ML20151C312

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Proposed Tech Specs Changes Re Operation During Cycle 7
ML20151C312
Person / Time
Site: Three Mile Island Constellation icon.png
Issue date: 04/05/1988
From:
GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20151C300 List:
References
NUDOCS 8804120314
Download: ML20151C312 (39)
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o o TABLE OF CONTENTS Section TECHNICAL SPECIFICATIONS Page 1 DEFINITIONS 1 -1 1.1 RATED POWER l -1 1.2 REACTOR OPERATING CONDITIONS 1 -1 1.2.1 COLD SHUTDOWN 1 -1 1.2.2 HOT SHUTDOWN 1 -1 1.2.3 REACTOR CRITICAL 1 -1 1.2.4 HOT STANDBY l -1 1.2.5 POWER OPERATION 1 -1 1.2.6 REFUELING SHUTDOWN 1 -1 1.2.7 REFUELING OPERATION 1-2 1.2.8 REFUELING INTERVAL 1-2 1.2.9 STARTUP 1-2 1.2.10 TAVG l-2 1.2.11 HEATUP-COOLDOWN MODE 1-2 1.2.12 STATION, UNIT, PLANT AND FACILITY l-2 1.3 OPERABLE 1-2 1.4 PROTECTIVE INSTRUMENTATION LOGIC 1-2 1.4.1 INSTRUMENT CHANNEL l-2 1.4.2 REACTOR PROTECTION SYSTEM 1-2 1.4.3 PROTECTION CHANNEL l-3 1.4.4 REACTOR PROTECTION SYSTEM LOGIC 1-3 1.4.5 ENGINEERED SAFETY FEATURES SYSTEM 1-3 1.4.6 DEGREE OF REDUNDANCY l-3 1.5 INSTRUMENTATION SURVEILLANCE l-3

1. 5.1 TRIP TEST 1-3 1.5.2 CHANNEL TEST 1-3 1.5.3 INSTRUMENT CHANNEL CHECK l-3 1.5.4 INSTRUMENT CHANNEL CALIBRATION 1-4 1.5.5 HEAT BALANCE CHECK 1 -4 1.5.6 HEAT BALANCE CALIBRATION 1-4 1.6 POWER DISTRIBUTION 1-5 1.6.1 QUADRANT POWER TILT 1-5 1.6.2 AXIAL POWER IMBALANCE 1-5 l 1.7 CONTAINMENT INTEGRITY l-5 1.8 FIRE SUPPRESSION WATER SYSTEM 1-5 1.9 CHANNEL CALIBRATION 1-6 1.10 ERANNEL CHECK 1-6 1.11 CHANNEL TEST l-6 1.12 DOSE EQUIVALENT I-131'" 1-6a 1.13 SOURCE CHECK l-6a 1.14 3DlIDIFICATION 1-6a 1.15 0FFSITE DOSE CALCULATION MANUAL I-7 1.16 PROCESS CONTROL PROGRAM l-7 1.17 GASEOUS RADWASTE TREATMENT SYSTEM 1-7 1.18 VENTILATION EXHAUST TREATMENT SYSTEM 1 -?

1.19 PURGE-PURGING l-7 1.20 VENTING l-7 1 . 21 REPORTABLE EVENT l-7a e804120314 G80405 PDR ADOCK 05000289 p DCD -i-Amendment Nu. 11, 72, 129

LIST OF TABLES TABLE TITLE PAGE 1.2 Frequency Notation 1-8 2.3-1 Reactor Protection System Trip Setting Limits 2-10 3.1.6.1 Pressure Isolation Check Valves Between the Primary 3-15a Coole.nt System and LPIS 3.5-1 Instruments Operating Conditions 3-29 3.5-1A Quadrant Tilt Limits 3-34a 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 Liquid Effluent Monitoring Instrumentation 3-101 3.23-1 Radiological Environmental Monitoring Program 3-122 3.23-2 Reporting Levels for Radioactivity Concentration 3-126 in Environental Samples 4.1-1 Instrument Surveillance Requiremerits 4-3 4.1 -2 Minimum Equipent Test Frequency 4-8

, 4.1 -3 Minimum Sampling Frequency 4-9 4.1 -4 Post Accident Monitoring Instrumentation 4-10a 4.19-1 Minimum Number of Steam Generators to be Inspected 4-84 During Inservice Inspection 4.19-2 Steam Generator Tube Inspection 4-85 4.21 -1 Radioactive Liauid Effluent Monitoring 4-88 Instrumentation Surveillance Requirements 4.21 -2 Radioactive Gaseous Effluent Monitoring 4-91 Instrumentation Surveillance Requirements 4.22-1 Radioactive Liquid Waste Sampling & Analysis Program 4-98 4.22-2 Radioactive Gaseous Waste Sampling & Analysis Program 4-106 4.23-1 Maximum Yalves for the Loner Limits of Detection (LLD) 4-118 vi Amendment No. 59, 72, 100, 106, 118

o -

LIST OF FIGURES Figure Title 2.1 -1 TMI-1 Core Protection Safety Limit 2.1 -2 TMI-1 Core Protection Safety Limits 2.1 -3 TMI-1 Core Protection Safety Bases 2.3-1 THI-1 Protection System Maximum Allowable Set Points 2.3-2 Protection System Maximum Allowable Set Points for Axial Power Imbalance, TMI-1 3 .1 -1 Reactor Coolant System Heatup/Cooldown Limitations (Applicable to 10 EFPY) 3.1-2 Reactor Coolant System, Inservice Leak and Kydrostatic Test Limitations (Applicable to 10 EFPY) 3.1 -3 Limiting Pressure vs. Temperature Curve for 100 STD cc/ Liter H20 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, THI-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 Pump Operation after 100+10/-0 EFPD, THI-1 3.5-2G Rod Position Setooints for 2 Pump Operation from 0 to 40+10/-0 EFPD, TMI-1 3.5-2H Rod Position Setpoints for 2 Pump Operation from 40+10/-0 to 100+10/-0 EFPD, THI-1 3.5-2I Rod Position Setpoints for 2 Pump Operation after 100+10/-0 EFPD, TMI-1 3.5-2J Axial Power Inbalance Envelope for Operation from 0 to 40+10/-0 EFPD, THI-1 vii Amendment Nos. 11, 17, 29, 39, 45, 50, 59, 72, 106, 109, 120, 126, 134

m O

  • LIST OF FIGURES Figure Titl e 3.5-2K Axial Power Idalance Envelope for Operation from 40+10/-0 to 100+10/-0 EFPD, TMI-l 3.5-2L Axial Power Idalance Envelope for Operation after 100+10/-0 EFPD, TMI-l
3. 5-2M LOCA Limited Maximum Allowable Linear Heat Rate l 3.5-1 Incore Instrumentation Specification Axial Imbalance Indication, TMI-l 3.5-2 Incore Instrumentation Specification Radial Flux Tilt Indication,T MI-l 3.5-3 Incore Instrumentation Specification 3.11-1 Transfer Path to and from Cask Loading Pit 4.17-1 Snubber Functional Test - Sample Plan 2 5-1 Extended Plot Plan TMI 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-l Onsite Organization viii Amendment Nos. 72, 77, 126

1.6 POWER DISTRIBUTION 1.6.1 QULDRANT POWER TILT 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 _

The quadrant tilt limits are stated in Specification 3.5.2.4.

1.6.2 AXIAL POWER IMBALANCE l Axial power imbalance 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. Inbalance is monitored continuously by the RPS using input from the power range channels. Imbalance limits are defined in Specification 2.1 and imbalance setpoints are defined in Specification 2.3.

1.7 CONTAINMENT INTEGRITY Containment 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 each 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 modification.

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 I

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 sprcy system riser.

Amendment No. 36 1-5 (1-7-78)

l l

2. SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS, REACTOR CORE Applicability Applies to reactor thermal power, axial power imbalance, reactor l coolant system pressure, ccolant temperature, and coolant flow during power operation of the plant.

Objective To maintain the integrity of the fuel cladding.

Specification 2.1.1 The combination of the reactor system pressure and coolant temperature shall not exceed the safety limit as defined by $

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

2.1.2 The combination 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 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-imbalance 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 overheating of the cladding under normal operating conditions. This is accomplished 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 temperature. 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 temperature and the possibility of cladding failure. Although DNB is not an observable parameter during reactor operation, the observable parameters of neutron power, reactor coolant. flow, temperature, and pressure can be related to DNB through the use of a critical heat flux (CHF) correlation. The B&W-2(1) and BWC(2) correlations have been developed to predict DNB and the location of DNB for axially uniform and non-uniform heat flux distributions. The B&W-2 correlation applies to Mark-9 fuel and the BWC correlation applies to Mark BZ fuel. The local DNB ratio (DNBR), defined as the ratio of the heat flux that would cause DNB at a particular core location to the actual heat flux, is indicative of the margin to DNB. The minimum value of the DNBR, during steady-state operation, normal operational Amendment No.17 (5-18-76) 2-1

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 been considered in determining the core protection safety limits.

l The curve presented in Figure 2.1-1 represents the conditions at which the minimum allowable DNBR or greater is predicted for the limiting combination of thermal power and number of operating reactor 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 associated thermal-hydraulic situation has been established for the hot channel, then all other corrbinations of axial flux shapes and their accompanying radials must result in a condition which will not riolate 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:

a. The DNBR ligit produced by a nuclear power peaking factor of P 2.82 of the combination of the radial peak, axial peak,dan=d position of the axial peak that yields no less 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 imbalance produced by the power peaking.

2-2 Amndmnt No. 17, 50, 90, 126

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

The curve of Figure 2.1-1 is the most 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 number of reactor coolant pumps in operation or the local quality at the point of minimum DNBR is equal to 22 percent, (B&W-2)(4), or 26 percent (BWC)(2) whichever 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 instrumentation error. The maximum thermal power for other reactor coolant pump conditions is produced in a similar manner.

Using a local quality limit of 22 percent (B&W-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 quality 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 minimum DNBR less than 22 percent (B&W-2), or 26 percent (BWC) for the particular reactor coolant pump situation. Curve 1 is more restrictive than any other reactor coolant pump situation because any pressure / temperature point above and to the lef t of this curve will be above and to the left of the other curves.

REFERENCES i

(1) FSAR, Section 3.2.3.1.1 l (2) BWC Correlation of Critical Heat Flux, BAW-10143P-A, Babcock & hilcox, Lynchburg, Virginia, April 1985 (3) FSAR, Section 3.2.3.1.1.3 l

(4) FSAR, Section 3.2.3.1.1.11 I 2-3 Amendment No. 17, 29, 39, 50, 120, 126 l __ __ _ ._

2400 .

2200

,, ACCEPTABLE

OPERATION a.

J

$ 2000 E

UNACCEPTABLE 3

a OPERATION g 1800 7 v

1600 580 600 620 640 660 Reactor Outlet Temperature, OF CORE PROTECTION SAFETY LIf41T Tf41-1 Figure 2.1-1 Amendment No. 50 (3-1G-79)

Thennal Power Level, %

120

(-43.8.112) 1 (37.8,112)

ACCEPTABLE 4 PUMP OPERATION -

- 100

(-43.8,89.3) 3 (37.0,89.3)

(-58.5,80.4) AC EPT

-- 80 (53.0,80.4)

OPERATION

(-43.8,62.0) 3 (37.8g2.0)

(-58.5,57.8) ACCEPTABLE 60 (53.0,57.8) 2,3, & 4 PUMP OPER,tTION

-- 40

(-58.5,30.4) (53.0,30.4)

- 20 l t t I t I f f I I I I ' I f

.3 60 40 20 -10 0 10 20 30 40 50 60 70 80 Axial Po ..' Imbalance. *;

Curve Reactor Coolant Flow (lb/hr) 1 139.8 x 10 6 2 ;J4.5 x 10 6 3 68.8 x 10 0 CORE PROTECTION SAFETY LIMITS TMI-1 l Amendment No. 11, 2S, 33 AE, Ep,JZP,126 F,gure 2.1 ,c

-i., .,

2400

~

2200 1 g

~

h7 g --

E -

2 4

u _

y 2000 E

c. -

"v s

8 E J S 1800 ff F

1600 ^

580 600 620 640 660 U

Reactor Outlet Temperat tre, F l Reactor Coolant Flow l Curve (Ibs/hr) Power PumpsOperating(TypeofLimit) l 1 139.8 x 106(100%)* 112% FourPumps(DNBRLimit) 2 104. 5 x 106(74.7%) 89.4% Three Pumos (Quality Limit) 3 68.8 x 100 (49.2%) 62.0% OnePumpinEachLoop(QualityLimit)

  • 106.5% of Cycle 1 Design Flow I CORE PROTECTION SAFFTY BASES TMI-1 Amendment flo. 50, 126 7

2.3 LIMITING SAFETY SYSTEM SETTINGS, PROTECTION INSTRUMENTATION Applicability Applies to instruments monitoring reactor power, axial power l imbalance, reactor coolant system pressure, reactor coolant outlet temperature, flow, number of pumps in operation, and high reactor building pressure.

Objective To provide automatic protection action to prevent any combination of process 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.

Bases The r :ctor protection system consists of four instrument channels to monitor each of several selected plant conditions which will cause a reactor trip if any one of these conditions deviates from a pre-selected operating range to the degree that a safety limit may be reachad.

The trip setting limits for protection system instr'Jmentation 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 instrumentation trip set points plus calibration and instrumentation errors.

Nucigar Overpower A reactor trip at high power level (neutron flux) is provided to prevent danage to the fuel c! adding from reactivity excursions too rapid to be detected by pressure and temperature measurements.

During nornal plant operations with all reactor coolant pumps i operating, reactor trip is initiated when the reactor power level reaches 105.1% of rated power. Adding to this the possible variation in trip set points due to calibration and instrument errors, the maximum actual power at which a trip would be actuated could be 112%, which is the value used in the safety analysis (1).

l l

l 2-5 Amendment No. 13, 17, 28, 126

- - _ =. .

a. Overpower trip based on flow and imbalance The power level trip set point produced by the re actor coolant system flow is based on a power-to-flow ratio which has been established to accommodate the most sevtre thermal transient considered in the design, the loss-of-coolant flow accident from high power. Analysis has demonstrated that the specified power to flow ratio is adequate to prevent a DNBR 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 p wer level trip set point produced by the power to flow ratio provides overpower DNB protection for all modes of pump operation. For every flow rate there is a maximum permissible power level, and for every power level there is a minimum ptirmissible low flow rate. Typical power level and low flow rate combinations for the pump 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 power 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 pumps operating) if the power is 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 maxinum 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 vcat valve surveillance program during each refueling outage.

For safety analysis calculations the maximum calibration and instrae3ntation errers for the power level were used.

The pcwe.r-imbalance boundaries ar" established in order to prevent reactor thermal limits from being exceeded. These thermal limits are either nower peaking Kw/ft limits or DNBR limits. The axial l power imbalance (power in the top half of the core minus power in 2-6 Amendnent No. 13, 17, 25, 28, 39, 50, 125

the bottom half of core) reduces the power level trip produced by 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-imbalance 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 number 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 limit 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 1cwering of the high pressure trip setpoint and raising of the setpoint for the Power Operated Relief Valve (PORV), from 2255 psig to 2450 psig, has the effect of reducing the challenge rate to the PORY while maintaining ASME Code Safety Valve capability.

A B&W analysis completed in September 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 B&W generic ECCS analysis, however, assumed a low l pressure trip of 1900 psig and, to establish conformity with thir analysis, the low pressure trip setpoint has been raised 4

to

  • 9 more conservative 1900 psig. Aoplication of the B&W 2-7 Amendment No. 17, 28, 39, 45, 78, 126, 135

crossflow model resulted in safety 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 temperatore 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 calibated 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 ai; a value no higher than 620*F even under worst case conditions.

The safety analysis used a high temperature 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, linearity, repeatability, etc. This does not imply that the equipment is restricted to operation within the calibrated range. Additional testing has demonstrated that in fact, the temperature channel is fully operational approximately 10% sbove the calibrated range.

Since it has been established that the channel will trip et a value of RC outlet temperature no higher than 620*F even in the worst case, and since the channel is fully operational approximately 10% above the calibrated range and exhibits no hysteresis or foldover characteristics, it is concluded that the instrument design is acceptable,

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

2-8 Amendment No. 45, 78, 135

f. Shutdown bypass 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 huh reactor coolant system pressure trip set point of 1720 psig is automatically imposed.

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 pressur9 trip set point so that the reactor must 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 remove 5.0 percent of rated power if none of the reactor coolant pumps were operating.

keferences i (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.

l (7) "ECCS Analysis of B&W'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 i Pressure," BAW-1890, Rev. O, Babcock and Wilcox, September 1985.

l l

?-9 i

Amendment W. 13, 17, 28, 39, 45, 78, 90, 126, 135 t

1 Table 2.3 -1 .

REACTOR PROTECTION SYSTEM TRIP SETTING LIMITS (5) l .

Y Four Reactor Coolant Three Reactor Coolant One Reactor Coolant i Pumps Operating Pumps Operating Pump Operating in (Nominal Operating (Nominal Operating Each Loop (Nominal Shutdown Power - 100%) Power - 75%) Operating Power 49%) Bypass

1. Nuclear power, max. 105.1 105.1 105.1 5.0(2) I

% of rated power

2. Nuclear power based on 1.08 times flow 1.08 times flow 1.08 times flow minus Bypassed g flow (1) and imbalance minus reduction due minus reduction due reduction due to max. of rated power to inbalance to imbalance imbalance
3. Nuclear power based NA NA 55% Bypassed l (4) on pump monitors, max. % of rated power
4. High reactor coolant 2355 2355 2355 1720(3) l system pressure, p;ig max.

'?

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

6. Reactor coolant temp. 61 8.8 61 8.8 618.8 61 8.8 F., max.

. 7. High Reactor Building 4 4 4 4 pressure, psig nax.

(1 ) Reactor coolant system flow, %. l (2) Administratively controlled reduction set only during reactor shutdown. [

(3) Automatically set when other segments of the RPS (as specified) are bypassed. g (4) The pury monitors also produce a trip on: (a) loss of two reactor coolant pumps in one reactor coolant loop, l and (b) loss of one or two reactor coolant pumps durir.g two-pump operation.

(5) Trip settings limits are setting limits on the setpoint side of the protection system bistable connectors. 1

2500 P = 2355 psig .

2300 - ,

en T = 618.8 F

ACCEPTABLE a OPERATION 2100 -

O t

a.

{ -

P = 1900 psig 3 1900 u UNACCEPTABLE b OPERATION E

1700 -

1500 i i e i 540 560 580 600 620 640 Reactor Outlet Temperature. OF l

l l

TMI-1 l PROTECTION SYSTEM MAXIMUM ALLCWABLE 3ETPOINTS l

l Amendment No. 13, 17, 28, 39. H . Figure 2.3-1 78, 126, 135 1

i

Thermal Power Level, %

- 120

(-30.0.108) __ (24.5,108)

ACCEPTABLE l m3 = 1.900 l 4 pugp - 100 l

m2 = -1.854 l OPERATION l l

1(-30.0,80.6 ) (24.5,80.6 )

ACCEPTABLE 80

(-50.0,70.0) P

-- (45.0,70.0)

OP R T f l ..

(-30.0,53.1 ) 6f24.5,53.1)

(-50.0,42.6 )

l ACCEPTABLE -

2,3, & 4 l (45.0,42.6 )

l 40 PUMP - l l OPERATION l l --

l e o l e l

(-50.0,15.1 ) l 20 9 g l N (45.0.15.1 )

n n

! -- n l n e i e

l i i e i rii l i i e i i i I

70 50 30 10 0 10 20 30 40 50 60 70 80 Axial Power Imbalance, %

TR0TECTION SYSTEM MAXIMUM ALLC',lABLE SETPOINTS FOR AXIAL POWER IMBALANCE l TMI-1 Amendment No. .U. 79, 39, 70, #5, Figure 2.3-2 50, 120, 126  :

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 trirmned to the same position. Normal ope:ation of 100 percent of the thermal power allowable for the reactor coolant pump combination may then continue provided that the rod that was declared inoperable is maintained within allowable group average position limits in 3.5.2.5.

3.5.2.3 The worth of single incerted 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 4

exceed the values in Table 3.5-1 A as determined using the full incore detector system, l

b. When the full incore detector system is not available and except for physics tests quadrant tilt shall not exceed the values in Table 3.5-1 A 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 3.5-1 A as determined using the minimum incore 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 pmp operation, thermal power shall be reduced 2 percent of the thermal power allowable for the reactor coolant punp combination for each 1 percent tilt in excess of the tilt limit.

e. Within a period of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, the quadrant power tilt shall be reduced to less than the tilt ifmit except for physics tests, or the following adjustments in setpoints and limits shall be made:

3-34 Amendment No. 17, 29, 39, 40, 50, 90, 126

. o.

I

1. The protection system reactor )ower/inbalance envelope trip setpoints shall >e reduced 2 percent in power for each I 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 th .n 60% full power.
2. The control rod group withorawal limits (Figures 3.5-2A to 3.5-21) shall be reduced 2 percent in power for each 1 percent tilt in excess of the tilt limit.
3. The operational imbalance limits (Figures 3.5-2J, 3.5-2K, and 3.5-2L) shall be reduced 2 percent in l power for each 1 nercent tilt in excess of the tilt limit.
f. Except for physics or diagnostic testing, i' quadrant tilt is in excess of +16.80% determined usir;g the full incore detector system (FIT), or +14.2% determined using the out of core detector system (OCT) 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 teJting during power operation ff th a quadrant tilt is permitted provided that the thermal power allowable is restricted as stated its 3.5.2.4.d above.
g. Quadrant tilt shall be monitored on a minimum frequen:y of once every two hours during power operation above 15 percent of rated power.

Table 3.5-1 A - Quadrant Tilt Limits Tilt Limit Tilt Limit (indicated power (indicated power s 50%) > 50%)

Quadrant Tilt as Indicated By:

Full incore detector 6.83% 4.12%

system Power range channels 4.05% 1.96%

Minimum incore 2.80% 1.90%

l detector system i

3-34a Amendment No. 29, 38, 39, 40, 45, 50, 120, 126

3.5.2.5 Control Rod Positions

a. Operating rod group overlap shall not ucNd 25 percent +5 percent, between two sequential groups except for phytics 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 Figares 3.5-2D, 3.5-? E, and 3.5-2F for three pump operation. Two pump  ;

operation is specified on Figu.es 3.5-2G, 3.5-2H, and i 3.5-21. If any of these control rod position limits are exceeded, corrective measures shall be taken immediately to achieve an acceptable control rod position. Acce'ptable control rod positions shall be attained within four hours. ,

c. Deleted ,
d. Axial power imbalance 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, 3.5-2K, and 3.5-2L. If the imbalance is not within l the envelopes defined by Figures 3.5-2J, 3.5-2K, or 3.5-2L at the appropriate time in cycle, corrective l measures Shall be taken to achieve an accept?ble imbalance. If an acceptable imbalance is not achieved

within four hours, reactor power shall be reduced until imbalance limits are met,
e. Safety rod limits are given in 3.1.3.5.

l 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 l effective full power days using the incore instruentation '

detection system to verify the power distribution is within the limits shown in Figure 3.5-2H. l 3-35

, Amndent No. 10, 17, 29, 38, 39, 50, 120, 126

. _ _ _ . _ . _ _ _ ._ _ _ _ . . _ ,_ _ _. _ _ ~ _ . _ _ .

Bases The axial power imbalance envelopes defined in Figures 3.5-2J, 3.5-2K, and 3.5-2L are based on LOCA analyses which have defined tne maximum linear heat rate (see Figure 3.5-2M) such that the maximum clad temperature will not exceed the Final Acceptance Criteria

, (2200*F). Operation outside of the axial power inbalance 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 '

limited by the Final Acceptance Criteria only if the control rods are at the withdrawal / insertion limits as defined by Figures 3.5-2A, 3.5-28, 3.5-2C, 3.5-2D, 3.5-2E, 3.5-2F, 3.5-2G, 3.5-2H, 3.5-21, and if quadrant tilt is at the limit. The effects of the gray APSRs are also included. Additional conservatism is introauced by application of:

a. Nuclear uncertainty fat tors
b. Thermal calibration uncertainty
c. Fuel densification effects
d. Hot rod manufacturing tolerance factors
e. Postulated fuel rod bow effects
f. Peaking limits based on initial condition for Loss of Coolant Flow transients.

l The axial power imbalance envelopes given in Figures 3.5-2J, 3.5-2K, l and 3.5-2L have been error adjusted for observability and measurement uncertainties. Therefore, the limits specified in these figures are the maximum axial power imbalance alarm setpoints for power operation.

The Rod index versus Allowable Power curves of Figures 3.5-2A, 3.5-28, 3.5-2C, 3.5-20, 3.5-2E, 3.5-?F, 3.5-2G, 3.5-2H, and 3.5-2I describe three regions. These three regions are:

1. Permissible operating Region  !

l 2. Restricted Regions i

3. Prohibited Region (Operation in this region is not allowed) l NOTE: Inadvertent operation within the Restricted Region for a period of four hours is not considered a violation of a i

l 3-35a l Amendnent No. 17, 29, 38, 39, 50, 120, 126

The 25+5 percent overlap between successive control rod groups is alloweT 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 Safet) 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 position at power is for group 7 to be partially inserted.

The rod position limits are based on the mos; limiting of the following three criteria: ECCS power peaking, shutdown margin, and potential ejected rod worth. As discussed above, compliance with the ECCS power peaking criterion is ensured by the rod position limits. The minimum available rod worth, consistent with the rod position limits, provides for achieving hot shutdown by reactor trip at any time, assuming the highest worth control rod that is withdrawn remains in the full out position (1). The rod position  ;

limits also ensure that inserted rod groups will not contain single rod worths 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 single inserted control rod worth 1.0% ak/k at beginning of life, 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 liniits given in Figures 3.5-2A, 3.5-28, 3.5-20, 3.5-20, 3.5-2E, 3.5-iF, 3.5-2G, 3.5-2H, and 3.5-21 have been error adjusted for observat flity and measurement uncertainties.

Therefore, the limits specified in these figures are the maximum rod position alarm setpoints for operation.

The pisnt computer will scan for tilt and imbalance and will satisfy the technical specification requirements. If the cog uter is out of service, than manual calculation for tilt above 15 percent power and imbalance above 40 percent power nust be performed at least every two hours until the ccmputer is returned to service.

3-36 Amendment No. 17, 29. 39, 40, 50, 126

0

  • 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 incore instrumentation with statistically cosined 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 L an actual core tilt of +7.50% which is equivalent to a +6.83% tilt $

measured with the full incore instrumentation with statistically cosined 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 Set;;0 int 0 <5%

15 50%

40 50%

50 60%

75 85%

>75 105.1%

REFERENCES (1) FSAR, Section 3.2.2.1.2 (2) FSAR, Lection 14.2.2.2 3-36a Amndent No. 39,126

(300,102) 100 - 9 *0'  ?

(275.9.102)

SHUTDOWN 90 -

MARGIN (273.5.90)

NOT ALLOWED LIMIT 80 -

(249.5,78) g 70 - RESTRICTED c.

m

$ 60 a

50 (38.5.48) (201.5.48) f 40 -

30 -

! 20 "

6 PERMIS".*BLE j (0,11.5)

~

10 '

Oi (0,,2. 6 ) , , , , , , , , , ,

0 25 50 75 100 125 150 175 200 225 250 275 300 Indicated Rod Index, % Withdrawn 0 25 50 75 100 I i 1 f i Group 7 0 25 50 75 100 l i l l 1 Group 6 0 25 50 75 100 i i l 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, 50, 126 Figure 3.5-2A

(300,102) 100 -

(92.5.102)

I (270.1.102) 90 - SHUTDOWN (266.5,90)

MARGIN NOT ALLOWED LIMIT 80 -

u (249.5,78) i 70 -

E RESTRICTED

] 50 -

e 50 -

(38.5,48) (201.5,48) g 40 -

E 30 -

b 0

(0,l'. 5) PERMISSIBLE

~

10 ' -

0,2.6) 0 a f I f I t g 3 g g  ;

O 25 50 75 100 125 150 175 200 225 250 275 300 Indicated Rod Index, % Withdrawn 0 25 50 75 100 I I I I I Group 7 0 25 50 75 100 l I l I i Group 6 0 25 50 75 100 L I e i I Group 5 R0D POSITION SETPOINTS FOR 4 PUMP OPERATION FROM 40

+10/-0 TO 100 +10/-0 EFPD TMI-1 Amendment No. ID, 17, 29, 39 AB, 50, 126 Figure 3.5-2B 1

g a 4 l

l (300,102) 100 -

I19 '6'l )

(266.5.102) 90 -

~ " (249.5,78) i e 70 -

RESTRICTED NOT ALLOWED "3 l e 60 -

o

~

, (116.5.48) (201.5,48) y 40 -

c.

30 -

fa 20 -

PERMISSIBLE 10 - (58.5.13) 0' I.O ,2. 7 ), , , , , , , , , ,

0 25 50 75 100 125 150 175 200 225 250 275 300 Indicated Rod Index, % Withdrawn i

O 25 50 75 100 l l I I I Group 7 0 25 50 75 100 I i e i Group 6 0 25 50 75 100 l f f f l Group 5 R0D POSITION SETPOINTS FOR 4 PUMP OPERATION AFTER 100

+10/ 0 EFPD TMI-1 Amendment No. 17, 29, 39, 50, 126 1

l 100 -

90 -

(300,77) 80 -

(93.2.77) (276.0,77)i b 70 -

(273.5,67)

[ NOT ALLOWED SHUTDOWN

" 60 - MARGIN 3 LIMIT (249.5,58) f O

50 -

RESTRICTED

, 40 -

g (38.5,36)

(201.5,35.5) 2 30 -

20 -

.h

] 10 , (0.8.6) PERMISSIBLE O< ,

O'l 4) , , , , , , , i ,

0 25 50 75 100 125 150 175 200 225 250 275 300 Indicated Rod Index. % Withdrawn 0 25 50 75 100 i f I f I j Group 7 l

0 25 50 75 100

I I I f f .

l Group 6

' l 0 25 50 75 100 i f f f I Group 5 R0D POSITION SETPOINTS FOR 3 PUMP OPERATION FROM 0 TO l' 40 +10/-0 EFPD TMI-1 Figure 3.5-20 Amendnent No. 17, 29, 39, 45, 50, 126

100 -

90 -

~

(300,77)

(93.2,77) (270.3,77)p NOT ALLOWED SHUTOOWN (266.5,67)

I?> 60 - MARGIN j 'IMIT (249.5,58) t 50 -

RESTRICTED H

f 40 -

3 (38.5.36) (201.5,35.5) 30 -

T>

$ 20 -

I 19 (0,8.6) PERMISSIBLE OI 0,1.4) e i I I t I i I i I ,f 0 25 50 75 100 125 150 175 200 225 250 275 300 Indicated Rod Index, % Withdrawn 0 25 50 75 100 l i I I l Group 7 0 25 50 75 100 l I I I I Group 6 0 25 50 75 100 l l l l l Group 5 I

ROD POSITION SETPOINTS FOR 3 PUMP OPERATION FROM 40

+10/-0 TO 100 +10/ 9 EFPD l

TMI-1

~

Figure.3.5-2E Amendment No. 17, 29, 39, 40, 50, I20, 126 l

1

100 -

90 -

80 '

(198.5,77) (266.5,77) h 70 -

SHUTDOWN (266.5,67) f 60 -

NOT ALLOWED OfN

{ (249.5,58)

I SO -

RESTRICTED o

'#4 40 -

116.5,36)

(201.5.35.5) f 30 -

b 20 -

g

'il 4 10 -

(58.5,9.7) PERMISSIBLE 0' p . 2. 0), , , , , , , , , ,

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 I Group 7 0 25 50 75 100 I I I I I 25 O 50 75 100 I I I I l Group 5 R0D POSITION SETPOINTS FOR 3 PUMP OPERATION AFTER 100

+10/-0 EFPD TMI-1 Figure 3.5-2F Amendment No. 17, 29, 39, #5, 50, 120, 126

100 90 -

80 -

E 70 B

60 - SHUTDOWN 3 MARGIN (300,52) f o

50 -

NOT ALLOWED LIMIT (94.5,52) (276.2,52)J (273.5.44)

. 40 -

t (249.5.38) g RESTRICTED

o. 30 -

20 -

(201.5.23) 10 0,5.7) PERMISSIBLE (0,0.3) O m i i i i i e i i i i  :

0 25 50 75 100 125 150 175 200 225 230 275 300 Indicated Rod Index, % Withdraw.1 0 25 50 75 100 t I i l l l

Group 7 0 25 50 75 100 l I I f I Group 6 0 25 50 75 100 l i I i f Group 5 ROD POSITION SETPOINTS FOR 2 PUMP OPERATION FROM 0 TO 40 +10/-0 EFPD THI-1 Figure 3.5-2G Amendment No. 27, 29, 39, 50, 90, 126

1 100 -

90 -

80 -

g 70 -

E 60 - SHUTDOWN

?u MARGIN tigr7 a gg _ (94.5,52) (270.5.52)(300,52) e NOT ALLOWED (266.5,44) 40 -

RESTRICTED (249.5,38)

$ 30 -

20 - (38.5.24) (201.5.23) 9 2~ 10 PERMISSIBLE

- (0,5. 7)

(0,0.3) '

0a e i e i e i t i i e 0 25 50 75 100 125 150 175 200 225 250 275 300 Indicated Rod Index, % Withdrawn l

0 25 50 75 100 i i e i i Group 7 0 25 50 75 100 l i I f i Group 6 0 25 50 75 100 t I f I l Group 5 R0D POSITION SETPOINTS FOR 2 PUMP OPERATION FROM 40

+10/-0 TO 100 +10/-0 EFPD THI-1 Amendment No. 29, 39, 40, #5, 50. I20, 126 Figure 3.5-2H

100 90 80 -

b x

2 70 -

?

3 60 o (200.5.52) (300.32)

, 50 -

_ (266.5.52)-

  • aHUTDOWN L NOT ALLOWED MARGIN (266.5,44) ,

E 40 LIMIT RESTRICTED 2 (249.5.38)

] 30 -

0 (116.5.24) j 20 - (201.5.23) 10 -

PERMISSIBLE (58.5,6.5) 0i i i i e i , , , , ,

0 25 50 75 100 125 150 175 200 225 250 275 300 Indicated Rod Index, % Withdrawn l i e i , ,3 Group 7 0 25 50 75 100 l l I i t 0 25 50 75 100 I I f I l Group 5 ROD POSITION SETPOINTS FOR 2 PUvPi OPERATION AFTER 100 +10/-0 EFPD TMI-1 Amendment No. 120, 126

Indicated Power, % of Rated Power

. 110

(-13.9.102) (19.7,102)

- - 100

(-14.1,92), <(19.7,92)

- - 90

(-22.0,80), . . 80 ,(25.8,80)

- 70 RESTRICTED PERMISSIBLE

- 60 RESTRICTED REGION OPERATING REGION REGION --

50

-- 40

.. 30 20

. 10 l l i f f I I i i 1

-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 3.5-23 Amendment No. 126

Indicated Power, of Rated Power

-- 110

(-19.7,102)  ; ,, (21.6.102)

(-20.8,92) ,

(21.8,92)

,, gg

(-24.9,80)< -- 80 ,(27.8,80)

.. 70 RESTRICTED PERMISSIBLE - - 60 RESTRICTED REGION OPERATING REGION REGION 50

-. 40

-. 30

- 20

-- 10 i e i i i i e _j i _ i_

-50 30 10 0 10 20 30 40 50 Indicated Axial Power Imbalance, %

AXIAL POWER IMBALANCE ENVELOPE FOR OPERATION FR0!1 40 +10/-0 TO 100 +10/-0 EFPD TMI-1 Amendment No. 126 Figure 3.5-2K i

i Indicated Power, % of Rated Power

. 110

(-22.6,102)

-- 100

(-22.8,92) ' < (22.8,92)

- . 90

(-27.8,80)< - - 80 ,(28.7,80)

- - 70 RESTRICTED PERMISSIBLE

- 60 RESTRICTED REGION OPERATING REGION REGION -- 50

.. 40

-- 30

... 20

. -- 10 l i I I I t i i t 1 40 -30 10 0 10 20 30 40 50 Indicated Axial Power Imbalance, %

AXIALP0h!ERIMBALANCE ENVELOPE FOR OPERATION AFTER 100 +10/-0 EFPD TMI-1 fvnendnent No. 126 Figure 3.5-2L l

20 C

k a

18 -

3 a

s

~

k 16 -

/

f,#*s' a

a y

/ //

  • b ,

I 14 - / '

S E

h

2 12 0-1000 mwd /mtU 1000-2600 mwd /mtU After 2600 t1Wd/mtU 10 I i e i O 2 4 6 8 10 , 12 Axial Location From Bottom of Core, ft.

I L

f I

LOCA LIMITED MAXIMUM

! ALLOWABLE LINEAR HEAT RATE i

TM1-1 Figure 3.5-2M I

5.3 REACTOR Applicability Applies to the design features of the reactor core and reactor coolant system.

Objective To define the significant design features cf the reactor core and reactor coolant system.

Specification 5.3.1 . REACTOR CORE 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 gnemblies. Each fuel assembly contains 208 fuel rods.tlitzt 5.3.1.2 The reactor core shall approximate a right circular cylinder an activewith an of height equivalent 142 inches. diager of 128.9 inches and 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 enrichment is less than 3.7 weight percent U235, l 5.3.1.4 There are 61 full-length control rod asse211es (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 silygr-indium-cadmium alloy clad with stainless steel.L31 The gray APSRA contain a 63 inch length of Inconel.

5.3.1.5 The core will have 68 burnable poison spider assemblies 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 assemblies 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 COOLANT SYSTEM 5.3.2.1 The reactor coolant system shall be designed and constructed in accordance with code requirements.(4) i 5.3.2.2 The reactor coolant system and any connected auxiliary systems exposed to the reactor coolant conditions of temperature and pressure, shall be designed for a pressure of 2,500 psig and a temperature of 650 F. The pressurizer and pressurizer qurge line shall be designed for a tempera-ture of 670 F.(5; 5-4 Amendment No.126

_ . _ _ _ _ _ _ _ _ ~ _ _

-

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