power in the bottom half of the core expressed as a percentage of the i

rated power) shall not exceed the safety limit as defined by the locus of points (solid line) for the specified flow set forth in Figure 2.1-2. If the actual-reactor-thermal-power /

axial-power-12alance point is above the line for the specified flow, the safety limit is exceeded.

Bases To maintain the integrity of the fuel cladding and to prevent fission product release, it is necessary to prevent oierheating of the cladding under normal operating conditions. This is accouplished by operating within the nucleate boiling regime of heat transfer, wherein the heat transfer coefficient is large enough so that the clad surface temperature is only slightly greater than the coolant tesperature.

The upper boundary of the nucleate boiling regime is termed, departure from nucleate boiling (DNB). At this point there is a sharp reduction of the heat transfer coefficient, which could l

result in excessive cladding tenperature and the possibility of cladding failure. Although DNB is not an observable parameter during reactor operation, the observable parameters of neutron power, react ~ coolant flow, temperature, and pressure can be related to L8 through the use of a critical heat flux (CHF) correlation. The 84W-2(1) and BWC(2) correlations have been developed to predict DNS and the locatfor, of DNS for axially uniform i

and non-uniform heat flux distributions. The 84W-2 correlation applies to ;4 art-B fuel and the BWC correlation applies to Mark BZ l

fuel. The local DNB ratio (DNBR), defined as the ratio of the heat flux that would cause DN8 at a particular core location to the i

actual heat flux, is indicative of the margin to DNB. The minimiin value of the DNBR, during steady-state operation, normal operational 4

Amendsent No. }7,142

transients, and anticipated transients is limited to 1.30 (B&W-2)

. and 1.18 (BWC). A DNBR of 1.30 (B&W-2) or 1.18 (BWC) corresponds to a 95 percent probability at a 95 percent confidence level that DNB will not occur; this is considered a conservative margin to DNB for all operating conditions. The difference between the actual core outlet pressure and the indicated reactor coolant system pressure has oeen considered in determining the core protection safety limits.

The curve presented in Figure 2.1-1 represents the conditions at which the minimum allowable DNBR or greater is predicted for the If miting combination of thermal power and number of operating reacter coolant pumps.

This curve is based on the following nuclear power peaking factors (3):

N N

N F

= 2.82, F

= 1. 71 ; F

= 1.65 q

AH z

The 1.65 cosine axial flux shape in conjunction with FN AH = 1.71 define the reference design peaking condition in the core for opera-tion at the maximum overpower.

Once the reference peaking condition and the associ6ted thermal-hydraulic situation has been established for the hot channel, then all other codinations of axial flux shapes and their accompanying radials must result in a condition which will not violate the previously established design criteria on DNBR. The flux shapes examined include a wide range of positive and negative offset for steady state and transient conditions.

These design limit power peaking factors are the most restrictive calculated at full power for the range from all control rods fully withdrawn to maximum allowable control rod insertion, and form the core DNBR design basis.

The curves of Figure 2.1-2 are based on the more restrictive of two thermal limits and include the effects of potential fuel densification and fuel rod bowing:

o a.

The DNBR Ifpit produced by a nuclear power peaking I

factor of F" 2.82 of the codination of the radial peak, axial peak,9 =d position of the axial peak that yields no l

an 1ess than the DNBR limit.

b.

The combination of radial and axial peak that prevents central fuel melting at the hot spot.

The limit is 70.50 kW/ft.

Power peaking is not a directly observable quantity and therefore limits have been established on the basis of the axial power l

12alance produced by the power peaking.

2-2 Amendsent No.1/j, 53, 9g,1/h,142

The specified flow rates for curves 1, 2, and 3 of Figure 2.1-2 correspond to the expected rainimum flow rates with four pumps, three pumps, and one pug in each loop, respectively.

The curve of Figure 2.1-1 is the nost restrictive of all possible reactor coolant pump-maximum thermal power combinations shown in Figure 2.1-2.

The curves of Figure 2.1-3 represent the conditions at which the DNBR limit is predicted at the maximum possible thermal l

power for the nunter of reactor coolant pugs in operation or the local quality at the point of minf aum DNBR is equal to 22 percent, (B&W-2)(4), or 26 percent (BWC)(2) whichaver condition is more l

restrictive.

The maximum thermal power for three pump operation is 89.3 percent due to a power level trip produced by the flux-flow ratio (74.7 per cent flow x 1.08 = 80.6 percent power) plus the maximum calibration and instrutentation error.

The maximum thermal w2r for other reactor coolant pump conditions is produced in a similar manner.

Using a local quality limit of 22 percent (B4W-2), or 26 percent (BWC) at the point of minimum DNBR as a basis for curves 2 and 3 of Figure 2.1-3 is a conservative criterion even though the quality at the exit is higher than the quility at the point of minimum DNBR.

The DNBR as calculated by the B&W-2 or BWC correlation continually l

increases from the point of minimum DNBR, so that the exit DNBR is always higher and is a function of the pressure.

For each curve of Figure 2.1-3, a pressure-temperature point above and to the left of the curve would result in a DNBR greater than 1.30 (B&W-2) or 1.18 (BWC) or a local quality at the point of minfmum DNBR less than 22 percent (B&W-2), or 26 percent (BWC) for the particular Nactor coolant pump situation.

Curve 1 is more restrictive than any other reactor coolant pug situation becau:e any pressure / temperature point above and to the left of this curve will be zbove and to the left of the other curves.

REFERENCES (1) FSAR, Section 3.2.3.1.1 (2) BWC Correlation of Critical Heat Flux, BAW-10143P-A, Babcock & Wilcox, Lynchburg, Virginia, April 1955 I

(3) FSAR, Section 3.2.3.1.1.3

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(4) FSAR, Sectioa 3.2.3.1.1.11 l

2-3 Amendsent No., M, M, 99, )?9, }M,142

2400 2200 ACCEPTABLE OPERATION a J $ 2000 $e UNACCEPTABLE a OPERATION g 1800 v 1600 580 600 620 640 660 Reactor Outlet Temperature, OF CORE PROTECTION SAFETY LIMI* TMI-1 Figure 2.1-1 Amendment No. W.142 S

Thgrmal Power Loyol, % 120 (-43.8,112) 1 (37.8,112) ACCEPTABLE 4 PUMP OPERATION - 100 (_ -43.8,89.3) 2 (37.8,89.3) (-58.5,80.4) l CEPT. 80 (53.0,80.4) OPERATIC'i (-43.8,62.0) 3 (37.8,62.0) (-38.5.57.8) ACCEPTABLE 60 (53.0,57.8) 2,3, & 4 PUMP OPERATION 40 (-58.5,30.4) (53.0.30.4) ~ 20 l l t i t I i 8 I I I I I I ' 70 50 30.20 -10 0 10 20 30 40 50 60 70 80 Axial Power Imbalance, % Curve Reactor Coolant Flow (lb/hr) 6 1 139.8 x 10 6 2 104.5 x 10 6 3 68.8 x 10 CORE PROTECTION SAFETY LIMITS TMI-1 Amendment No. 11, 23. M. M. M. Il#. M. 142 Figure 2.1-2

r 2400 ~ 2200 1 fDT ~ .e E ~ _.2 [. 2000 2 n. Tc 8 b E d 1800 ff 1600 580 600 620 640 660 Reactor Outlet Temperature, UF Reactor Coolant Flow Curve (1bs/hr) Power Pumps Operating (Type of Limit) 1 139.8 x 106(100%)* 112% FourPumps(DN3RLir.it) 2 104. 5 x 100(74.7%) 89.4% Three Pumos (Quality Limit) 3 68.8 x 106(49.2%) 62.0% One Pump in Each Loop (Quality Limit)

• 106.5% of Cycle 1 Design Flow CORE PROTECTION SAFETY BASES TMI-1 icendment No. 50, J7p, 142 Figure 2.1-3

2.3 LIMITING SAFETV SYSTEM SETTINGS, PROTECTION INSTRUMENTATION Applicability Applies to instrumnts mnitoring reactor power, axial power l italance, reactor coolant system pressure, reactor coolant outlet temerature, flow, nuder of pumps in operation, and high reactor building pressure. Objective To provide automatic protection action to prevent any codination of proces< variables from exceeding a safety limit. Specification 2.3.1 The reactor protection system trip setting limits and the permissible bypasses for the instrument channels shall be as stated in Table 2.3-1 and Figure 2.3-2. E Bases The reactor protection system consists of four instrument channels to mnitor each of several selected plant conditions which will cause a reactor trip if any one of these conditions deviates from a pre-selected operLting range to the degree that a safety limit may be reached. The trip setting limits for protection system instrumentation are listed in Table 2.3-1. These trip setpoints are setting limits on the setpoint side of the protection system bistable comparators. The safety analysis has been based upon these protection system instrumatation trip set points plus calibration and instrumentation errors. Nuclear Overpower A reactor trip at high power level (neutron flux) is provided to i prevent damage to the fuel cladding from reactivity excursions too rapid to be detected by pressure and temperature measurements. During normal plant operations with all reactor coolant pumps I operating, reactor trip is initiated when the reactor power level i reaches 105.1% of rated power. Adding to this the possible l variation in trip set points due to calibration and instrument errors, the maximum actual power at which a trip would be actuated could be 1125, which is the value used in the safety analysis (1). 2-5 Anendnent No. J3, J7, 28, J26,142

a. Overpower trip based on flow and imbalance i

The power level trip set point produced by the reactor coolant system flow is based on a power-to-flow ratio which has been established to accommodate the nest severe thermal transient considered in the design, the loss-of-coolant flow accident from high power. Aralysis has degenstrated that the specified power to flow ratio is adequate to prevent a ONBR of less than 1.30 (B&W-2) or 1.18 (BWC) should a low flow l condition exist due to any malfunction. The power level trip set point produced by the power-to-flow ratio provides both high power level and low flow protection in the event the reactor power level increases or the reactor coolant flow rate decreases. The power level trip set point produced by the power to flow ratio provides overpower DNB protection for all modes of pug operation. For every flow rate there is a maximum permissible power level, and for every power level there is a minimum permissible low flow rate. Typical power level and low flow rate codinations for the pug situations of Table 2.3-1 are as follows:

1. Trip would occur when four reactor coolant pumps are operating if power is 108 percent and reactor flow rate is 100 percent, or flow rate is 92.5 percent and power level is 100 percent.

2. Trip would occur when three reactor coolant pumps are operating if powec is 80.6 percent and reactor flow rate is 74.7 percent or flow rate is 69.4 percent and power level is 75 percent.

3. Trip would occur when one reactor coolant pump is operating in each loop (total of two pugs operating) if the power 1' 53.1 percent and reactor flow rate is 49.2 percent or flow rate is 45.3 percent and the power level is 49 percent.

The flux / flow ratios account for the maximum calibration and instrumentation errors and the maximum variation from the average value of the RC flow signal in such a manner that the reactor protective system receives a conservative indication of the RC flow. No penalty in reactor coolant flow through the core was taken for an open core vent valve because of the core vent valve surveillance program during each refueling outage. For safety analysis calculations the maximum calibration and instrumentation errors for the power level were used. The power-idalance boundaries are established in order to prevent reactor thermal if aits from being ex::eeded. These thermal limits are either power peaking Kw/ft limits or DNBR limits. The axial l power idalance (power in the top half of the core minus power in 1 2-6 Amendment No. 73,77,25,28,39,50,226,142

the bottom half of core) reduces the power level trip produced b,v I the power-to-flow ratio so that the boundaries of Figure 2.3-2 are produced. The power-to-flow ratio reduces the power level trip and associated reactor power / axial power-12alance boundaries by 1.08 percent for a one percent flow reduction,

b. Pump Monitors The redundant pump monitors prevent the minimum core DNBR from decreasing below 1.30 (B&W-2) or 1.18 (BWC) by tripping the reactor due to the loss of reactor coolant pump (s). The pump monitors also restrict the power level for the nuder of pumps in operation.

c. Reactor coolant system pressure During a startup accident from low power or a slow rod withdrawal from high power, the system high pressure trip setpoint is reached before the nuclear overpower trip setpoint. The trip setting if mit shown in Figure 2.3-1 for high reactor coolant system pressure ensures that the system pressure is maintained below the safety limit (2750 psig) for any design transient (6).

Due to calibration and instrument errors, the safety analysis assumed a 45 psi pressure error in the high reactor coolant system pressure trip setting. As part of the post-TMI-2 accident modifications, the high pressure trip setpoint was lowered from 2390 psig to 2300 psig. (The FSAR Accident Analysis Section still uses the 2390 psig high pressure trip setpoint.) The lowering of the high pressure trip setpoint and raising of the setpoint for the Power Operated Relief Yalve (PORV), from 2255 psig to 2450 psig, has the effect of reducing the challenge rate to the PORY while maintaining ASME Code Safety Yalve capability. A B&W analysis cogleted in Septesber of 1985 concluded that the high reactor coolant system pressure trip setpoint could be raised to 2355 psig with negligible impact on the frequency of opening of the PORY during anticipated overpressurization transients (8). The high pressure trip setpoint was subsequently raised to 2355 psig. The potential safety benefit of this action is a reduction in the frequency of reactor trips. The low pressure (1800 psig) and variable low pressure (11.75 Tout-5103) trip setpoint were initially established to maintain the DNB ratio greater than or equal to 1.3 for those design accidents that result in a pressure reduction (3,4 and 7). The B4W generic ECCS analysis, however, assumed a low l pressure trip of 1900 psig and, to estabitsh conformity with this analysis, the low pressure trip setpoint has been raised to the more conservative 1900 psig. Application of the 84W 2-7 Amendment No. 31, 22, pp, 95,12, Ath, Abb,142

crossflow model resulted in safe 2y limits (see Figures 2.1-1 and 2.1-3) outside the acceptable operating region formed by the low pressure, high pressure, and high temperature trip setpoints (see Figure 2.3-1) which justifies the removal of the variable low pressure trip,

d. Coolant outlet temperature The high reactor coolant outlet temperature trip setting limit (618.8F) shown in Figure 2.3-1 has been established to prevent excessive core coolant temperature in the operating range The calibrated range of the temperature channels of the RPS is 520' to 620*F.

The trip setpoint of the channel is 618.8F. Under the worst case environment, power supply perturbations, and drift, the accuracy of the trip string is 1.2*F. This accuracy was arrived at by summing the worst case accuracies of each module. This is a conservative method of error analysis since the normal procedure is to use the root mean square method. Therefore, it is assured that a trip will occur at a value no higher than 620'F even under worst case conditions. The safety analysis used a high tempe"ature trip set point of 620*F. The calibrated range of the channel is that portion of the span of indication which has been qualified with regard to drift, if nearity, repeatability, etc. This does not igly that the equipment is restricted to operation within the calibrated range. Additional testing has demonstrated that in fact, the temerature channel is fully operational approximately los above the calibrated range. Since it has been established that the channel will trip at a value of RC outlet temperature no higher than 620*F even in (Ae worst case, and since the channel is fully operational approximately los above the calibrated range and exhibits no hysteresis or foldover characteristics, it is concluded that the instrveent design is acceptable,

e. Reactor building pressure The high reactor building pressure trip setting limit (4 prig) provides positive assurance that a reactor trip will occur in the unlikely event of a steam liree failure in the reactor building or a loss-of-coolant accident, even in the absener of a low reactor coolant system pressure trip.

I 2-8 Amendsent No. ($, /fS, YJ7,142 i

f. Shutdown bypass I

In order to provide for control rod drive tests, zero power physics testings, and startup procedures, there is provision for bypassing certain segments of the reactor protection system. The reactor protection system segments which can be bypassed are shown in Table 2.3-1. Two conditions are imposed when the bypass is used:

1. By administrative control the nuclear overpower trip set point must be reduced to value < 5.0 percent of rated power during reactor shutdown. ~

2. A high reactor coolant system pressure trip set point of 1720 psig is automatically igosed.

The purpose of the 1720 psig high pressure trip set point is to prevent normal operation with part of the reactor protection system bypassed. This high pressure trip set point is lower than the normal low pressure trip set point so that the reactor sust be tripped before the bypass is initiated. The overpower trip set point of < 5.0 percent prevents any significant reactor power from lieing produced when performing the physics tests. Sufficient natural circulation (5) would be available to nuove 5.0 percent of rated power if none of the reactor coolant pumps were operating. References (1) FSAR, Section 14.1.2.3 (2 FSAR, Section 14.1.2.2 (3) FSAR, Section 14.1.2.7 (4) FSAR, Section 14.1.2.9 l (5) FSAR, Section 14.1.2.6 (6) Technical Specification Change Request No. 31, January 16, 1976, and Technical Specification Change Request No. 84, June 23,1978. (7) "ECCS Analysis of 84W's 177-FA Lowered Loop NNS," BAW-10103-A, Rev. 3, Babcock and Wilcox, Lynchburg, Virginia, July 1977. (8) "Justification for Raising Setpoint for Reactor Trip on High Pressure," 8AW-1890, Rev. O, Babcock and Wilcox, September 1985. 2-9 Amendment No. /A, //, A8, AA, AA, /A, A0, AA6, AA6,142

Tchle 2.3-1 REACTOR PROTECTION SYSTEM TRIP SETTING LIMITS (5) l~ E k Four Reactor Coolant Three Reactor Coolant One Reacter Coolant g Pumps Oper,1 ting Pumps Operating Pump Operating in 3 (Nominal Operating (Nominal Operating Each Loop (Nominal Shutdown Pwer - 1001) Power - 751) Ope.ating Power 49%) Bypass 2*

1. Nuclear power, max.

105.1 105.1 105.1 5.0(2) I ? % of rated power 5

2. Nuclear power based on 1:08 times flow 1.08 times flow 1.08 times flow minus Bypassed

? flow (1) and idalance minus reduction due minus reduction due reduction due to g M max. of rated power to i dalance to fealance idalance T t

3. Nuclear power based NA NA 55%

Bypassed l ? (4) on pump monitors, g max. 1 of rated power

4. High reactor coolant 2355 2355 2355 1720(3) l system pressure, psig max.

Es

5. Low reactor coolant 1900 1900 1900 Bypassed system pressure, psig min.

6. Reactor coolant temp.

61 8.8 61 8.8 618.8 618.8 l F., max.

7. High Reactor Building 4

4 4 4 g pressure, psig max. (1) Reactor coolant system flow,1. l (2) Administratively controlled reduction set only during reactor shutdown. l (3) Automatically set when other segments of the RPS (as specified) are bypassed. g (4) The pump monitors also produce a trip on: (a) loss of two reactor coolant pumps in one reactor coolant luop, g and (b) loss of one or two reactor coolant pumps during two-pump operation. (5) Trip settings ifmits are setting limits on the setpoint side of the protection system bf stable connectors. I

2500 P = 2355 psig 2300 ACCEPTABLE a-OPERATION I 2100 t a. { P = 1900 psig g 1900 u UNACCEPTABLE h OPERATION t 2* 1700 1500 i i i i 540 560 580 600 620 640 Reactor Outlet Tetperature 'F 1 TMI-1 PROTECTION SYSTEM MAXIMUM ALLOWABLE SETPOINTS i Amendment No. I3, 17, 28, 39, H, Figure 2.3-1 78,126. DE,142 r

Thermal Power Levol, % - - 120 (-30.0.108) (24.5.108) ACCEPTABLE l 3 = 1.900 l 4 pygp . 100 l 2 = -1.854 m m l OPERATION l l(-30.0,80.6 ) l ~~ (24.5,80.6 ) ACCEPTABLE 80 (5.0,70.0) l (-50.0,70.0) l OP R T l (-30.0,53.1 ) 6f24.5,53.1) l l ACCEPTABLE ~ l ~ g (-50.0,42.6 ) l 2,3, & 4 (45.0,42.6 ) PUMP 40 l l OPERATION l l l l l l e o e (-50.0.15.1 ) yl 20 I d (45.0.15.1 ) l a = m c pl el l ,i, , 70 50 30 10 0 10 20 30 40 50 60 70 80 Axial Power Imbalance, % l l PROTECTION SYSTEM MAXIMUM ALLOWABLE SETPOINTS FOR AX1AL POWER IMBALANCE l THI-1 Amendment No. 17,29,35,(p,(5, Figure 2.3-2 59, 228, ) M 142

f. If a control rod in the regulating or axial power shaping groups is declared inoperable per Specification 4.7.1.2.,

operation may continue provided the rods in the group are positioned such that the rod that was declared inoperable is maintained within allowable group average position limits of Specification 4.7.1.2.

g. If the inoperable rod in Paragraph "e" above is in groups 5, 6, 7, or 8, the other rods in the group may be trinened to the same position.

Normal operation of 100 percent of the thermal power allowable for the reactor coolant pump conbination may then continue provided that the rod that was declared inoperable is maintained within allowable group average position If mits in 3.5.2.5. 3.5.2.3 The worth of single inse;ted control rods during criticality is limited by the restriction of Specification 3.1.3.5 and the Control Rod Position Limits defined in Specification 3.5.2.5. 3.5.2.4 Quadrant Tilt: a. Except for physics tests the quadrant tilt shall not exceed the values in Table 3.5-1 A as determined using l the full incore detector system. b. When the full incore detector system is not available and except for physics tests quadrant tilt shall not exceed the valuas in Table 3.5-1A as determined using l the power range channels displayed on the console for each quadrant (out of core detection system), c. When neither detector system above is available and, except for physics tests, quadrant tilt shall not exceed the values in Table 3.5-1A as determined using the minimum incere detector system. d. Except for physics tests if quadrant tilt exceeds the tilt limit, allowable power shall be reduced 2 percent l for each 1 percent tilt in excess of the tilt limit. For less than four pump operation, thermal power shall be reduced 2 percent of the thermal power allowable for the reactor coolant puuv con 6fnation for each 1 percent tilt in excess of the tilt limit. e. Within a period of 4 hours, the quadrant power tilt shall be reduced to less than the tilt limit except for physics tests, or the following adjustsents in setpoints and limits shall be made: l 3-34 Amendsent No. //, Ap, Ap, /p, Ap, AD, At,6,142

o 1. The protection system reactor power /italance i envelope trip setpoints shall be reduced 2 percent in pwer for.each 1 percent tilt, in excess of the tilt If mit, or when thermal power is equal to or less than 50% full power with four reactor coolant pumps running, set the nuclear overpower trip setpoint equal to or less than 60% full power. 2. The control rod group withdrawal ifmits (Figures 3.5-2A to 3.5-2!) shall be reduced 2 percent in power for each 1 percent tilt in excess of the tilt limit. ] 3. The operational iealance limits (Figures 3.5-2J. l 3.5-2K, and 3.5-2L) shall be reduced 2 percent in l power for each 1 percent tilt in excess of the tilt limit. f. Except for physics or diagnostic testing, if quadrant tilt is in excess of +16.80% determined using the full incore detector syste,n (FIT), or +14.2% determined using the out of core detector system (0CT) if the FIT is not available, or +9.5% using the minimum incore detector system (MIT) when neither the FIT nor OCT are available, the reactor will be placed in the hot shutdown condition. Diagnostic testing during power operation with a quadrant tilt is permitted provided that the thermal power allowable is restricted as I l stated in 3.5.2.4.d above, g. Quadrant tilt shall be nonitored on a minimum frequency of once every two hours during power operation above 15 percent of rated power. Table 3.5-1 A - Quadrant Tilt Limits l Tilt Limit Tilt Limit i (indicated power (indicated power s 50%) > 50%) Quadrant Tilt as Indicated By: Full incore detector 6.831 4.125 system Power range channels 4.05% 1.965 Minimum incore 2.805 1.90% detector system 3-34a Amendment No. W,,fW, W, AQ, W, $Q, W, M,142

\\ 3.5.2.5 Control Rod positions a. Operating rod group overlap shall not exceed 25 percent 15 percent, between two sequential groups l except for physics tests. ) b. Position limits are specified for regulating control rods. Except for physics tests or exercising control rods, the regulating control rod insertion / withdrawal limits are specified on Figures 3.5-2A, 3.5-2B, and 3.5-2C for four pump operation and Figures 3.5-20, 3.5-2E, and 3.5-2F for three pumn operation. Two pump operation is specified on Figures 3.5-2G, 3.5-2H, and 3.5-21. If any of these control rod position If mits are exceeded, corrective measures shall be taken immediately to achieve an acceptable control rod position. Acceptable control rod positions shall be attained within four hours, c. Deleted d. Axial power frbalance shall be monitored on a minimum frequency of once every two hours during power operation above 40 percent of rated power. Except for physics tests, corrective measures (reduction of imbalance by APSR movements and/or reduction in reactor power) shall be taken to maintain operation within the envelopes defined by Figures 3.5-2J, 3.5-2X, and 3.5-2L. If the 12alance is not within l the enve* opes defined by Figures 3.5-2J. 3.5-2X, or 3.5-2L at the appropriate time in cycle, corrective l measures shall be taken to achieve an acceptable imbalance. If an acceptable 12alance is not achieved within four hours, reactor power shall be reduced until 12alance limits are set. e. Safety rod limits are given in 3.1.3.5. 3.5.2.6 The control rod drive patch panels shall be locked at all times with limited access to be authorized by the superintendent. 3.5.2.7 A power map shall be taken at intervals not to exceed 30 effective full power days using the incore instrumentation detection system to verify the power distribution is within the limits shown in Figure 3.5-2M. l 3-35 Amendsent No. )p, )/, Jp, pp, )), Jp, JJS, JJA,142

- Bases The axial power 12alance envelopes defined in Figures 3.5-2J, 3.5-2K, and 3.5-2L are based on LOCA analyses which have defined the maximum linear heat rate (see Figure 3.5-2M) such that the maximum clad tengerature will not exceed the Final Acceptance Criteria (2200'F). Operation outside of the axial nower tubalance envelope l alone does not constitute a situation that would cause the Final Acceptance Criteria to be exceeded should a LOCA occur. The axial l power imbalance envelope represents the boundary of operation If mited by the Final Acceptance Criteria only if the control rods are at the withdrawal / insertion If mits as defined by Figures 3.5-2A, 3.5-2B, 3.5-20, 3.5-20, 3.5-2E, 3.5-2F, 3.5-2G, 3.5-2H, 3.5-21, and if quadrant tilt is at the If mit. The effects of the gray APSRs are also included. Additional conservatism is introduced by application of:

a. Nuclear uncertainty factors

b. Therrai ct'ibration uncertainty

c. Fuel ar6tification effects

d. Hot rod manufacturing tolerance factors

e. Postulated fuel rod bow effects

f. Peaking if mits based on initial redition for Loss of Coolant Flow transients.

The axial power tu6alance envelopes given in Figures 3.5-2J, 3.5-2X, and 3.5-2L have been error adjusted for observability and measurement uncertainties. Therefore, the Ifmits specified in these figures are the maximum axial power is6alance alarm setpoints for power operation. The Rod index versus Allowable Power curves of Figures 3.5-2A, 3.5-28, 3.5-20, 3.5 20, 3.5-2E, 3.5-2F, 3.5-2G, 3.5-2H, and 3.5-21 describe three regions. These three regions are:

1. Permissible operating Region

2. Restricted Regions

3. Prohibited Region (0peration in this region is not allowed)

NOTE: Inadvertent operation within the Restricted Region for a period of four hours is not considered a violation of a Itatting conditten for operatten, The 1 tatting criteria within the,*.estricted Region are potential ejected rod worth - and ECCS power peaking and since the protiabtitty of these. accidents is very low, especially in a 4 hour time frame, inadvertant operation within the Restricted Region for a period of 4 hours is allowed. 3-35a Amendent No. }}. M. M. Me 99. M9 99,142 L I

The 25+5 percent overlap between successive control rod groups is alloweif since the worth of a rod is lower at the upper and lower part of the stroke. Control rods are arranged in groups or banks defined as follows: Group Function 1 Safety 2 Safety 3 Safety 4 Safety 5 Regulating 6 Regulating 7 Regulating 8 APSR (axial power shaping rod bank) Control rod groups are withdrawn in sequence beginning with group 1 Groups 5,6 and 7 are overlapped 25 percent. The normal por.ition at power is for group 7 to be partially inserted, The rod position If mits are based on the most ifmiting of the following three criteria: ECCS power peaking, shutdown margin, and potential ejected rod worth. As discussed Aove, compliance with the ECCS power peaking criterion is ensured by the rod position t limits. The minimum ava'lable rod worth, consistent with the rod position limits, provides for achieving hot shutdown by reactor urIp at any time, assuming the highest worth control rod that is withdrawn remains in the full out position (1). The rod position ifmits also ensure that inserted rod groups will not contain single rod warthe greater than: 0.65% Ak/k at rated power. These values have been shown to be safe by the safety analysis (2) of the hypothetical rod ejection accident. A maximum single inserted control rod worth of 1.0% Ak/k is allowed by the rod position limits at hot zero power. A si'1gle inserted control rod worth 1.0% Ak/k at beginning of If fe, hot, zero power would result in a lower transient peak thermal power and, therefore, less severe environmental consequences than 0.65% Ak/k ejected rod worth at rated power. The rod position lim;ts given in Figures 3.5-2A, 3.5-28, 3.5-20, 3.5-20, 3.5-2E, 3.5-2F, 3.5-2G, 3.5-2H, and 3.5-2I have been error adjusted for observability and measurement uncertainties. Therefore, the limits specified in these figures are the maximum rod position alare setpoints for operation. The plant computer will scan for tilt and is6alance and will satisfy the technical specification requirements. If the cog uter is out of service, then manual calculation for tilt above 15 percent power and imbalance above 40 percent power must be performed at least every two hours until the computer is returned to service. l 3-36 Amendment No. 77,/29, 79, (9, 59, 726, 142

The quadrant power tilt limits for thermal power greater than 50% l set forth in Specification 3.5.2.4 have been established within the thermal analysis design base using an actual core tilt of +4.92% which is equivalent to a +4.12% tilt measured with the full incere instrumentation with statistically combined measurement uncertainties included. The quadrant power tilt limits for thermal power less than or equal to 50% set forth in Specification 3.5.2.4 have been established within the thermal analysis design base using i an actual core tilt of +7.50% which is equivalent to a +6.83% tilt I measured with the full incore instrumentation with statistically conbined measurement uncertainties included. The maximum allowable quadrant power tilt setpoint of +16.8% tilt measured with the full incore detector system represents a +20% actual core tilt and includes bounding measurement uncertainty allowances. Reduction of the nuclear overpower trip setpoint to 60% full power when thermal power is equal to or less than 50% full power maintains both core protection and an operability margin at reduced power similar to that at full power. During the physics testing program, the high flux trip setpoints are administrative 1y set as follows to assure an additional safety margin is provided: Test power Test Setpoint 0 <5% l 15 50% i 40 50% 50 60% 75 85% >75 105.1% REFERENCES (1) FSAR, Section 3.2.2.1.2 1 (2) FSAR, Section 14.2.2.2 l l 1 l 3-36a Amendeent No. M, IM,142

(300,102) 100 (275.9.102)7 SHUT 00WN MARGIN (273.5,90) 90 NOT ALLOWED LIMIT 80 (249.5,78) ul 70 RESTRICTED i n. o j 60 a: 50 (38.5,48) (201.5,48) J 40 l i 30 6 20 PERMISSIBLE j (0,11.5) ) 10 ' 0, (0,,2.6)i i 0 25 50 75 100 125 150 175 200 225 250 275 300 Indicated Rod Index, 5 W1thdrawn 0 25 50 75 100 i i f t Group 7 0 25 50 75 100 .i i i l Group 6 0 25 50 75 100 l I i i I Group 5 ROD POSITION SETPOINTS FOR 4 PUMP OPERATION FROM 0 TO 40 +10/-0 EFPD THI-1 Amendment No. 17, 29, 39, 59, Wti',142 Figure 3.5-2A ~ ~ ) i

(92.5.102) (300,102) 100 (270.1.102) I 90 SHUTDOWN (266.5,90) MARGIN NOT ALLOWE0 LIMIT 80 (24 9.5,78) u 70 c-RESTRICTE0 ]o 60 a (38.5,48) (201.5,48) u 40 Eo 30 b 0 (0,11.5) PERMISSIBLE ~ 10 ' 0,2.6) o i i e i i i i i 0 25 50 75 100 125 150 175 200 225 250 275 300 Indicated Rod Index, % Withdrawn 0 25 50 75 100 t l l t l Group 7 0 25 50 75 100 l i I l Group 6 0 25 50 75 100 l l t l l Group 5 ROD POSITION SETPOINTS FOR 4 PUMP OPERATION FROM 40 +10/-0 TO 100 +10/-0 EFPD THI-1 Amendment No. IB, 17, 29, 39, 6, 50, YF/, 142 Figum 3.5-2B

(300,102) I '0' ) (266.5.102) 100 j m (266.5,90) q0 SHUTOC4fN MARGIN ~ (249.5,78) u Wf 70 RESTRICTED ?, NOT ALLOWED 60 o (116.5,48) (201.5,48) u g 40 c. 30 20 PERMISSIBLE 3 10 (58.5,13) O' I.0,2. 7 ), 0 25 50 75 100 125 150 175 200 225 250 275 300 Indicated Rod.inciex, % Withdrawn O 25 50 75 100 l i I I I Group 7 0 25 50 75 100 1 I I i i Group 6 0 25 50 75 100 t i I l i Group 5 ROD POSITION SETP0!i35 FOR 4 PUMP OPERATION AFILf( 100 +10/-0 EFPD THI-1 Amendment No. 17, 29, 39, 59, 7/$,142

l t 100 90 (300,77) 80 (93.2,77) (276.0,77) g 70 (273.5,67) [ NOT ALLOWED SHUTDOWN MARGIN 60 3 LIMIT (249.5,58) f 50 RESTRICTED C 40 b (38.5,36) (201.5,35.5) [ 30 b 20 .9 3 10, (0.8.6) PERMISSIBLE o< O'lf) i i i i I 0 25 50 75 100 125 150 175 200 225 250 275 300 Irdicated Rod Index, % Withdrawn 0 25 50 75 100 i f f I i 6POUP 7 0 25 50 75 100 l l l I I Group 6 0 25 50 75 100 1 l i l i Group 5 ROD POSITION SETPOINTS FOR 3 PUMP OPERATION FROM 0 TO 40 +10/-0 EFPD TMI-1 Figure 3.5-20 Amendnent No. 17, 29, 39, 6, H, 26/,142

100 90 80 ( } (93.2,77) (270.3,77) '- bg 70 c. NOT ALLOWED SHllTDOWN (256.5,67) E 60 MARGIN g LIMIT (249.5,58) 50 RESTRICTED w 40 f (38.5,36) (201.5,35.5) 30 T> 20 .a 3 (0,8.8) PE!*ISSIBLE 19 oI 0,1.4) f f i t i f f I l t 0 25 50 75 100 125 150 175 200 225 250 275 300 Indicated Rod Index, % Withdrawn O 25 50 75 100 l I f l Group 7 0 25 50 75 100 L l i I I Group 6 0 25 50 75 100 l i I I I Group 5 ROD POSITION SETPOINTS FOR 3 PUMP OPERATION FROM 40 +10/-0 TO 100 +10/-0 EFP0 TMI-1 ~ Figure 3.5-2E Amendment No. 17, 29, 39, 6, 50,120, aff,142 l

- W 4* 100 90 80 (198.5,77) (266.5,77) s. g 70 SHUTDOWN (266.5,67) N g NOT ALLOWED 60 j (249.5,58) = RESTRICTED 50 g w 40 g (116.5,36) (201.5.35.5) E 30 b 20 2 10 (58.5,9.7) PERMISSIBLE 0' @,2. 0 ), 0 25 50 75 100 125 150 175 200 225 250 275 300 Indicated Rod Index, % Withdrawn O 25 50 75 100 t t I i I Group 7 0 25 50 75 100 1 I I f I O 25 50 75 100 I i f I i Group 5 ROD POSITION SETPOINTS FOR 3 PUMP OPERATION AFTER 100 +10/-0 EFPD THI-1 1 Figure 3.5-2F Amendment No. 17, 29, 39, 45, 56,129, YN,142

i 100 90 80 D 70 E .I SHUTDOWN 60 3 MARGIN 'MI (300,52) f, 50 (94.5,52) (276.2,52)i o NOT ALLOWED (?73.5,44) 40 (249.5,38) k RESTRICTED s 30

• ' 4)

(201.5,23) 'O d 5 10 0 5 7) PERMISSIBLE (0,0.3) 0 i e i i i e i i 0 25 50 75 100 125 150 175 200 225 250 275 300 Indicated Rod Index, % Withdrawn O 25 50 75 100 i f f I l Group 7 0 25 50 75 100 t i l 1 i Group 6 0 25 50 75 100 i f f i i Group 5 ROD POSITION SETPOINTS FOR 2 PUMP OPERATION FROM 0 TO 40 +10/-0 EFPD TISI-1 Figure 3.5-2G Amendment No. 17, 29, 39, 59, 99. T/9,142

k 100 90 80 70 k n. SHUT 00WN 60 o3 MARGIN } (94.5,52) (270.5,52) 50 NOT ALLOWED (266.5,44) 40 RESTRICTED 5-

• 0' a) 30 20 (38.5,24)

(201.5,23) 3 10 ISSIBLE (0,0.3) i O' i i 25 250 275 300 M tcd Rod Index, ; Withdrawn 0 25 50 75 100 f I i 0 25 50 75 100 I 6 0 25 50 75 100 Group 5 R00 POSITION SETPOINTS FOR 2 PUMP OPERATION FROM 40 +10/-0 T0 100 +10/-0 EFPD THI-1 kendment No. 29, 39, (p, (3, 50,120, /!/,6,142 Figure 3.5-2H

100 90 80 b d 70 ? g" 60 50 (200.5,52) (300.32) o (266.5,52) SHUTDOWN L NOT ALLOWED MARGIN (266.5,44) E 40 LIMIT RESTRICTED 2 (249.5.38) ] 30 20 (201.5.23) 10 PERMISSIBLE (0,1.3) (58.5,6.5) 0i i i i e i i 0 25 50 75 100 125 150 175 200 225 250 ^75 300 Indicated Rod Index, % Withdrawn i 0 25 50 75 100 i t t t t Group 7 0 25 50 75 100 l I i l l Group 6 0 25 50 75 100 I t t l l Group 5 ROD POSITION SETPOINTS FOR 2 PUMP OPERATION AFTER 100 +10/-0 EFPD THI-1 Figure 3.5-21 Amendment flo.129, JM,142

Indicated Power, % of Rated Power - 110 (-13.9,102)

(19.7,102)

(-14.1.92) (19.7,92) 4 - 90 (-22.0,80), . 30 ,(25.8,80) 1 - 70 - 60 RESTRICTED PERMISSIBLE RESTRICTED REGION OPERATING 'IGION REGION 50 40 I 30 20 10 I f f f I I I i i t -50 30 -20 -10 0 10 20 30 40 50 Indicated Axial Power Imbalance, % AXIAL POWER IMBALANCE ENVELOPE FOR OPERATION FROM 0 TO 40 +10/-0 EFPD TMI-1 Figure 3.5-2J AmendmentNo.pg,142

Iniicated Power, : of Rated Power -- 110 (-19.7,102) - (21.6.102) - 00 (-20.8,92) (21.8,92) 4 (-24.9,80)< 80 ,(27.8,80) 70 RESTRICTED PERMISSIBLE - 60 RESTRICTED REGION OPERATING REGION REGION 50 40 30 20 10 f f i f f I f f I f -50 30 10 0 10 20 30 40 50 Iaiicated Axial Power Imbalance, % AXIAL POWER IMBALANCE ENVELOPE FOR OPERATION FROM 40 +10/-0 TO 100 +10/-0 EFPD TMI-1 Amendment No. VM,142 Figure 3.5-2K

l ( Indicated Power, % of T ted Power i i - 110 ) (-22.6,102) - 100 (-22.8,92) (22.8,92) , gg (-27.8,80)i - 80 >(28.7,80) - 70 60 RESTRICTED PERMISSIBLE RESTRICTED REGION OPERATING REGION REGION 50 40 30 20 10 l i I I I I t I i i ' 40 -30 10 0 10 20 30 40 50 Indicated Axial Power Imbalance, AXIAL POWER IMBALANCE ENVELOPE FOR OPERATION AFTER 100 +10/-0 EFPD THI-1 Amendment No.1EW,142 Figure 3.5-2L

20 C s3 3 18

• z E

1. ././*

16 / 8 // 3 ,/ s / g 14 f l E 12 0-1000 mwd /mtU 1000-2600 mwd /mtU After 2600 mwd /mtU 10 t i i e O 2 4 6 8 10 12 Axial Location From Bottom of Core, ft. LOCALIMiTEDMAXIMUM ALLOWABLE LINEAR HEAT RATE TMI-1 Amendment No. 142 Figure 3.5-2M

~ ^ 5.3 REACTOR Applicability l Applies to the design features of the reactor core and reactor coolant system. Objective To define the significant design features of the reactor core and reactor coolant system. Specification 5.3.1 REACTOR CORE I 5.3.1.1 The reactor core contains approximately 93.1 metric tons of slightly enriched uranium dioxide pellets. The pellets are encapsulated in Zircaloy-4 tubing to form fuel rods. The reactor core is made up of 177 fugl qpemblies. Each fuel asse21y contains 208 fuel rods il)('l 5.3.1.2 The reactor core shall approximate a right circular cylinder with an equivalent diameter of 128.9 inches and an active height of 142 inches.(21 l 5.3.1.3 The average initial enrichment of the current core for Unit 1 is a nominal 3.02 weight percent of U235 The I highest enrichnent is less than 3.7 weight percent U235, l 5.3.1.4 There are 61 full-length control rod assemblies (CRA) and 8 axial power shaping rod assemblies (APSRA) distributed in the reactor core as shown in FSAR Figure 3.2-1. The full-length CRA contain a 134 inch length of silver-indium-cadsfum alloy clad with stainless steel.LU The gray APSRA contain a 63 inch length of Inconel. 5.3.1.5 The core will have 68 burnable poison spider asseelies with similar dimensions as the full-length control rods. The cladding will be zircaloy-4 filled with alumina-boron.

5. 3.1. 6 Reload fuel asseelfes and rods shall conform to design and evaluation described in FSAR and shall not exceed an enrichment of 4.3 percent of U235, l

5.3.2 REACTOR C0OLANT SYSTEM 5.3.2.1 The reactor coolant system shall be designed and constructed in accordance with code requirements.(4) 5.3.2.2 The reactor coolant system and any connected auxiliary systems r.xposed to the reactor coolant conditions of-temperature and pressure, shall be designed for a pressure of 2..%0 psig and a temperature of 650 F. The pressurizer and pressurizer turge line shall be designed for a tempera-ture of 670 F.(Si 5-4 Ameadment No. W6',142 t ,, + - ,---,----.-,---,-,--,_,,,----,,,,,,_m_,,,,,w-,_,_ ,.,,-,,wn,,,,,,,w,,-- -,,.,,-,,,,,,.,-..,-______,,---,,,-e,---}}