Text

Amend 15 to License NPF-30,modifying Tech Specs to Support Transition from Westinghouse 17X17 low-parasitic Fuel Assembly Fueled Core to Westinghouse 17X17 Optimized Fuel Assembly Fueled Core
	ML20155F323
Person / Time
Site:	Callaway
Issue date:	04/08/1986
From:	Youngblood B; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20155F327	List: ML20155F323; ML20155F334;
References
	NUDOCS 8604210468
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k UNITED STATES 8

NUC' LEAR REGULATORY COMMISSION o

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WASHINGTON,0. C. 20555

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UNION ELECTRIC COMPANY CALLAWAY PLANT, UNIT 1 DOCKET NO.

50-483 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 15 License No. NPF-30 1.

The Nuclear Regulatory Commission (the Comission) has found that:

A.

The application for amendment to the Callaway Plant, Unit 1 (the facility) Facility Operating License No. NPF-30 filed by Union Electric Company (the licensee) dated November 15, 1985, as

-supplemented December 13, 1985, January 28, 1986, February 18, 1986, February 24, 1986, and Februiry 28, 1986, complies with the standards and requirements.of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulatjons set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application,

~

the provisions of the Act, and the rules and regulations of the Commission;

~~

C.

There is reaconable 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 comp 11ance with the Comission's ' regulations; D.

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

The issuance of this amendment is in accordance with 10 CFR Part~51 of the Comission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment and Paragraph 2.C.(2) of Facility Operating License No. NFP-30 is hereby amended to read as follows:

8604210468 860408 PDR ADOCK 05000403 P

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Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.15, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into the license.

UE shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION fd !?L f)'h

/,1 B. J. Youngblood, Director PWR Project Directorate #4 Division of PWR Licensing-A, NRR

Attachment:

Changes to the Technical Specifications Date of Issuance: April 8, 1986 e

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g.

ATTACHMENT TO LICENSE AMENDMENT NO. 15 OPERATING LICENSE NO. NPF-30 DOCKET N0. 50-483 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of r.hange. Corresponding overleaf pages are provided to maintain document completeness.

Amended Page Overleaf Page 4-1U I

II 1-2 1-1 1-3 1-4 1-5 1-6 1-7 1-8 1-9 2-2 2-1 2-4 2-3 2-7 B2-1 82-2 B2-3 B2-4 3/4 1-19 3/4 1-20 3/4 1-22 3/4 1-21 3/4 2-8 3/4 2-7 3/4 2-9 3/4 2-10 3/4 10-1 3/4 2-11 3/4 2-12 3/4 2-13 3/4 2-14 3/4 10-2 B3/4 2-1 B3/4 2-2 B3/4 2-3 B3/4 2-4 B3/4 2-5 B3/4 4-1 B3/4 4-2 2-8 2-9 Delete page 3/4 2-15.

Delete page B3/4 2-6.,

s 4

~

INDEX DEFINITIONS SECTION PAGE 1.0 DEFINITIONS 1.1 ACTI0N.............................................................

1-1 1.2 ACTUATION LOGIC TEST........................,......................

1-1 1.3 ANALOG CHANNEL OPERATIONAL TEST....................................

1-1 1.4 AXIAL FLUX DIFFERENCE..............................................

1-1

1. 5 CHANNEL CALIBRATION................................................

1-1 1.6 CHANNEL CHECK......................................................

1-1 1.7 CONTAINMENT INTEGRITY..............................................

1-2

1. 8 CONTROLLED LEAKAGE.................................................

1-2

- 1.9 CORE ALTERATION....................................................

1-2 1.10 DESIGN THERMAL P0WER...............................................

1-2 1.11 DOSE EQUIVALENT I-131..............................................

1-2 1.12 -E-AVERAGE DISINTEGRATION ENERGY....................................

1-3 1.13 ENGINEERED SAFETY FEATURES RESPONSE TIME...........................

1-3 1.14 FREQUENCY N0TATION.................................................

1-3 1.15. I D ENTI FI ED L EA KAG E.................................................

1-3 1.16 MASTER RELAY TEST..................................................

1-3 1.17 MEMBER (S) 0F THE PU8LIC............................................

1-3 1.18 0FFSITE DOSE CALCULATION MANUAL....................................

1-4 1.19 OPERABLE - OPERABILITY.............................................

1-4 1.20 OPERATIONAL MODE - MODE...........................................

1-4 1.21 PHYSICS TESTS......................................................

1-4 1.22 PRESSURE BOUNDARY LEAKAGE..........................................

1-4 1.23 PROCESS CONTROL PR0 GRAM............................................

1-4 1.24 P U R G E - P U R G I NG....................................................

1-4 1.25 QUADRANT POWER TILT RATI0..........................................

1-5 1.26 RATED THERMAL P0WER................................................

1-5 1.27 REACTOR TRIP SYSTEM RESPONSE TIME..................................

1-5 1.28 REPORTABLE EVENT:..................................................

1-5 1.29 SHUTDOWN MARGIN....................................................

1-5 CALLAWAY - UNIT 1 I

Amendment No. 15 i

l l

l

- s.

F INDEX DEFINITIONS SECTION PAGE DEFINITIONS (Continued) 1.30 SITE B0VNDARY.....................................................

1-5 1.31 S LAVE R E LAY T E ST..................................................

1-5 1.32 SOLIDIFICATION....................................................

1-5 1.33 ' SOURCE CHECK......................................................

1-5 1.34 STAGGERED TEST BASIS..............................................

1-6 1.35 THERMAL P0WER.....................................................

1-6 1.36 TRIP ACTUATING DEVICE OPERATIONAL TEST............................

1-6 1.37 UNIDENTIFIED LEAKAGE..............................................

1-6 1.38. UNRESTRICTED AREA.................................................

1-6 1.39 VENTILATION EXHAUST TREATMENT SYSTEM..............................

1-6 1.40 VENTING...........................................................

1-6 1.41 WASTE GAS HOLDUP SYSTEM...........................................

1-7 TABLE 1.1 FREQUENCY N0TATION...........................................

1-8 2^

TABLE 1.2 OPERATIONAL M0 DES............................................

1-9 CALLAWAY - UNIT 1 II Amendment No. 15

1. 0 DEFINITIONS 5

4 The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications.

ACTION 1.1 ACTION shall be that part of a Technical Specification which prescribes remedial measures required under designated conditions.

ACTUATION LOGIC TEST 1.2 An ACTUATION LOGIC TEST shall be the application of various simulated input combinations in conjunction with each possible interlock logic state and verification of the required logic output. The ACTUATION LOGIC TEST shall include a continuity check, as a minimum, of output devices.

ANALOG CHANNEL OPERATIONAL TEST 1.3 An ANALOG CHANNEL OPERATIONAL TEST shall be the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY of alare, interlock and/or trip functions. The ANALOG CHANNEL OPERATIONAL TEST shall include adjustments, as necessary, of the alarm, inter-lock and/or Trip Setpoints such that the Setpoints are within the required range and accuracy.

AXIAL FLUX DIFFERENCE

.)

1.4 AXIAL FLUX DIFFERENCE shall be the difference in normalized flux signals between the top and bottom halves of a two section excore neutron detector, i

. CHANNEL CALIBRATION j

1. 5 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel such that it responds within the required range and accuracy to known values of input. The CHANNEL CALIBRATION shall encompass the entire channel including the sensors and alarm, interlock and/or trip functions and may be performed by any series of sequential, overlapping, or total channel steps such that the entire channel is calibrated.

CHANNEL CHECK 1.6 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.

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g CALLAWAY - UNIT 1 1-1 I

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c DEFINITIONS a

' CONTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when:

a.

All penetrations required to be closed during accident conditions are either:

1)

Capable of being closed by an OPERABLE containment automatic isolation valve system, or 2)

Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except as provided in Table 3.6-1 of Specification 3.6.3.

b.

All equipment hatches are closed and sealed, c.

Each air lock is in compliance with the requirements of Specification 3.6.1.3, d.

The containment leakage rates are within the limits of Specification 3.6.1.2, and e.

The sealing mechanism associated with each penetration (e.g., welds, bellows, or 0-rings) is OPERABLE.

CONTROLLED LEAKAGE

.. ~

' 1. 8 CONTROLLED LEAKAGE shall be that seal water flow from the reactor coolant pump seals.

CORE ALTERATION 1.9 CORE ALTERATION shall be the raovement or manipulation of any component within the reactor vessel with the vessel head removed and fuel in the vessel.

Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position.

DESIGN THERMAL POWER 1.10 DESIGN THERMAL POWER shall be a design total reactor core heat transfer rate to the reactor coolant of 3565 MWt.

DOSE EQUIVALENT I-131 1.11 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie / gram) l which alone would produce the same thyroid dose as the quantity atd isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present.

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

CALLAWAY - UNIT 1 1-2 Amendment No. 15 I

DEFINITIONS I - AVERAGE DISINTEGRATION ENERGY 1.12 I shall be the average (weighted in proportion to the concentration of l

each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half-lives greater than 15 minutes, making up at least 95% of the total noniodine activity in the coolant.

ENGINEERED SAFETY FEATURES RESPONSE TIME 1.13 The ENGINEERED SAFETY FEATURES (ESF) RESPONSE TIME shall be that time l

interval from when the monitored parameter exceeds its ESF Actuation Setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable.

FREQUENCY NOTATION 1.14 The FREQUENCY NOTATION specified for the performance of Surveillance l

Requirements shall correspond to the intervals defined in Table 1.1.

IDENTIFIED LEAKAGE 1.15 IDENTIFIED LEAKAGE shall be:

l

~

a.

Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or b.

Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of Leakage Detection Systems or not to be PRESSURE BOUNDARY LEAKAGE, or c.

Reactor Coolant System leakage through a steam generator to the Secondary Coolant System.

MASTER RELAY TEST 1.16 A MASTER RELAY TEST shall be the energization of each master relay and l

verification of OPERABILITY of each relay.

The MASTER RELAY TEST shall include a continuity check of each associated slave relay.

~

MEMBER (S) 0F THE PUBLIC 1.17 MEMBER (S) 0F THE PUBLIC shall include all persons who are not occupa-l tionally associated with the plant.

This category does not include employees of the licensee, its contractors or vendors.

Also excluded from this category are persons who enter the site to service equipment or to make deliveries.

This category does include persons who use portions of the site for recrea-tional, occupational, or other purposes not associated with the plant.

CALLAWAY - UNIT 1 1-3 Amendment No. 15 9

DEFINITIONS z-OFFSITE' DOSE CALCULATION MANUAL-1.,.

1.18 The OFFSITE DOSE CALCULATION MANUAL (00CM) shall contain the methodology l

and parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm / Trip Setpoints, and in the conduct of the Environmental Radiological Monitoring Program.

OPERA 8LE - OPERABILITY 1.19 A system, subsystem, train, component or device shall be OPERABLE or l

-have OPERA 8ILITY when it is capable of performing its specified function (s),

and when all necessary attendant instrumentation, controls, electrical power, 1

cooling or seal water, lubrication or other auxiliary equipment that are required for the system,. subsystem, train, component, or device to perform ~its function (s) are also capable of performing their related support function (s).

OPERATIONAL MODE - MODE 1.20 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive l

combination of core reactivity condition, power level, and average reactor coolant temperature specified in Table 1.2.

PHYSICS TESTS 1.21 PHYSICS TESTS shall be those tests performed to measure the fundamental l

nuclear characteristics of the core and related instrumentation:

(1) described

,in Chapter.14.0 of the FSAR, or (2) authorized under the provisions of

10 CFR 50.59, or (3) otherwise approved by the Commission.

PRESSURE BOUNDARY LEAKAGE 1.22 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube l

leakage) through a nonisolable fault in a Reactor Coolant System component body, pipe wall, or vessel wall.

-PROCESS CONTROL PROGRAM 1.23 The PROCESS CONTROL PROGRAM shall contain the current formula, sampling, l

. analyses, tests, and determinations to be made to ensure that the processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Part 20, 10 CFR Part 71 and Federal and State regulations, burial ground requirements, and other requirements governing the dispos,a1 of the radioactive waste.

PURGE - PURGING 1.24 PURGE or PURGING shall be any controlled process of discharging air or l

gas from a confinement to maintain temperature, pressure, humidity, concentra-tion or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

I CALLAWAY - UNIT 1 1-4 Amendment No.15

DEFINITIONS QUADRANT POWER TILT RATIO 1.25 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore l

detector calibrated output to the average of the upper excore detector cali-brated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater. With one excore detector inoperable, the remaining three detectors shall be used for computing the average.

RATED THERMAL POWER 1.26 RATED THERMAL POWER shall be a total core heat transfer rate to the l

reactor coolant of 3411 MWt.

REACTOR TRIP SYSTEM RESPONSE TIME 1.27 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from

]

when the monitored parameter exceeds its Trip Setpoint at the channel sensor until loss of stationary gripper coil voltage.

REPORTABLE EVENT 1.28 A REPORTABLE EVENT shall be any of those conditions specified in l

Section 50.73 to 10 CFR Part 50.

SHUTDOWN MARGIN 1

l.29 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which l

,the reactor is subcritical or would be subcritical from its present condition assuming all full-length rod cluster assemblies (shutdown and control) are fully inserted excapt for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn.

SITE BOUNDARY 1.30 The SITE BOUNDARY shall be that line beyond which the land is neither l

owned, nor leased, nor otherwise controlled by the licensee.

SLAVE RELAY TEST 1.31 A SLAVE RELAY TEST shall be the energization of each slave relay and l

verification of OPERABILITY of each relay.

The S*. AVE RELAY TEST shall include a continuity check, as a minimum, of associated testable actuation devices.

SOLIDIFICA_ TION 1.32 SOLIDIFICATION shall be the conversion of wet wastes into a form that l

meets shipping and burial ground requirements.

SOURCE CHECK 1.33 A SOURCE CHECK shall be the qualitative assessment of channel response l

when the channel sensor is exposed to a source of increased radioactivity.

CALLAWAY - UNIT 1 1-5 Amendment No.15

DEFINITIONS STAGGERED TEST BASIS l

1.34 A STAGGERED TEST BASIS shall consist of:

I a.

A. test schedule for n systems, subsystems, trains, or other designated components obtained by dividing the specified test interval into n equal subintervals, cad b.

The testing of one system, subsystem, train, or other designated component at the beginning of each subinterval.

THERMAL POWER 1.35 THERMAL POWER shall be the total core heat transfer rate to the reactor l

coolant.

TRIP ACTUATING DEVICE OPERATIONAL TEST 1.36 A TRIP ACTUATING DEVICE OPERATIONAL TEST shall consist of operating the l

Trip Actuating Device and verifying OPERABILITY of alarm, interlock and/or trip functions. The TRIP ACTUATING DEVICE OPERATIONAL TEST shall include adjustment, as necessary, of the Trip Actuating Device such that it actuates at the required Setpoint within the required accuracy.

UNIDENTIFIED LEAKAGE 1.37 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE l

or CONTROLLED LEAKAGE.

~~

UNRESTRICTED AREA

'1.38 An UNRESTRICTED AREA shall be any area at or beyond the SITE B0UNDARY l

access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials, or any area within the SITE BOUNDARY used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes.

VENTILATION EXHAUST TREATMENT SYSTEM 1.39 A VENTILATION EXHAUST TREATMENT SYSTEM shall be any system designed and l

installed to reduce gaseous radiciodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters.for the purpose of removing iodines or partic-ulates from the gaseous exhaust stream prior to the release to the environment.

Such a system is not considered to have any effect on noble gas effluents.

Engineered Safety Features (ESF) Atmospheric Cleanup Systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

VENTING 1.40 VENTING shall be any controlled process of discharging air or gas from a l

confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is not pro-vided or required during VENTING.

Vent, used in system names, does not imply a VENTING process.

CALLAWAY - UNIT 1 1-6 Amendment No.

15

o DEFINITIONS WASTE GAS HOLDUP SYSTEM o.

p.

1.41 A WASTE GAS HOLDUP SYSTEM shall be any system designed and installed to l

j reduce radioactive gaseous effluents by collecting Reactor Coolant System off-gases from the Reactor Coolant System and providing for delay or holdup for the

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purpose of reducing the total radioactivity prior to release to the environment.

l-t i

et CALLAWAY - UNIT 1 1-7 Amendment No.15

TABLE 1.1 FREQUENCY NOTATION

):.

NOTATION FREQUENCY S

At.least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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

W At least once per 7 days.

M At least once per 31 days.

Q At least once per 92 days.

SA At least once per 184 days.

R At least once per 18 months.

S/U Prior to each reactor startup.

N.A.

Not applicable.

P Completed prior to each release.

f

O O

O CALLAWAY - UNIT 1 1-8 Amendment No. 15 l

TABLE 1.2 OPERATIONAL MODES REACTIVITY

% RATED AVERAGE COOLANT MODE CONDITION, K,7f THERMAL POWER

TEMPERATURE 1.

POWER OPERATION

> 0.99

> 5%

> 350*F 2.'

STARTUP

> 0.99

< 5%

> 350*F 3.

HOT STANDBY

< 0.99 0

> 350*F 4.

HOT SHUTDOWN

< 0.99 0

350*F > T

> 200*F avg 5.

COLD SHUTDOWN

< 0.99 0

< 200*F 6.

REFUELING **

5 0.95 0

$ 140*F

" Excluding decay heat.

- Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.

O l

CALLAWAY - UNIT 1 1-9 Amendment No. 15

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temperature (T,yg) shall not exceed the limits shown in Figure 2.1-1 for four loop operation.

APPLICABILITY: MODES 1 and 2.

ACTION:

Whenever the point defined by the combination of the highest operating loop average temperature and THERMAL POWER has exceeded the appropriate pres-surizer pressure line, be in HOT STANOBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and comply with the requirements of Specification 6.7.1.

REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig.

APPLICABILITY: MODES 1, 2,' 3, 4, and 5.

ACTION:

MODES 1 and 2:

Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STAND 8Y with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and comply with the requirements of Specification 6.7.1.

~~

MODES 3, 4, and 5:

Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes, and comply with the requirements of Specification 6.7.1.

e CALLAWAY - UNIT 1 2-1

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0 0.2 0.4 0.6 0.8 1.0 1.2 FRACTION OF RATED THERMAL POWER FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION CALLAWAY - UNIT 1 2-2 Amendment No. 15

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS f.

2.7 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The Reactor Trip System Instrumentation and Interlocks Setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2-1.

APPLICABILITY: As shown for each channel in Table 3.3-1.

ACTION:

a.

With a Reactor Trip System Instrumentation or Interlock Setpoint less conservative than the value shown in the Trip Setpoint column but more conservative than the value shown in the Allowable Value column of Table 2.2-1, adjust the Setpoint consistent with the Trip Setpoint value.

b.

With the Reactor Trip System Instrumentation or Interlock Setpoint less conservative than the value shown in the Allowable Values column of Table 2.2-1, either:

1.

Adjust the Setpoint consistent with the Trip Setpoint value of Table 2.2-1 and determine within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that Equation 2.2-1 was satisfied for the affected channel, or 2.

Declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1 until the channel is restored to OPERABLE status with its Setpoint adjusted consistent with the Trip Setpoint value.

Equation 2.2-1 Z + R + 5 < TA Where:

Z=

The value from Column Z of Table 2.2-1 for the affected channel,

~

R=

The "as measured" value (in percent span)'of rack error for the affected channel, S=

Either the "as measured" value (in percent span) of the sensor error, or the value from Column 5 (Sensor Error) of Table 2.2-1 for the affected channel, and TA = The value from Column TA (Total Allowance) of Table 2.2-1 for the affected channel.

w CALLAWAY - UNIT 1 2-3

5

.e-TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS TOTAL SENSOR ERROR r-

[

FUNCTIONAL UNIT ALLOWANCE (TA)

Z_

QJ TRIP SETPOINT ALLOWABLE VALUE 1.

Manual Reactor Trip N.A.

N.A.

g 2.

Power Range,' Neutron Flux q

a. High Setpoint 7.5 4.56 0

$109% of RTP*

$112.3% of RTP*

b. Low Setpoint 8.3 4.56 0

<2S% of RTP*

<28.3% of RTP*

3.

Power Range, Neutron Flux, 2.4 0.5 0

<4% of RTP* with

<6.3% of RTP* with

'High Positive Rate i time constant i time constant 22 seconds 22 seconds 4.

Power Range, Neutron Flux, 2.4 0.5 0

<4% of RTP* with

<6.3% of RTP^ with High Negative Rate i time constant i time constant

>2 seconds 22 seconds y

5.

Intermediate Range, 17.0 8.41 0

Neutron Flux

~<25% of RTP*

-<35.3% of RTP*

6.

Source Range, Neutron Flux 17.0 10.01 0

$105 cps

$1.6 x 105 cps 7.

Overtemperature AT 9.3 6.77 2.06 See Note 1 See Note 2

+1.24***

8.

Overpower AT 5.7 1.49 0.5 See Note 3 See Note 4 1

9.

Pressurizer Pressure-Low 5.0 2.21 2.0 21885 psig 21874 psig l

10.

Pressurizer Pressure-High 7.5 4.96 1.0

$2385 psig

$2400 psig g

11.

Pressurizer Water Level-High 8.0 2.18 2.0 592% of instrument 193.8% of instrument g

span span 12.

Peactor Coolant Flow-Low 2.5 1.7 0.6 190% of loop

>89.2% of loop r

minimum measured minimum measured g

flow **

RTP = RATED THERMAL POWER

- Minimum Neasured Flow = 95,660 gpm

- - Two Allowances (temperature and pressure, respectively)

TABLE 2.2-1 (Continued) n it TABLE NOTATIONS E

g NOTE 1: OVERTEMPERATURE AT

+ 1 s [T (y,1

]

aT I

1 5) 1 AT, {Kg -K 7sg) - T'] + K (P - P') - f (AI)}

1 I

l 2

3 3

3 i'i w

s Where:

AT Measured AT by RTD Manifold Instrumentation;

=

1 Lead-lag compensator on measured AT;

=

3 It, T2 Time constants utilized in lead-lag compensator for AT, i t = 8 s.

=

T2 =.3 s; 1

Lag compensator on measured AT;

=

y.

3

}

T3 Time constant utilized in the lag compensator for AT. Ta = 0 s;

=

AT, 61.8*F (Referenced AT at DESIGN THERMAL POWER);

=

1.15; K

=

0.0251/*F; K

=

2 1

f

= The function generated by the lead-lag compensator for T,yg dynamic compensation; Time constants utilized in the lead-lag compensator for T,yg, t4 = 28 s,

=

T4, Is is = 4 s; g

Average temperature, *F; T

=

I g

y.

3 Lag compensator on measured T,yg;

=

5 f

Ts Time constant utilized in the measured T,yg lag compensator, is = 0 s;

=

U,

j g

TABLE'2.2-1 (Continued)

TABLE NOTATIONS (Continued)

NOTE 1:

(Continued)

T'

$ 588.4 F (Referenced T,yg at DESIGN THERMAL POWER);

0.00116; K

=

3 Pressurizer pressure, psig; P

=

P' 2235 psig (Nominal RCS operating pressure);

=

Laplace transform operator, s 1; 5

=

and f (AI) is a function of the indicated difference between top and bottom detectors of the t

power-range neutron ion chambers; with gains to be selected based on measured instrument response during plant STARTUP tests such that:

y between -35% and + 6%, f (AI) = 0, where q and q are percent DESIGN THERMAL l

(i) For gt - 9b t

t b

POWER in the top and bottom halves of the core respectively, and q

9 is total THERMAL t

b l

q[j POWER in percent of DESIGN THERMAL POWER; (ii) For each percent that the magnitude of qt 9 exceeds -35%, the AT Trip Setpoint shall b

be automatically reduced by 1.91% of its value at DESIGN THERMAL POWER; and l

(iii) For each percent that the magnitude of q 9 exceeds +6%, the AT Trip Setpoint shall t

b be automatically reduced by 1.89% of its value at DESIGN THERMAL POWER.

k NOTE 2:

The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 2.0%

g of AT span.

5 E

F

.. ; _ z.c;

w. :.. '. ; -'

'. ; K : :... e.2 9.~ L

','?2.::.

T ';. ;* ~T _ y lll.,

f_"

" ' ' _ * ~. '

q ~;

..: a -

+

c.

TABLE 2.2-1 (Continued) n TABLE NOTATIONS (Continued) 95 NOTE 3:

OVERPOWER AT E

ATfl (1 f

3) 5 AT {K4

) T - Ks [T (y I

3) - T"] - f (AI)}

(

-K 5 1 15 1

2 y

Ta g

15 7

8

-e H

Where:

AT Measured AT by RTD Manifold Instrumentation;

=

1+rS lead-lag compensator on measured AT;

=

1+TS2 I.

T2 Time constants utilized in lead-lag compensator for AT,

=

It, = 8 s..

T2 = 3 s; 1

Lag compensator on measured AT;

=

y.

3 ie Is Time constant utilized in the lag compensator for AT, r3 = 0 s;

=

61.8*F (Referenced AT at DESIGN THERMAL POWER);

AT

=

g K

1.080;

=

4 3

0.02/ F for increasing average temperature and 0 for decreasing average K

=

temperature; 4

b y[7 The function generated by the rate-lag compensator for T,yg dynamic

=

3 compensation; y

T7 Tirre constant utilized in the rate-lag compensator for Tav9, 17= 10 s;

=

e=

1 g-y.

3 Lag compensator on measured T,yg;

=

5 Time constant utilized in the measured T lag compensator, Ts = 0 s; y

is

=

avg C

s

+ ~ u :.;

.:. w, w a :: m y c k o w.i : M.-' s -> :. +.. ;.

z. f c+.

',+.1.; y 4..,.

TABLE 2.2-1 (Continued) 9

~

{

TABLE NOTATIONS (Continued)

I NOTE 3:

(Continued) 0.0065/*F for T > T" and Ks = 0 for T 5 T";

l C5 Ks

=

w Average Temperature. *F; T

=

Indicated T,yg at DESIGN THERMAL POWER (Calibration temperature for AT T"

=

instrumentation, 5 588.4*F);

Laplace transform operator, s 1; and S

=

f (AI) 0 for all AI.

=

2 NOTE 4:

The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than m

4 4.1% of AT span.

~

o j

l E

s R

3e 1

2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission prod-ucts to the reactor coolant.

Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime could re-sult in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coef-ficient.

DNB is not a directly measurable parameter during operation and there-fore THERMAL POWER and Reactor Coolant Temperature and Pressure have been re-lated to DNB.

This relation has been developed to predict the DNB flux and the location of DNB for axially uniform and non uniform heat flux distributions.

The local DNB heat flux ratio (DNBR) defined as the ratio of the heat flux that would cause DNB at a particular core location to the lucal heat flux, is indica-tive of the margin to DNB.

The DNB design basis is as follows:

there must be at least a 95 percent probability that the minimum DNBR of the limiting rod during Condition I and II events is greater than or equal to the DNBR limit of the DNB correlation being used (the WRB-1 correlation in this application).

The correlation DNBR limit is established based on the entire applicable experimental data set such that there is a 95 percent probability with 95 percent confidence that DNB will not occur when the minimum DNBR is at the DNBR limit (1.17 for the WRB-1 Correla-tion).

In meeting this design basis, uncertainties in plant operating parameters, nuclear and thermal parameters, and fuel fabrication parameters are considered statistically such that there is at least a 95% probability with 95% confidence level that the minimum DNBR for the limiting rod is greater than or equal to the DNBR limit. The uncertainties in the above plant parameters are used to determine the plant DNBR uncertainty.

This DNBR uncertainty, combined with the correlation DNBR limit, establishes a design DNBR value which must be met in plant safety analyses using values of input parameters without uncertainties.

For Callaway, the design DNBR values are 1.32 and 1.34 for thimble and typical cells, respectively.

In addition, margin has been maintained in the design by meeting safety analysis DNBR limits of 1.42 for thimble and 1.45 for typical cells in performing safety analyses.

The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature below which the calcu-lated DNBR is no less than the design DNBR value or the average enthalpy at the vessel exit is less. than the enthalpy of saturated liquid.

CALLAWAY - UNIT 1 B 2-1 Amendment No. 15

SAFETY LIMITS BASES 2.1.1 REACTOR CORE (Continued)

The curves are based on a nuclear enthalpy rise hot channel factor, F H' of 1.49 and a reference cosine with a peak of 1.55 for axial power shape. An allowance is included for an increase in F at reduced power based on the g

expression:

F H = 1.49 [1+ 0.3 (1-P)]

where P is the fraction of RATED THERMAL POWER.

These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance is within the limits of the f (AI) function of the Overtemperature trip. When the axial power imbalance t

.is not within the tolerance, the axial power imbalance effect on the Overtem-perature aT trips will reduce the setpoints to provide protection consistent with core safety limits.

2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this safety limit protects the integrity of the Reactor Coolant System (RCS) from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.

The reactor vessel, pressurizer, and the RCS piping and valves are designed to Section III of the ASME Code for Nuclear Power Plants which permits a maximum transient pressure of 110% (2735 psig) of design pressure.

The Safety Limit of 2735 psig is therefore consistent with the design criteria and associated Code requirements.

The entire RCS is hydrotested at greater than or equal to 125% (3110 psig) of design pressure to demonstrate integrity prior to initial operation.

CALLAWAY - UNIT 1 B 2-2 Amendment No. 15

2. 2 LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS The Reactor Trip Setpoint Limits specified in Table 2.2-1 are the nominal values at which.the Reactor trips are set for each functional unit.

The Trip Setpoints have been selected to ensure that the core and Reactor Coolant System are prevented from exceeding their Safety Limits during normal operation and design basis anticipated operational occurrences and to assist the Engineered Safety Features Actuation System in mitigating the consequences of accidents.

The Setpoint for a Reacter Trip System or interlock function is considered to be adjusted consistent with the nominal value when the "as measured" Setpoint is within the band allowed for calibration accuracy.

To accommodate the instrument drift assumed to occur between operational tests and the accuracy to which Setpoints can be measured and calibrated, Allow-able Values for the Reactor Trip Setpoints have been specified in Table 2.2-1.

Operation with Setpoints less conservative than the Trip Setpoint but within the Allowable Value is acceptable since an allowance has been made in the safety analysis to accommodate this error.

An optional provision has been included for determining the OPERABILITY of a channel when its Trip Setpoint is found to exceed the Allowable Value.

The methodology of this option utilizes the "as measured" deviation from the specified calibration point for rack and sensor components in conjunction with a statistical combination of the other uncer-tainties of the instrumentation to measure the process variable and the uncer-tainties in calibrating the instrumentation.

In Equation 2.2-1, Z + R + S

'the interactive effects of the errors in the rack and the sensor, and the "$ TA, as

-measured" values of the errors are considered.

Z, as specified in Table 2.2-1, in percent span, is the statistical summation of errors assumed in the analysis excluding those associated with the sensor and rack drift and the accuracy of their measurement.

TA or Total Allowance is the difference, in percent span, between the Trip Setpoint and the value used in the analysis for Reactor trip.

R or Rack Error is the "as measured" deviation, in percent span, for the affected channel from the specified Trip Setpoint.

S or Sensor Error is either the "as measured" deviation of the sensor from its calibration point or the value specified in Table 2.2-1, in percent span, from the analysis assumptions.

For functions which have multiple input values, due to more than one parameter providing input to the function, multiple values for 5 are noted which are applicable to the primary input channels.

(See Westinghouse statistical set-point study for protection systems provided for justification).

Use of Equa-tion 2.2-1 allows for a sensor drift factor, an increased rack drift factor, and provides a threshold value for REPORTABLE EVENTS.

The methodology to derive the Trip Setpoints is based upon combining all of the uncertainties in the channels.

Inherent to the determination of the Trip Setpoints are the magnitudes of these channel uncertainties.

Sensors and other instrumentation utilized in these channels are expected to be capable of operat-ing within the allowances of these uncertainty magnitudes.

Rack drift in excess of the Allowable Value exhibits the behavior that the rack has not met its allowance.

Being that there is a small statistical chance that this will happen, an infrequent excessive drift is expected.

Rack or sensor drift, in excess of the allowance that is more than occasional, may be indicative of more serious problems and should warrant further investigation.

CALLAWAY - UNIT 1 B 2-3 Amendment No. 15

IIMilING SAFilY SY% TEM SETTINGS BASES REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS (Continued)

The various Reactor trip circuits automatically open the Reactor trip breakers whenever a condition monitored by the Reactor Trip System reaches a preset or calculated level.

In addition to redundant channels and trains, the design approach provides a Reactor Trip System which monitors numerous system variables, therefore providing Trip System functional diversity.

The func-tional capability at the specified trip setting is required for those antici-patory or diverse Reactor trips for which no direct credit was assumed in the safety analysis to enhance the overall reliability of the Reactor Trip System. The Reactor Trip System initiates a Turbine trip signal whenever

. Reactor trip is initiated. This prevents the reactivity insertion that would otherwise result from excessive Reactor Coolant System cooldown and thus avoids unnecessary actuation of the Engineered Safety Features Actuation System.

Manual Reactor Trip The Reactor Trip System includes manual Reactor trip capability.

Power Range, Neutron Flux In each of the Power Range Neutron Flux channels there are two independent bistables, each with its own trip setting used for a High and Low Range trip I

setting.

The Low Setpoint trip provides protection during subcritical and low i ;/

s

^

.ppwer operations to mitigate the consequences of a power excursion beginning from low power, and the High Setpoint trip provides protection during power operations to mitigate the consequences of a reactjvity excursion from all power levels.

a-

~

The Low Setpoint trip may be manually blocked above P-10 (a power level of approximately 10% of RATED THERMAL POWER) and is automatically reinstated below the P-10 Setpoint.

~~~

Power Range, Neutron Flux, High Rates The Power Range Positive Rate trip provides protection against rapid flux

~

increases which are characteristic of a rupture of a control rod drive housing.

Specifically, this trip complements the Power Range Neutron Flux High and Low trips to ensure that the criteria are met for rod ejection from mid power.

The Power Range Negative Rate trip provides protection for control rod drop accidents. At high power a single or multiple rod drop accident could cause local. flux peaking which could cause an unconservative local DNBR to exist. The Power Range Negative Rate trip will prevent this from occurring by tripping the reactor.

No credit is taken for operation of the Power Range Negative Rate trip for those control rod drop accidents for which DNBRs will be greater than the limit value.

CALLAWAY - UNIT 1 B 2-4

T e

REACTIVITY CONTROL SYSTEMS ROD DROP TIME

-LIMITING CONDITION FOR OPERATION 3.1.3.4 'The individual full-length shutdown and control rod drop time from the fully withdrawn position shall be less than or equal to 2.4 seconds from l

beginning of decay of stationary gripper coil voltage to dashpot entry with:

T,yg greater than or equal to 551*F, and a.

b.

All Reactor Coolant pumps operating.

APPLICABILITY:

MODES 1 and 2.

ACTION:

a.

With the rod drop time of any full-length rod determined to exceed the above limit, restore the rod drop time to within the above limit prior to proceeding to MODE 1 or 2.

b.

With the rod drop times within limits but determined with three reactor coolant pumps operating, operation may proceed provided THERMAL POWER is restricted to less than or equal to 66% of RATED THERMAL POWER.

. SURVEILLANCE REQUIREMENTS

~

4.1.3.4 The rod drop time of full-length rods shall be demonstrated through measurement prior to reactor criticality:

For all rods following each removal of the reactor vessel head, a.

b.

For specifically affected individual rods following any maintenance on or modification to the Control Rod Drive System which could affect the drop time of those specific rods, and c.

At least once per 18 months.

CALLAWAY - UNIT 1 3/4 1-19 Amendment No.

15 9

=

REACTIVITY CONTROL SYSTEMS SHUTDOWN R00 INSERTION LIMIT LIMITING CONDITION FOR OPERATION 3.1.3.5 All shutdown rods shall be fully withdrawn.

APPLICABILITY: MODES 1*'and 2*#.~

ACTION:

With a maximum of one shutdown rod not fully withdrawn, except for surveillance testing pursuant to Specification 4.1.3.1.2, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either:

a.

Fully withdraw the rod, or b.

Declare the rod to be inoperable and apply Specification 3.1.3.1.

~.

SURVEILLANCE REQUIREMENTS I

4.1.3.5 Each shutdown rod shall be determined to be fully withdrawn:

a.

Within 15 minutes prior to withdrawal of any rods in Control Bank A, B, C, or D during an approach to reactor criticality, and b.

At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereaf ter.

See Special Test Exceptions Specifications 3.10.2 and 3.10.3.

With K,ff greater than or equal to 1.

CALLAWAY - UNIT 1 3/4 1-20

REACTIVITY CONTROL SYSTEMS CONTROL ROD INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.6 The control banks shall be limited in physical insertion as shown in Figure 3.1-1.

APPLICABILITY: MODES 1* and 2*#.

ACTION:

With the control banks inserted beyond the above insertion limits, except for surveillance testing pursuant to Specification 4.1.3.1.2:

Restore the control banks to within the limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or a.

b.

Reduce THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the bank position-using the above figure, or Be in at least HOT STANDBY within 6 hpurs.

c.

SURVEILLANCE REQUIREMENTS s

'4.1.3.6 The position of each control bank shall be determined to be within

-the insertion limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the Rod Insertion Limit Monitor is inoperable, then verify the individual rod positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

"See Special lest Exceptions Specifications 3.10.2 and 3.10.3.

With K,77 greater thaa or equal to 1.

(

CALLAWAY - UNIT 1 3/4 1-21

(FULLY WITHDR AWN) -

Ste ES -

St 200 ISO 4

g 1s0 4

140

=

B a 130 2

w I 100 2

g o

s0 i

M 2:

e9 O

4 0"

40 t

to 0

0 20 40 40 80 100 (FULLY INSERTED)

RATED THERMAL POWER (Percent)

FIGURE 3.1-1 R0D BANK INSERTION LIMITS VERSUS RATED THERMAL POWER - FOUR LOOP OPERATION 1

CALLAWAY - UNIT 1 3/4 1-22 Amendment No.15

.,,.,,..---e...-------.-.-------%

--,--,.-..--e,

..+------..--.------...---~--.-..,----4.--e,.=-w

-.r

p-

,POWl:R_ D151RIRU110N LIMITS

~SilRVI.ItiANCE REQUIREMENTS (Continued)'

2)

When the I is less than or equal to the F limit. for the appropriate measurcel core plane, additional power distribution C

RP maps shall be taken and F compared to F and F, at least once per.31 EIPD.

e.

lhe i limits for RATED TliERMAL POWER (FRTP) shall be provided for xy xy all core planes containing Bank "D" control rods and all unrodded core planes in a Radial Peaking factor limit Report per Specificallon 6.9.1.9; f.

The F,y limits of Specification 4.2.2.2e., above, are not applicable in the following core planes regions as measured in percent of core height from the bottom of the fuel:

1)

Iower core region from 0 to 15%, inclusive, 2)

Upper core region from 85 to 100%, inclusive, 3)

Grid plane regions at 17.8 i 2%, 32.1 1 2%, 46.4 1 2%, 60.6 1 2%

8

[,.'

and 74.9 i 2%, inclusive, and 4)

Core plane regions within i 2% of core height ( 2.88 inches) about the bank demand position of the Bank "0" control rods.

~

g.

With F, exceeding F, the effects of F,y on F (Z) shall be n

evaluated to determine if F (Z) is within its limits.

q 4.2.2.3 When F (Z) is measured for other than F determinations, an overall

~

9 xy measured I (I), hall be obtained from a power distribution map and increased q

by 37, to account f or manuf acturing tolerances and further increased by 5% to account 1or measurement uncertainty.

O e

CAtlAWAY - UNIT 1 3/4 2-7 c

y POWER DISTRIBUTION LIMITS N

3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR - F3H LIMITING CONDITION FOR OPERATION Fhshallbelimitedbythefollowingrelationship:

3.2.3 F g i 1.49 [1 + 0.3 (1-P)]

where P _ THERMAL POWER RATED THERMAL POWER F g = Measured values of F H btained by using the movable incore detectors to obtain a power distribution map.

The measured values of F shall be used since an uncertainty of 4% for H

incore measurement of F has been included in the above limit.

H APPLICABILITY:

MODE 1 ACTION:

With Fh exceeding its limit:

a.

Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either:

~

1.

Restore the F to within the above limits, or g

2.

Reduce THERMAL POWER TO LESS THAN 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux-High Trip Setpoint to 5 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b.

Demonstrate through in-core flux mapping that F is within H

its limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding the limit or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , and c.

Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER above the reduced limit required by a or b, above; subsequent POWER OPERATION may pro-is demonstrated through in-core flux

,ceed provided that F g mapping to be within its limit at a nominal 50% of RATED THERMAL POWER prior to exceeding this THERMAL POWEP, at a nominal 75% of RATED THERMAL POWER prior to exceeding this THERMAL power and within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after attaining 95% or greater RATED THERMAL POWER.

CALLAWAY - UNIT 1 3/4 2-8 Amendment No.

15

POWER DISTRIBUTION LIMITS N

NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR - F3H SURVEILLANCE REQUIREMENTS 4.2.3.1 F

shall be determined to be within its limit by using the movable g

incore detectors to obtain a power distribution map:

a.

Prior to operation above 75% of RATED THERMAL POWER after each fuel loading, and b.

At least once per 31 Effective Full Power Days.

c.

The provisions of Specification 4.0.4 are not applicable.

CALLAWAY - UNIT 1 3/4 2-9 Amendment No.

15

POWER DISTRIBUTION LIMITS 3/4.2.4 QUADRANT POWER TILT RATIO LIMITING CONDITION FOR OPERATION 3.2.4 The QUADRANT POWER TILT RATIO shall not exceed 1.02.

,0PLICABILITY:

MODE 1, above 50% of RATED THERMAL POWER.*

1 ACTION:

a.

With the QUADRANT POWER TILT RATIO determined to exceed 1.02 but less than or equal to 1.09:

1.

Calculate the QUADRANT POWER TILT RATIO at least once per hour until either:

a)

The QUADRANT POWER TILT RATIO is reduced to within its limit, or b)

THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER.

2.

Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either:

a)

Reduce the QUADRANT POWER TILT RATIO to within its limit, or b)

Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1 and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

3.

Verify that the QUADRANT POWER TILT RATIO is within its limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding the limit or reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High Trip Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , and 4.

Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified acceptable at 95%

or greater RATED THERMAL POWER.

See Special Test Exception Specification 3.10.2.

CALLAWAY - UNIT 1 3/4 2-10 Amendment No.15 y

-r--

POWER DISTRIBUTION LIMITS LIMITING' CONDITION FOR OPERATION ACTION (Continued) b.

With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to misalignment of either a shutdown or control rod:

1.

Calculate the QUADRANT POWER TILT RATIO at least once per hour until either:

a)

The QUADRANT POWER TILT RATIO is reduced to within its limit, or b)

THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER.

2.

Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1, within 30 minutes; 3.

Verify that the QUADRANT POWER TILT RATIO is within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after exceeding the limit or reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High Trip Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; and 4.

Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified acceptable at 95%

or greater RATED THERMAL POWER.

c.

With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to causes other than the misalignment of either a shutdown or control rod:

1.

Calculate the QUADRANT POWER TILT RATIO at least once per hour until either:

a)

The QUADRANT POWER TILT RATIO is reduced to within its limit, or b)

THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER.

CALLAWAY - UNIT 1 3/4 2-11 Amendment No. 15

r POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION ACTION (Continued) 2.

Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High Trip Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; and 3.

Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified at 95% or greater RATED T:lERMAL POWER.

d.

The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.2.4.1 The QUADRANT POWER TILT RATIO shall be determined to be within the

, limit above 50% of RATED THERMAL POWER by:

a.

Calculating the ratio at least once per 7 days when the alarm is OPERABLE, and b.

Calculating the ratio at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during steady-state operation when the alarm is inoperable.

4.2.4.2 The QUADRANT POWER TILT RATIO shall be determined to be within the limit when above 75% of RATED THERMAL POWER with one Power Range channel inoperable by using the movable incore detectors to confirm that the normalized symmetric power distribution, obtained from two sets of four symmetric thimble locations or a full core flux map, is consistent with the indicated QUADRANT POWER TILT RATIO at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

CALLAWAY - UNIT 1 3/4 2-12 Amendment No.15 9

POWER DISTRIBUTION LIMITS 3/4.2.5 DNB PARAMETERS LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB related parameters shall be maintained within the limits shown on Table 3.2-1:

Reactor Coolant System T,yg, a.

b.

Pressurizer Pressure, and c.

Reactor Coolant System Total Flow Rate.

APPLICABILITY:

MODE 1.

ACTION:

With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS 4.2.5.1 Each of the parameters of Table 3.2-1 shall be verified to be within their limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

~4.2.5.2 The calculated RCS total flow rate shall be determined to be greater than or equal to 382,630* GPM.

a.

Prior to operation above 75% of RATED THERMAL POWER after each fuel loading, and b.

At least once per 31 Effective Full Power Days.

4.2.5.3 The RCS loop flow rate indicators shall be subjected to a CHANNEL CALIBRATION at least once per 18 months.

4.2.5.4 The RCS total flow rate shall be determined by precision heat balance measurements at least once per 18 months.

Within 7 days of performing the precision heat balance, the instrumentation used for determination of steam pressure, feedwater temperature, and feedwater venturi AP in the calorimetric calculations shall be calibrated.

4.2.5.5 The feedwater venturi shall be inspected for fouling and cleaned as necessary at least once per 18 months.

The calculated value of RCS total flow rate shall be used since uncertainties of 2.2% for flow (including 0.1% for feedwater venturi fouling) measurement have been included in the above surveillance.

CALLAWAY - UNIT 1 3/4 2-13 Amendment No. 15

7-s TABLE 3.2-1 DNB PARAMETERS LIMITS Four Loops in PARAMETER-Operation Indicated Reactor Coolant System T,yg 5 593.4'F Indicated Pressurizer Pressure

> 2220 psig*

Calculated Reactor Coolant System Total Flow Rate

> 382,630** GPM

Limit not applicable during either a THERMAL POWER ramp in excess of 5% of RATED THERMAL POWER per minute or a THERMAL POWER step in excess of 10% of RATED THERMAL POWER.

- The calculated value of RCS total flow rate shall be used since uncertainties of 2.2% for flow (including 0.1% for feedwater venturi fouling) measurement have been inclLded i'n the above operating limit.

CALLAWAY - UNIT 1 3/4 2-14 Amendment No. 15

3/4.10 SPECIAt IEST EXCEPTIONS

=

3/4.10.1 SHUf00WN MARGIN LIMITING CONDITION FOR OPERATION 3.10.1 The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 may be suspended for measurement of control rod worth.and SHUTDOWN MARGIN provided reactivity equivalent to at least the highest estimated control rod worth is available for trip insertion from OPERABLE control rod (s).

APPIICA!11L11Y: MODE 2.

ACTION:

With any full-length control rod not fully inserted and with less

.i.

than the above reactivity equivalent.available for trip insertion, immediately initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm-boron or its equivalent until the SHUTDOWN MARGIN required by Specifi-cation 3.1.1.1 is restored.

b.

With all full-length control rods fully inserted and the reactor subcritical by less than the above reactivity equivalent, immediately initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm boron or its

-J equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1

-/

is restored.

[,

liitVI ILL ANCL RIQtJIREMENTS 4.10.1.1 lhe position of each full-length control rod either partially or fully withdrawn shall bc determined at least once per 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

4.10.J.7 Each full-length control rod not fully inserted shall be demonstrated capable of full insertion when tripped from at least the 50% withdrawn position within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to reducing the SHU100WN MARGIN to less than the limits of Specification 3.1.1.1.

(

Call AWAY - UNil 1 3/4 10-1

_n__,

SPECIAL TEST EXCEPTIONS 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION 3.10.2 The group height, insertion, and power distribution limits of Specifi-cations 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.1, and 3.2.4 may be suspended during the performance of PHYSICS TESTS provided:

a.

The THERMAL POWER is maintained less than or equal to 85% of RATED THERMAL POWER, and b.

The limits of Specifications 3.2.2 and 3.2.3 are maintained and determined at the frequencies specified in Specification 4.10.2.2 below.

APPLICABILITY:

MODE 1.

ACTION:

With any of the limits of Specifications 3.2.2 or 3.2.3 being exceeded while the requirements of Specifications 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.1 and 3.2.4 are suspended, either:

a.

Reduce THERMAL POWER sufficient to satisfy the ACTION requirements of Specifications 3.2.2 and 3.2.3, or

]

b.

Be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.10.2.1 The THERMAL POWER shall be determined to be less than or equal to 85%

of RATED THERMAL POWER at least once per hour during PHYSICS TESTS.

4.10.2.2 The requirements of the below listed specifications shall be performed at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during PHYSICS TESTS:

a.

Specifications 4.2.2.2 and 4.2.2.3, and b.

Specification 4.2.3.1.

CALLAWAY - UNIT 1 3/4 10-2 Amendment No. 15

~, - - - - -,

3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by:

(1) maintaining the minimum DNBR in the core at or above the safety analysis DNBR limits during normal operation and in short-term transients, and (2) limiting the fission gas release, fuel pellet temperature, and cladding mechanical properties to within assumed design criteria.

In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200*F is not exceeded.

The definitions of certain hot channel and peaking factors as used in these specifications are as follows:

F (Z)

Heat Flux Hot Channel Factor, is defined as the maximum local 0

heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods; F

Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of H

the integral of linear power along the rod with the highest integrated power to the average rod power; and xy(Z)

Radial Peaking Factor, is defined as the ratio of peak power density F

to average power density in the horizontal plane at core elevation Z.

3/4.2.1 AXIAL FLUX DIFFERENCE The limits on AXIAL FLUX DIFFERENCE (AFD) assure that the F (Z) upper bound 9

envelope of 2.32 times the normalized axial peaking factor is not exceeded during either normal operation or in the event of xenon redistribution following power changes.

Target flux difference is determined at equilibrium xenon conditions.

The full-length rods may be positioned within the core in accordance with their respective insertion limits and should be inserted near their normal position for steady-state operation at high power levels.

The value of the target flux difference obtained under these conditions divided by the fraction of RATED THERMAL POWER is the target flux difference at RATED THERMAL POWER for the associated core burnup conditions. Target flux differences for other THERMAL POWER levels are obtained by multiplying the RATED THERMAL POWER value by the appropriate fractional THERMAL POWER level.

The periodic updating of the target flux difference value is necessary to reflect core burnup considerations.

CALLAWAY - UNIT 1 B 3/4 2-1 Amendment No.

15 9

--~

p POWER DISTRIBUTION LIMITS BASES AXIAL FLUX DIFFERENCE (Continued)

Although it is intended that the plant will be operated with the AFD within the target band required by Specification 3.2.1 about the target flux difference, during rapid plant THERMAL POWER reductions, control rod motion will cause the AFD to deviate outside of the target band at reduced THERMAL POWER levels.

This deviation wil! not affect the xenon redistribution suffi-ciently to change the envelope of peaking factors which may be reached on a subsequent return to RATED THERMAL POWER (with the AFD within the target band) provided the time duration of the deviation is limited. Accordingly, a 1-hour penalty deviation limit cumulative during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is provided for operation outside of the target band but within the limits of Figure 3.2-1 while at THERMAL POWER levels between 50% and 90% of RATED THERMAL POWER.

For THERMAL POWER levels between 15% and 50% of RATED THERMAL POWER, deviations of the AFD outside of the target band are less significant.

The penalty of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months actual time reflects this reduced significance.

Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitor Alarm.

The computer determines the 1 minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for two or more OPERABLE excore channels are outside the target band and the THERMAL POWER is greater

.than 90% of RATED THERMAL POWER.

During operation at THERMAL POWER levels between 50% and 90% and between 15% and 50% RATED THERMAL POWER, the computer

, outputs an alarm message when the penalty deviation accumulates beyond the limits of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , respectively.

Figure B 3/4 2-1 shows a typical monthly target band.

3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR The limits on heat flux hot channel factor and nuclear enthalpy rise hot channel factor ensure that 1) the design limits on peak local power density and minimum DNBR are not exceeded and 2) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200*F ECCS acceptance criteria limit.

Each of these is measurable but will normally only be determined periodi-cally as specified in Specifications 4.2.2 and 4.2.3.

This periodic surveil-lance is s.pfficient to ensure that the limits are maintained provided:

a.

Control rods in a single c. cup move together with no individual rod insertion differing by more than i 12 steps, indicated, from the group demand position.

b.

Control rod 6anks are sequenced with overlapping groups as described in Specification 3.1.3.6.

CALLAWAY - UNIT 1 8 3/4 2-2 Amendment No.15

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i INDICATED AXIAL FLUX DIFFERENCE l

1 FIGURE B 3/4.2-1 TYPICAL INDICATED AXIAL FLUX DIFFERENCE VERSUS THERMAL POWER CALLAWAY - UNIT 1 B 3/4 2-3 Amendment No. 15

C z.

POWER DISTRIBUTION LIMITS BASES 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR AND NUCLEAR ENTHALPY RISE

-HOT CHANNEL FACTOR (Continued) c.

The control rod insertion limits of Specification 3.1.3.6 are maintained.

d.

The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.

FfH will be maintained within its limits provided conditions a. through

d. above are maintained.

The relaxation of F as a function of THERMAL POWER H

allows changes in the radial power shape for all permissible rod insertion limits.

When an F measurement is taken, an allowance for both experimental error q

and manufacturing tolerance must be made.

An allowance of 5% is appropriate for a full-core map taken with the incore detector flux mapping system and a 3% allowance is appropriate for manufacturing tolerance.

When Fh is measured, (i.e., inferred), no additional allowances are necessary prior to comparison with the limits of Section 3.2.3.

An error allow-tance of 4% has been-included in the limits of Section 3.2.3.

Margin between the

' ' safety analysis limit DNBRs (1.42 and 1.45 for thimble and typical cells, re-spectively) and the design limit DNBRs (1.32 and 1.34 for thimble and typical cells, respectively) is maintained.

A fraction of this margin is utilized to accommodate the transition core DNBR penalty (2%) and the appropriate fuel rod bow DNBR penalty (less than 3% per WCAP-8691, Rev.1). The 7% margin between design and safety analysis DNBR limits includes >2% margin for plant design flexibility.

The Radial Peaking Factor, Fxy (Z), is measured periodically to provide assurance that the Hot Channel Factor, F (Z), remains within its limit.

The RTP)q F

limit f r RATED THERMAL POWER (F as provided in the Radial Peaking xy x

Factor Limit Report per Specification 6.S.I.9 was determined from expected power control maneuvers over the full range of burnup conditions in the core.

3/4.2.4 QUADRANT POWER TILT RATIO The Q'UADRANT POWER TILT RATIO limit assures that the radial power distribution satisfies the design values used in the power capability analysis.

Radial power distribution measurements are made during STARTUP testing and periodically during power operation.

The limit of 1.02', at which corrective action is required, provides DNB and linear heat generation rate protection with x y plane power tilts.

A limit of 1.02 was selected to provide an allowance for the uncertainty associated with the indicated power tilt.

CALLAWAY - UNIT 1 B 3/4 2-4 Amendment No. 15

POWER DISTRIBUTION LIMITS BASES QUADRANT POWER TILT RATIO (Continued)

The 2-hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correc-tion of a dropped or misaligned control rod.

In the event such action does not correct the tilt, the margin for uncertainty on F is reinstated by reducing q

the maximum allowed power by 3% for each percent of tilt in excess of 1.

For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the movable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO. The incore detector monitoring is done with a full incore flux map or two sets of four symmetric thimbles.

The two sets of four symmetric thimbles is a unique set of eight detector locations.

These locations are C-8, E-5, E-11, H-3, H-13, L-5, L-11, N-8.

3/4.2.5 DN8 PARAMETERS The limits on.the DNB related parameters assure that each of the parameters is' maintained within the normal steady-state envelope of operation assumed in the transient and accident analyses.

The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain the safety analysis DNBR limit throughout each analyzed transient.

The indi-cated T,yg value of 593.4*F and the indicated pressurizer pressure value of 2220 psig correspond to analytical limits of 595.9*F and 2205 psig respectively, with allowance for measurement uncertainty.

The 12-hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.

When RCS flow rate is measured, no additional allowances are necessary prior to comparison with the limits of Section 3.2.5.

A measurement uncertainty of 2.2% (including 0.1% for feedwater venturi fouling) for RCS total flow rate has been allowed for in determination of the design DNBR value. The measure-i ment uncertainty for the RCS total flow rate is based upon performing a preci-sion heat balance and using the result to normalize the RCS flow rate indicators.

Potential fouling of the feedwater venturi which might not be detected could bias the result from the precision heat balance in a non-conservative manner.

Therefore, an inspection is performed on the feedwater venturi each refueling outage.

CALLAWAY - UNIT 1 B 3/4 2-5 Amendment No. 15

~

3/4.4 REACTOR COOLANT SYSTEM c

BASES 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with all reactor coolant loops in opera-tion and maintain DNBR above the safety analysis DNBR limits during all normal l

operations and anticipated transients.

In MODES 1 and 2 with one reactor coolant loop not in operation this specification requires that the plant be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

In MODE 3, two reactor coolant loops provide sufficient heat removal capability for removing decay heat even in the event of a bank withdrawal accident; however, single failure considerations require that three loops be

~

OPERABLE.

A single reactor coolant loop provides sufficient heat removal if a bank withdrawal accident can be prevented, i.e., by opening the Reactor Trip System breakers.

In MODE 4, and in MODE 5 with reactor coolant loops filled, a single reactor coolant loop or RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops (either RHR or RCS) be OPERABLE.

In MODE 5 with reactor coolant loops not filled, a single RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations, and the unavailability of the steam generators as a heat removing component, require that at least two RHR loops be OPERABLE.

~

The operation of one reactor coolant pump (RCP) or one RHR pump provides

' adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System.

The reactivity change rate associated with boron reduction will, J

therefore, be within the capability of operator recognition and control.

The restrictions on starting a reactor coolant pump in MODES 4 and 5 are provided to prevent RCS pressure transients, caused by energy additions from the Secondary Coolant System, which could exceed the limits of Appendix G E

to 10 CFR Part 50.

The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 50 F above each of the RCS cold leg temperatures.

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~

b CALLAWAY - UNIT 1 B 3/4 4-1 Amendment No.15 5

REACTOR COOLANT SYSTEM BASES 3/4.4.2 SAFETY VALVES The pressurizer Code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2735 psig.

Each safety valve is designed to relieve 420,000 lbs per hour of saturated steam.

The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during shutdown.

In the event that no safety valves are OPERABLE, an operating RHR loop, connected to the RCS, provides overpressure relief capability and will prevent RCS overpressurization.

In addition, the Over-pressure Protection System provides a diverse means of protection against RCS overpressurization at low temperatures.

During, operation, all pressurizer Code safety valves must be OPERABLE to prevent the MCS from being pressurized above its Safety Limit of 2735 psig.

The combined relief capacity of all of these valves is greater than the maximum surge rate resulting from a complete loss-of-load assuming no Reactor trip s,

and also assuming no operation of the power-operated relief valves or steam dump valves.

Demonstration of the safety valves' lift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Code.

3/4.4.3 PRESSURIZER The 12-hour periodic surveillance is sufficient to ensure that the para-meter is restored to within its limit following expected transient operation.

The maximum water volume also ensures that a steam bubble is formed and thus the RCS is not a hydraulically solid system.

The requirement that a minimum number of pressurizer heaters be OPERABLE enhances the capability of the plant to control Reactor Coolant System pressure and establish natural circulation.

3/4.4.4 RELIEF VALVES Ihe power-operated relief valves (PORVs) and steam bubble function to relieve RCS pressure.during all design transients up to and including the design step luad decrease with steam dump. Operation of the PORVs minimizes the undesirable opening of the spring-loaded pressurizer Code safety valves.

Lach PORV has a remotely operated block valve to provide a positive shutoff capability should a relief valve become inoperable.

CALLAWAY - UNil 1 B 3/4 4-2

_ _ _ _ _ _ _ _ _ _ _ _

ML20155F323

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