Text

Amend 28 to License NPF-30,modifying Tech Specs to Support Transition from Westinghouse 17 X 17 low-parasitic Fuel Assembly & Optimized Fuel Assembly Fueled Core to Westinghouse 17 X 17 Vantage 5 Fuel Assembly Fueled Core
	ML20236A687
Person / Time
Site:	Callaway
Issue date:	10/09/1987
From:	Wigginton D; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20236A690	List: ML20236A687; ML20236A697;
References
	NUDOCS 8710230032
	Download: ML20236A687 (51)
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_ _ _ - - _ _ - _ - _ _ _ _ _ _

/

'o UNITED STATES

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8

NUCLEAR REGULATORY COMMISSION o

>E WASHINGTON, D. C. 20555 UNION ELECTRIC COMPANY CALLAWAY PLANT, UNIT 1 DOCKET NO. STN 50-483 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 28 License No. NPF-30 1.

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

A.

The application for amendment filed by Union Electric Company (the licensee) dated March 31, 1987, as supplemented by letters dated April 15, June 5, June 18, July 16, July 28, August 7, August 13, August 31, September 9 and October 6, 1987 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR l

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

0.

The issuance of this amendment will not be inimical to the common I

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 Commission's regulations and all applicable requirements have been satisfied.

2.

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

8710230032 871009 PDR ADOCK 0500 3

P l

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. (2). Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 28, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are i

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 b.

j David L. Wiggint

, Acting Director Project Directorate III-3 Division of Reactor Projects

Attachment:

Changes to the Technical I

Specifications Date of Issuance:

October 9, 1987,

1 i

'I L___-----____---

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ATTACHMENT TO LICENSE AMENDMENT NO. 28' OPERATING LICENSE N0. NPF ;

DOCKET NO. 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 change. Corresponding overleaf pages are provided to maintain document completeness.

REMOVE INSERT l

1 I

II II V

V.

XX XX 1-5 1-5 q.

1-6 1-6 1-7 1-7 2-2 2-2 2-4 2-4 2-7 2-7 2-8 2-8 2-9 2-9 2-10 2-10 B 2-1 B 2-1 B 2-5 B~2-5 B 2-6 B 2-6 B 2 2-6(a) 3/4 1-19 3/4 1-19 3/4 2-1 3/4 2-1 3/4 2-2 3/4 2 3/4 2-2(a) i 3/4 2-6 3/4 2-6 i

I 3/4 2-7 3/4 2-7 3/42-7(a) 3/4 2-7(b)

J 3/4 2-14 3/4 2-14 3/4 3-12 3/4 3-12 3/4 3-12a 3/4 3-12a 3/4 5-1 3/4 5-1 l

B 3/4 2-1 8 3/4 2-1 B 3/4 2-2 B 3/4 2-2 B 3/4 2-4 B 3/4 2-4 8 3/4 2-5 B 3/4 2-5 B 3/4 2-6 B 3/4 2-6 6-21 6-21 6-22 6-22 i

i INDEX I

DEFINITIONS SECTION

_PAGE

_1. 0 DEFINITIONS 1.1 ACTI0N...........................................................

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

1-1 1.3

' ANALOG CHANNEL OPERATIONALTEST..................................

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 CONTROL L E D L EAKAG E....................................

1 - 2

-1.9 CO RE. AL TE RAT I ON......................................

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

1-2 1.11 DOS E EQU I VAL ENT I-131....................................

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

1-3 1.12 1.13 ENGINEERED SAFETY FEATURES RESPONSE TIME.........................

1-3 1.14 FREQUENCY NOTAT!0N...............................................

1-3 1.15 I DE NT I F I E D L EA KA G E...................................

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

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

1-3 1.18 0FFS I TE DOSE CALCUL ATION MANUAL......................

1.19 O P E RAB L E - O P E RAB I L I T Y..............................

1 - 4 1.20 OPERATIONAL MODE - M0DE..........................................

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

1-4 1.22 PRES SURE B0UNDARY L EA KAGE..............................

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

1-4 1.24 P U RG E - PU R G I N G........................................

1 - 4 1.25 QUADRANT POWER T ILT RATI0..................................

1 -5 1.26 RAT E D TH E RMAL P0WE R.....................................

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

1-5 1.28 R E PO RT AB L E E VE NT.......................................

1 - 5 1.29 RESTRICTE D AFD 0PERATION..................................

1 - 5 CALLAWAY - UNIT 1 I

Amendment No. JE, 28

i

.l INDEX i

DEFINITIONS.

SECTION PAGE DEFINITIONS (Continued) l 1.30 S HUT DOW N MAR G I N.................................................. 1 - 5

.l 1

1.31 S I T E B 0V N DARY.................................................... 1 - 5 j.

1 1.32-SLAVE RELAY TEST.................................................

1-5

]

l 1.33 SOLIDIFICATION...................................................

1 l l

1. 34. S O U RC E CH EC K.....................................................

1 - 6 l_

1.35 STAGGERED TEST BASIS.............................................

1-6 l.

l

'1.36 THERMAL P0WER....................................................

1-6 l

1.37. ' TRIP ACTUATING DEVICE 0PERATIONAL TEST........................... 1-6 l

1.38 UN I DE NT I F I E D L EA KAGE.............................................

1 - 6 l

1.39 UNRESTRICTED AREA............................................... 1-6 I

t.

I 1

1.40 VENTIL ATION EXHAUST TREATMENT SYSTEM.............................

1-7 l

I 1.41 VENTING...........................................................'l-7

~-l l

1.42 WASTE GAS HOLDUP SYSTEM..........................................

1-7 l

TABLE 1.1 FREQUENCY N0TATIONS.........................................

1-8 TABLE 1.2 OPERATIONAL M0DES...........................................

1-9 i

j.'

f.

)

l l-CALLAWAY - UNIT 1 II Amendment No. JE,28 L

i l

I i

a

{

l INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS l

SECTION PAGE l

3/4.2-POWER DISTRIBUTION LIMITS 3/4.2.1

-AXIAL FLUX DIFFERENCE.....................................

3/4 2-1 l

FIGURE 3.2-1 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATE D THERMAL P0WER.................................. 3/4 2-3 3/4.2.2 HEAT FLUX H0T CHANNEL FACTOR - Fg(Z)......................

3/4 2-4 l

FIGURE 3.2-2 K(Z)_ NORMALIZED F (Z) AS A FUNCTION OF CORE HEIGHT....

3/4 2-5 t

9 N

3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR - F3H............

3/4 2 l 3/4.2.4 QUADRANT POWER TILT RATI0.................................

3/4 2-10 3/4.2.5 DNB PARAMETERS............................................

3/4 2-13 l

TABLE 3.2-1 DNB PARAMETERS.........................................

3/4 2-14 3/4.3 INSTRUliENTATION 3/4.3.1 REACTOR-TRIP SYSTEM INSTRUMENTATION.......................

3/4 3-1 TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION....................

3/4 3-2 TABLE 3.3-2 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES.....

3/4 3-7 TABLE.4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS.........................................

3/4 3-9 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION.........................................

3/4 3-13 TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION..,...................................

3/4 3-14 TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETP0INTS.......................

3/4 3-22 TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES..............

3/4 3-29 TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS............

3/4 3-33 CALLAWAY - UNIT 1 V

Amendment No. 28

1 INDEX 1IMillNG CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS MC110N PAGE INSTRUMENTATION (Continued) 3/4.3.3 MONITORING INSTRUMENTATION l

Radiation Monitoring for Plant Operations.................

3/4 3-38

'ABLE 3.3 RADIATION MONITORING INSTRUMENTATION FOR PLANT 0PERATIONS.....................................

3/4 3-39 iABLE 4.3-3 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS SURVEILLANCE REQUIREMENTS...........

3/4 3-41 Movable'Incore Detectors..................................

3/4 3-42 Seismic Instrumentation...........................

3/4 3-43 TABLE 3.3-7 SEISMIC MONITORING INSTRUMENTATION.....................

3/4 3-44

~ABLE 4.3-4 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS............................

3/4 3-45 i

Meteorological Instrumentation..............

3/4 3-46

'ABLE 3.3-8 METEOROLOGICAL HONITORING INSTRUMENTATION..............

3/4 3-47 7ABLE 4.3-5 METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS...........................

3/4 3-48

-Remote Shutdown Instrumentation..........................

3/4 3-49

ABLE 3.3-9 REMOTE SHUTOOWN MONITORING INSTRUMENTATION............

3/4 3-50 TABLE 4.3-6 REMOTE SHUT 00WN MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS...........'................

3/4 3-51 i

Accident Monitoring Instrumentation......................

3/4 3-52

]

TABLE 3.3-10 ACCIDENT HONITORING INSTRUMENTATION......

3/4 3-53 TABLE 4.3 / ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS...........................

3/4 3-55 Fire Detection Instrumentation...........................

3/4 3-57 TABLE 3.3-11 FIRE DETECTION INSTRUMENTS............................

3/4 3-58 Loose-Part Detection System..............................

3/4 3-62 Radioactive Liquid Effluent Monitoring Instrumentation...

3/4 3-63 l

l l

I Call AWAY - UNil i VI l

l

INDEX J

ADMINISTRATIVE CONTROLS SECTION PAGE 6.1 RESPONS!81LITY................................................

6-1 6.2 ORGANIZATION 6.2.1 0FFSITE.....................................................

6-1 6.2.2 UNIT STAFF.......................................

6-1 FIGURE 6.2-1 0FFSITE ORGANIZATION.................................

6-3 FIGURE 6.2-2 UNIT ORGANIZAT10N....................................

6-4 TABLE 6.2-1 MINIMUM SHIFT CREW COMPOSITION.......................

6-5 6.2.3 INDEPENDENT SAFETY ENGINEERING GROUP (ISEG)

Function....................................................

6-6 Composition.................................................

6-6 Responsibilities...........................................

6-6 Records.....................................................

6-6 6.2.4 SHIFT TECHNICAL ADVIS0R.....................................

6-6 6.3 UNIT STAFF QUALIFICATIONS.....................................

6-6 6.4 TRAINING.......................

6-7 i

6.5 REVIEW AND AUDIT 6.5.1 ON-SITE REVIEW COMMITTEE (ORC)

Function....................................................

6-7 Composition................................................

6-7 Alternates..............

6-7 Meeting Frequency...........................................

6-7 Quorum......................................................

6-7 Responsibilities..................................

6-8 Records.....................................................

6-9 CALLAWAY - UNIT 1 XIX 1

INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.5.2 NUCLEAR SAFETY REVIEW BOARD (NSRB)

Function........................................................6-9 Composition.....................................................

6-10 A l te r n a te s...................................................... 6 - 10 C o n s ul t a n t s.....................................................

6-10 Meeting Frequency...............................................

6-10 Qualifications..................................................

6-10 Quorum..........................................................

6-10 Review..........................................................

6-11 Audits...............................

.......................... 6-11 Records.........................................................

6-12 6.5.3 TECHNICAL REVIEW AND CONTROL Activities......................................................

6-13 Records.........................................................

6-14 6.6 REPORTABLE EVENT ACTI0N...........................................

6-14 6.7 SAFETY LIMIT VIOLATION............................................

6-14 6.8 P ROC E DURE S AN D PR0G RAMS...........................................

6-15 6.9 REPORTING REQUIREMENTS 6.9.1 ROUTINE REP 0RTS.................................................

6-17 Startup Report..................................................

6-17 An n u a l Re po r t s..................................................

6-17 Annual Radiological Environmental Operating Report.............. 6-18 Semiannual Radioactive Effluent Release Report.................. 6-19 Mon thly Opera ti ng Re po r t........................................

6-20 Peaking Factor Limit Report.....................................

6-21 6.9.2 SPECIAL REP 0RTS.................................................

6-21 6.10 RECORD RETENTION.................................................

6-21 6.11 RADIATION PROTECTION PR0 GRAM.....................................

6-23 CALLAWAY - UNIT 1 XX Amendment No. 28

DEFINITIONS-

' QUADRANT POWER TILT RATIO 1.25 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper _ excore detector calibrated output to the average of the upper excore detector cali-brated outputs, or the ratio of the maximum lower excore detector calibrated cutput to the average of the lower excore detector calibrated outputs, which-ever 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 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 Section 50.73 to 10 CFR Part 50.

1 RESTRICTED AFD OPERATION 1.29 RESTRICTED AFD OPERATION (RAFDO) limits the AXIAL FLUX DIFFERENCE (AFD) to a +3% target band about the target flux difference and restricts power levels to between APLND and either APLRAFD0 or 100% RATED THERMAL POWER, whichever is less.

APLND and APLRAFD0 are defined in Specifications 3.2.1 and 4.2.2.3, respectively.

RAFD0 may be entered at the discretion of-the licensee.

SHUTDOWN MARGIN 1.30 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which l

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

l SITE 80VHDARY j

1. 31 The SITE B0UNDARY shall be that line beyond which the land is neither l

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

SLAVE RELAY TEST 1.32 A SLAVE RELAY TEST shall be the energization of each slave relay and l

verification of OPERABILITY of each relay.

The SLAVE RELAY TEST shall include e continuity check, as a minimum, of associated testable actuation devices, t

CALLAWAY - UNIT 1 1-5 Amendment No. JE, 28 2

DEFINITIONS SOLIDIFICATION 1.33 SOLIDIFICATION shall be the conversion of wet wastes into a form that meets shipping and burial ground requirements.

SOURCE CHECK 1

l 1.34 A SOURCE CHECK shall be the qualitative assessment of channel response l

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

STAGGERED 1EST BASIS 1.35 A STAGGERED TEST BASIS shall consist of:

l a.

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

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

THERMAL POWER 1.36 THERMAL POWER shall be the total core heat transfer rate to the reactor l

coolant.

TRIP ACTUATING DEVICE OPERATIONAL TEST 1.37 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 1.38 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE l

or CONTROLLED LEAKAGE.

UNRESTRICTED AREA 1.39 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 B0UNDARY used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes.

CALLAWAY - UNIT 1 1-6 Amendment No. H,28

1 1

~

DEFINITIONS'

~ VENTILATION EXHAUST TREATMENT SYSTEM i

1.40 A VENTILATION EXHAUST TREATMENT SYSTEM shall be any-system designed'and 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.

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

1 Engineered Safety Features (ESF) Atmospheric Cleanup Systems are not considered to be VENTILATION EXHAUST ~TREATNENT SYSTEM components, VENTING i

if 1

1

-1.41 VENTING shall be any controlled process of discharging air or gcs from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is not.

provided or required during VENTING.

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

' WASTE GAS HOLDUP SYSTEM

[

1.42 A WASTE GAS HOLDUP SYSTEM shall be any system designed and installed to reduce radioactive gaseous effluents by collecting Reactor' Coolant System off-j, gases from the Reactor Coolant System and providing for delay or holdup for.,

i the purpose of reducing the total radioactivity prior to release to the i

environment.

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CALLAWAY - UNIT 1 1-7 Amendment No. JE, 28

-TABLE 1.1 FREQUENCY NOTATION NOTATION FREQUENCY 5

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.

l R

At least once per 18 months.

S/U Prior to each reactor startup.

N.A.

Not applicable.

P Completed prior to each release.

l l

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

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3

.2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS q

i 2.l' SAFETY LIMITS I

REACTOR CORE-operating loop coolant temperature (Tavg) pressurizer pressure, and the hig 2.1.1 The combination of. THERMAL POWER, shall not exceed the limits shown in Figure 2.1-1 for four loop operation.

l l

APPLICABILITY:

MODES 1 and 2.

j

' ACTION:'

Whenever the point defined by the combination of the highest operating loop j

average temperature and THERMAL POWER has exceeded the appropriate pres-

^

surizer pressure line, be in HOT STANDBY 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 Reac^or Coolant System pressure has exceeded 2735 psig, be j

in HOT STANDBY 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:

l 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.

l l

1 CALLAWAY - UNIT 1 2-1

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L; Lu SAFETY' LIMITS 'AND LIMfTING SAFETY SYSTEM SETTfNGS L

2.2 LIMITING SAFETY SYSTEM SETTINGS y

a-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.

q j

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

' ACTION:

J 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 colamn of Table 2.2-1, adjust the Setpoint consistent with the Trip Setpoint value, j

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 l

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.

l Equation 2.2-1 2 + R + S < TA Where:

Z=

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

The"asmeasured"value(inpercentspan)ofrackerrorforthe affected channel, Eitherthe"asmeasured"value(in S=

error, or the value from Column S (percent span) of the sensor 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.

CALLAWAY - UNIT 1 2-3

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Il
e 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 products 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 result in excessive cladding temperatures because of the onset of departure i
from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient, DNB is not a directly measurable parameter during operation and I
therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB. This relation has been developed to predict the DNB flux and j
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 i
would cause DNB at a particular core location to the local heat flux, is indic-ative of the margin to DNB.
j 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 4
events is greater than or equal to the DNBR limit of the DNB correlation being used (the WRB-1 correlation for Optimized fuel (OFA) and the WRB-2 correlation for VANTAGE 5 fuel in this application). The correlation DNBR limit is estab-i lished 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 both the WRB-1 and WRB-2 correlations).
In meeting this design basis, uncertainties in plant operating parameters, 1
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 j
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, q
For Callaway, the design DNBR values are 1.32 and 1.34 for thimble and typical cells, respectively, for 0FA, and 1.32 and 1.33 for thimble and typical cells, respectively, for VANTAGE 5 fuel.
In addition, margin has been maintained in both fuel designs by meeting safety analysis DNBR limits of 1.42 and 1.45 for j
thimble and typical cells, respectively, for 0FA, and 1.61 and 1.69 for thimble and typical cells, respectively, for VANTAGE 5 fuel.
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 calculated DNBR is no less than the design DNBR value or the average enthalpy i
at the vessel exit is less than the enthalpy of saturated liquid.
CALLAWAY - UNIT 1 B 2-1 Amendment No. LB, 28
SAFETY LIMITS BASES 2.1.1 REACTOR CORE (Continued)
The curves are based on a nuclear enthalpy rise hot channel factor, Ffg, 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 3is 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.
1 CALLAWAY - UNIT 1 B 2-2 Amendment No. 15
4 LIMITING SAFETY SYSTEM SETTINGS BASES Intennediate and Source Range, Neutron Flux The Intermediate and Source Range, Neutron Flux trips provide core protec-tion during reactor startup to mitigate the consequences of an uncontrolled rod cluster control assembly bank withdrawal from a subcritical condition.
Neutron Flux channels. trips provide redundant protection to the Low Setpoint tri These The Source Range channels will initiate a Reactor trip at about 105 counts per second unless manually blocked when P-6 becomes active.
The Intermediate Range channels will initiate a Reactor trip at a current level equivalent to approximately 25% of RATED THERMAL POWER unless manually block when P-10 becomes active.
Overtemperature AT The Overtemperature AT trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power distribu-tion, provided that the transient is slow with respect to pi from the core to the temperature detectors (about 4 seconds) ping transit delays
, and pressure is within the range between the Pressurizer High and Low Pressure trips.
Setpoint is automatically varied with:
The (1) coolant temperature to correct for temperature induced changes in density and heat capacity of water and includes dynamic compensation for piping dela detectors, (2) pressurizer pressure,ys from the core to the loop temperature and (3) axial power distribution. With normal axial power distribution, this Reactor trip limit is always below the core Safety Limit as shown in Figure 2.1-1.
If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the Reactor trip is automatically reduced according to the notations in Table 2.2-1.
Delta-To, as used in the Overtemperature and Overpower AT trips, represents the 100% RTP value as measured by the plant for each loop.
This nonnalizes each loop's AT trips to the actual operating conditions existing at the time of measurement, thus forcing the trip to reflect the equivalent full power condi-tions as assumed in the accident analyses. These differences in vessel AT can arise due to several factors, the most prevalent being measured RCS loop flows greater than Minimum Measured Flow, and slightly asymetric power distributions between quadrants.
While RCS loop flows are not expected to change with cycle life, radial power redistribution between quadrants may occur, resulting in small changes in loop specific vessel AT values.
loop specific vessel AT value should be made when performinAccurate determination quarterly recalibration and under steady state conditions (g the Incore/Excore tions not affected by Xe or other transient conditions).
i.e., power distribu-Overpower AT The Overpower AT trip provides assurance of fuel integrity (e.g., no fuel pellet melting and less than 1% cladding strain) under all possible overpower conditions, limits the required range for Overtemperature AT trip, and provides l
CALLAWAY - UNIT 1 i
B 2-5 Amendment No. 28 j
I
LIMITING SAFETY SYSTEM SETTINGS BASES Overpower AT (Continued) a backup to the High Neutron Flux trip. The Setpoint is automatically varied with:
(1) coolant temperature to correct for temperature induced changes in density and heat capacity of water, and (2) rate of change of temperature for dynamic compensation for piping delays from the core to the loop temperature detectors, to ensure that the allowa'ble heat generation rate (kW/ft) is not exceeded. The Overpower AT trip provides protection to mitigate the conse-quences of various size steam breaks as reported in WCAP-9226, " Reactor Core
. Response to Excessive Secondary Steam Releases."
Delta-T, as used in the Overtemperature and Overpower AT trips, represents o
the 100% RTP value as measured by the plant for each loop. This normalizes each loop's AT trips to the actual operating conditions existing at the time of g
measurement, thus forcing the trip to reflect the equivalent full power condi-tions as assumed in the accident analyses. These differences in vessel AT can arise due to several factors, the most prevalent being measured RCS loop flows greater than Minimum Measured Flow, and slightly asymmetric power distributions between quadrants. While RCS loop flows are not expected to change with cycle life, radial power redistribution between quadrants may occur, resulting in small changes in loop specific vessel AT values. Accurate determination of the quarterly recalibration and under steady state conditions (g the Incore/Excore loop specific vessel AT value should be made when performin i.e., power distri-butions not affected by Xe or other transient conditions).
Pressurizer Pressure In each of the pressurizer pressure channels, there are two independent bistables, each with its own Trip Setting to provide for a High and Low Pressure trip thus limiting the pressure range in which reactor operation is permitted. The Low Setpoint trip protects against low pressure which could lead to DNB by tripping the reactor in the event of a loss of reactor coolant pressure.
On decreasing power the Low Setpoint trip is automatically blocked by P-7 (a power level of approximately 10% of RATED THERMAL POWER with turbine impulse chamber pressure at approximately 10% of full power equivalent); and on increasing power, automatically reinstated by P-7.
The High Setpoint trip functions in conjunction with the pressurizer relief and safety valves to protect the Reactor Coolant System against system overpressure.
l Pressurizer Water Level l
The Pressurizer High Water Level trip is provided to prevent water relief through the pressurizer safety valves. On decreasing power the Pressurizer High Water Level trip is automatically blocked by P-7 (a power level of CALLAWAY - UNIT 1 B 2-6 Amendment No.
28
i LIMITING' SAFETY SYSTEM SETTINGS l
BASES i
Pressurizer Water Level (Continued) l l
approximately.10% of EATED THERMAL POWER with a turbine impulse chamber pressure at approximately 10% of full power equivalent); and on increasing power, automatically reinstated by P-7.
Reactor Coolant Flow i
The Low Reactor Coolant Flow trips provide core protection to prevent DNB by mitigating the consequences of a loss of flow resulting from the loss of one or more reactor coolant pumps.
On increasing power above P-7 (a power level of approximately 10% of RATED THERMAL POWER or a turbine impulse chamber pressure at approximately 10% of full power equivalent) an automatic' Reactor trip will occur if the flow in more than one loop drops below 90% of nominal full loop flow. Above i
P-8 (a power level of approximately 48% of RATED THERMAL POWER) an automatic Reactor trip will occur if the flow in any single loop drops below 90% of nominal full loop flow. Conversely, on decreasing power between P-8 and P-7 an automatic Reactor trip will occur on low reactor coolant flow in more than one loop and below P-7 the trip function is automatically blocked.
I i
i l
l CALLAWAY - UNIT 1 B 2-6(a)
Amendment No.
28 i
e
'i 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 full'y withdrawn position shall be less than or equal to 2.7. 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 th'e rod drop time of any full-length red 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.
J SURVEILLANCE REQUIREMENTS 4.1.3.4 The rod drop time of full-length rods shall be demonstrated through measurement prior to reactor criticality:
a.
For all rods following each removal of the reactor vessel head, i
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. JE,28
I RfACTIVfTY CONTROL SYSTEMS
.]
SHUT 00WN ROD INSERTION LIMIT l
1 l
LIMITING CONDITION FOR OPERATION
'i l
3.1.3.5 All shutdown rods shall be fully withdrawn.
APPLICABILITY: MODES 1* and 2*#.
I 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.
I SURVEILLANCE REQUIREMENTS 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 thereafter.
I
~~See Special Test Exceptions Specifications 3.10.2 and 3.10.3.~~
~~With K,ff greater than or equal to 1.~~
I i
l 1
' CALLAWAY - UNIT 1 3/4 1-20 L__ __ _ _ _ __
3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within the following target band (flux difference units) about the target flux difference:
a.
+3%, -12% for Normal Operation b.
+3% for RESTRICTED AFD OPERATION The indicated AFD ma deviate outside the applicabggequired targt band at greater than or equa to 50% but less than 0.9 APL or 90% of TED TriERMAL POWER, whichever is less, provided the indicated AFD is within the Acceptable Operation Limits of Figure 3.2-1 and the cumulative penalty deviation times does not exceed 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
The indicated AFD may deviate outside the applicable required target band at I
greater than 15% but less than 50% of RATED THERMAL POWER provided the cumula-tive penalty deviation time does not exceed 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
APPLICABILITY: MODE 1, above 15% of RATED THERMAL POWER *,#
l ACTION:
a.
With the indicated AFD outside of the applicable required target ND**
band and with THERMAL POWER greater than or equal to 0.9 APL or 90% of RATED THERMAL POWER, whichever is less, within 15 minutes, either:
1.
Restore the indicated AFD to within the applicable required I
target band limits, or
~~See Special Test Exception Specification 3.10.2.~~
Surveillance testing of the Power Range Neutron Flux channel may be per-formed pursuant to Specification 4.3.1.1 provided the indicated AFD is maintained within g0g, Acceptable Operation Limits of Figure 3.2-1 and h
THERMAL POWER 1APL A total of 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months operation may be accumulated with the AFD outside of the applicable required target band during testing without penalty deviation.
~~APLND is the minimum allowable power level for RESTRICTED AFD OPERATION and will be provided in the Peaking Factor Limit Report per Specification 6.9.1.9.~~
~~APLN0 is equal to the maximum~~
~~" 2.32 K(Z)~~
~~100 over Z~~
,F(Z)* W(2)N0 and F (2) and W(Z)N0 are defined in 4.2.2.2.c.
CALLAWAY - UNIT 1 3/4 2-1 Amendment No. 28
7 8:
.1
\\
l I
POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION
~
L ACTION (Continued) 1 2.
Reduce THERMAL' POWER to less than 0.9 APLND** or 90% of RATED I
THERMAL POWER, whichever is less, and discontinue RESTRICTED I
AFD OPERATION (if ' applicable).
b.
With the indicated AFD outside of the applicable required target a
band for more than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of cumulative penalty deviation time during ths previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or outside the Acceptable Operation Limits of Figure 3.2-1 and with THERMAL POWER less.than 0.9 APLND**
or 90%, whichever is less, but equal to or' greater than 50% of RATED THERMAL POWER, reduce:
1.
THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes, and 2.
The Power Range Neutron Flux-High Setpoints to less than or i
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 .
c.
With the indicated AFD outside of the applicable required target I
band for more than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of cumulative penalty deviation time during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and with THERMAL POWER less than 50% but greater than 15% of RATED THERMAL POWER, the THERMAL POWER shall not be increased equal to or greater than 50% of RATED THERMAL POWER until the indicated AFD is within the applicable required target band.
SURVEILLANCE REQUIREMENTS 4.2.1.1 The indicated AFD shall be determined to be within its limits during POWER OPERATION above 15% of RATED THERMAL POWER by:
a.
Monitoring the indicated AFD for each OPERABLE excore channel:
1.
At least once per 7 days when the AFD Monitor Alarm is OPERABLE, and 2.
At least once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after restoring the AFD Monitor Alann to OPERABLE status.
b.
Monitoring and logging the indicated AFD for each OPERABLE excore channel at least once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and at least once per 30 minutes thereafter, when the AFD Monitor Alarm is inoperable. The logged values of the indicated AFD shall be assumed to exist during the interval preceding each logging.
I CALLAWAY - UNIT 1 3/4 2-2 Amendment No. 28
POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS
~~4. 2.1. 2 The indicated AFD shall be considered outside of its target band when two or more OPERABLE excore channels are indicating the AFD to be outside the target band.~~
Penalty deviation outside of the above required target band shall be accumulated on a time basis of:
1 a.
One minute penalty deviation for each 1 minute of POWER OPERATION outside of the target band at THERMAL POWER levels equal to or above 50% of RATED THERIML POWER, and b.
One-half minute penalty deviation for each 1 minute of POWER OPERA-TION outside of the target band at THERMAL POWER levels between 15%
and 50% of RATED THERMAL POWER.
4.2.1.3 The target flux difference of each OPERABLE excore channel shall be determined by measurement at least once per 92 Effective Full Power Days. The provisions of Specification 4.0.4 are not applicable.
4.2.1.4 The target flux difference shall be updated at least once per 31 Effec-tive Full Power Days by either determining the target flux difference pursuant to Specification 4.2.1.3 above or by linear interpolation between the most recently measured value and 0% at the end of the cycle life. The provisions of Specification 4.0.4 are not applicable.
CALLAWAY - UNIT 1 3/4 2-2(a)
Amendment No. 28 j
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i POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.
For Normal Operation, F (z) shall be evaluated to determine if F (z) 4.2.2.2 Q
Q is within its limit by:
a.
Using the movable incore detectors to obtain a power distribution j
map at any THERMAL POWER greater than 5% of RATED THERMAL POWER.
]
Increasing the measured F (z) component of the power distribution I
b.
Q map by 3% to account for manufacturing tolerances and further increasing the value by 5% to account for measurement uncertainties.
c.
Satisfying the following relationship:
M Fg (z) < 2.32 x K(z) for P > 0.5 P x W(z)NO M
Fq (z) < 2.32 x K(z) for P < 0.5 W(z)N0 x 0.5 whereFd(z)isthemeasuredFg(z)increasedbytheallowancesfor manufacturing) tolerances and measurement uncertainty, 2.32 is theis ga Fo limit, K(z POWER, and W(z)N0 is the cycle dependent, Normal Operation function that accounts for power distribution transients encountered during Normal Operation. This function is given in the Peaking Factor Limit Report as per Specification 6.9.1.9.
d.
Measuring FQ (z) according to the following schedule:
l M
1.
Upon achieving equilibrium conditions after exceeding, by 10%
or more of RATED THERMAL POWER, the THERMAL POWER at which Fq(z) was last determined,* or 2.
At least once per 31 Effective Full Power Days (EFPD), whichever i occurs first.
~~During power escalation at the beginning of each cycle, power level may be increased until a power level for extended operation has been achieved and a power distribution map obtained.~~
CALLAWAY - UNIT 1 3/4 2-6 Amendment No. 28
4 q
' POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) 4.2.2.2 (Continued) e.
With measurements indicating Fh(z) j maximum over z K(z)
+
M has increased since the previous detennination of FQ (z), either i
of the following actions shall be taken:
M Fg (z) shall be increased by 2% over that specified in 1.
Specification 4.2.2.2c., or >
M Q (z) shall be measured at least once per 7 Effective Full 2.
F Power Days until two successive maps indicate that maximum Fh(z) is not increasing.
over z K(z) f.
With the relationships specified in Specification 4.2.2.2c. above not being satisfied:
1.
Calculate the percent Fg(z) exceeds its limit by the following p
expression:
(max. over z of F((z) x W(z)NO )-1 x 100 for P > 0. 5 2.32 x K(2) p (max.overzof Fh(z) x W(z)N0
)-1 x 100 for P < 0.5 2.32 x K(z) 0.5 2.
Either one of the following actions shall be taken:
(a) Comply with the requirements of Specification 3.2.2 for Fo(z) exceeding its limit by the percent calculated aBove, or (b) Verify that the requirements of Specification 4.2.2.3 for RESTRICTED AFD OPERATION are satisfied and enter RESTRICTED AFD OPERATION.
g.
The limits specified in Specifications 4.2.2.2.c., 4.2.2.2.e., and 4.2.2.2.f. above are not applicable.in the following core plane regions:
1.
Lower core region from 0 to 15%, inclusive.
2.
Upper core region from 85 to 100%, inclusive.
i CALLAWAY - UNIT 1 3/4 2-7 Amendment No. 28 g
(
. y 6
POWER DISTRIBUTION LIMITS SURVEILLANCE' REQUIREMENTS (Continued) a 4.2.2.3 RESTRICTED AFD OPERATION (RAFD0) is permitted at powers above APLND if the following conditions are satisfied:
a.-
Prior to entering RAFD0, maintain THERMAL POWER above APLND and less j
than or equal to that allowed by Specification 4.2.2.2 for at least the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Maintain RAFD0 surveillance (AFD within +3%
of target flux difference) during this time period.
RAFD0 is then APLgd providing THERMAL POWER is maintained between APLND and per N
or between APL D and 100% (whichever is more limiting) and FQ surJv illance is maintained pursuant to Specification 4.2.2.4.
APLRAFuv is defined as:
APLRAFD0 = minimum '2.32 x K(z) x 100%
~
N ver z
.Fg(z)xW(z)kAFD0 F$(z)isthe.measuredF(z)increasedbytheallowancesfor where:
Q manufacturing tolerances and measurement uncertainty. The FQ limit is 2.32.
K(z) is given in Figure 3.2-2.
W(z)RAFD0 is the cycle dependent function that accounts for limited power distribution transients encountered during RAFD0. This function is given in the Peaking Factor Limit Report as per Specification 6.9.1.9.
b.
During RAF00, if the THERMAL POWER is decreased below APLND then the conditions of 4.2.2.3.a shall be satisfied before re-entering RAFD0.
During RAFD0, F (z) shall be evaluated to determine if F (z) is 4.2.2.4 Q
Q within its limits by:
a.
Using the movable incore detectors to obtain a power distribution map at any THERMAL POWER above APLND, i
b.
Increasing the measured Fg(z) component of the power distribution map by 3% to account for manufacturing tolerances and further increasing the'value by 5% to account for measurement uncertainties.
c.
Satisfying the following relationship:
EI ND F (z) < 2.32 x K(z) for P > APL q
_ P x W(z) MFD 0 i
MF(z)isthemeasuredF(z). The Fg limit is 2.32.
K(z)
)
where:
0 Q
is given in Figure 3.2-2.
P is the relative THERMAL POWER. W(z)RAFD0 is the cycle dependent function that accounts for limited power j'
distribution transients encountered during RAFDO. This function is given in the Peaking Factor Limit Report as per Specification 6.9.1.9.
.CALLAWAY - UNIT 1 3/4 2-7(a)
Amendment No. 28 i
I
i
[ POWER'D1STRIBUTIONLIMITS SURVEILLANCE REQUIREMENTS (Continued) 1 4.2.2.4 (Continued) d.
MeasuringF$(z)inconjunctionwithtargetfluxdifferencedetermi-i nation according to the following schedule:
+
1.
Prior to entering RAFD0 af ter satisfying Section 4.2.2.3 unless a full core flux map has been taken in the previous 31 EFPD wi above APL gh the relative thermal power having been maintained N
for the 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to mapping, and i
2.
At least once per 31 Effective Full Power Days.
f e.
With measurements indicating Fh(z) maximum over z
,K(z),
hasincreasedsincethepreviousdeterminationofFh(z)eitherof the following actions shall be taken:
F (z) shall be increased by 2 percent over that specified in 1.
9 4.2.2.4.c, or 2.
F (z) shall be measured at least once per 7 EFPD until two s ccessive maps indicate that
~h(z)~isnotincreasing, maximum F
over z K(z) f.
With the relationship specified in 4.2.2.4.c above not being satisfied, comply with the requirements of' Specification 3.2.2 for F (z) exceeding its limit by the percent calculated with the Q
following expression:
(max. over z of Fh(z)xW(z)RAFD0 )-1 x 100 for P > APLND 2.32 x K(z)
P g.
The limits specified in 4.2.2.4.c, 4.2.2.4.e, and 4.2.2.4.f above are not applicable in the following core plane regions:
1.
Lower core region from 0 to 15 percent, inclusive.
2.
Upper core region from 85 to 100 percent, inclusive.
4.2.2.5 When Fo(z) is measured for reasons other than meeting the re ments of Specification 4.2.2.2 or 4.2.2.4, an overall measured Fg(z) quire-shall be obtained from a power distribution map and increased by 3% to account for manufacturing tolerances and further increased by 5% to account for measure-ment uncertainty.
CALLAWAY - UNIT 1 3/4 2-7(b)
Amendment No. 28
POMER DISTRIBUTION LfMITS NUCLEARENTHALPYRISEHOTCHANNELFACTOR-Fh 3/4.2.3 LIMITING CONDITION FOR OPERATION 3.2.3 F shall be limited by the following relationship:
H F
1 1.49 [1 + 0.3 (1-P)]
H where P = THERMAL POWER RATED THERMAL POWER btained by using the movable incore F H = Measured values of F H detectors to obtain a power distribution sap. The measured values of Fh shall be used since an uncertainty of 4% f'or incore measurement of Fh has been included in the above limit.
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:
RestoretheFfH to within the above limits, or 1.
2.
Reduce THERMAL POWER TO LESS THAN 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux-High Trip Setpoint to 155% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .
Demonstrate through in-core flux mapping that Fh is within b.
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 Identify and correct the cause of the out of limit condition c.
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 reed provided that F H mapping to be within its limit at a nominal 50% of RATED THERMAL POWER prior to exceeding this THERMAL POWER, 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.
l CALLAWAY - UNIT 1 3/4 2-8 Amendment No.
15
' TABLE 3.2-1 DNB PARAMETERS
~ LIMITS Four loops in PARAMETER Operation
' Indicated Reactor Coolant System.T,yg
< 592.6 F
,_ 2220 psig*
d
' Indicated ' Pressurizer Pressure :
Calculated Reactor Coolant System Total Flow Rate
> 382,630**GPM-
j
.a I
i 1
,I
~l i
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I
~~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 included in the above operating limit.
CALLAWAY - UNIT 1 3/4 2-14 Amendment No. JE, 28 i
._-__2__-_________-_________
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TABLE 4.3-1 (Continued) 1 TABLE NOTATIONS
~~0nly if the Reactor Trip System bmakers happen to be closed and the~~
. {
Control Rod Drive System is capable of rod withdrawal.
/
The specified 18 month frequency may be waived for Cycie I provided the surveillance is performed prior to restart following tha first refueling outage or June 1,1986, whichever occurs first. Tne provisions of Specification 4.0.2 are reset from performance of this surveillance.
f
~~Below P-6 (Intermediate Range Neutron Flux interlock)' Setpcint.~~
~~Below P-10 (Low Setpoint Power Range Neutron Flux interlock) Setpoint.~~
(1)
If not performed in previous 31 days.
(2) Comparison of calorimetric to excore power dndication above 15% of RATED THERMAL POWER. Adjust excore channel gains consistent with calorimetric power if absolute difference is greater than 2%. The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.
(3) Single point comparison of incore to excore AXIAL FLUX DIFFERENCE above 15% of RATED THERMAL POWER. Recalibrates if the absolute dif fe.rence is greater than or equal to 3f The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.
(4) Neutron detectors may be excluded from CHANNEL CALIBRATION.
(5)
Detector plateau curves shall be obtained, evaluated and compared to manu-facturer's data. For the Intennediate Range and Power Range Neutron Flux' 3
channels the provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.
(6)
Incore - Excore Calibration, above 75% of RATED THERMAL POWER. The provi-sions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.
Determination of the loop specific vessel AT value should be made when performing the Incore/Excore quarterly recalibration, under steady state conditions.
(7) Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS.
The TRIP ACTUATING P VICE OPERATIONAL TEST shall independently verify the OPERABILITY of the bodervoltage and Shunt Trip Attachments of the Reactors i
Trip Breakers.
l l
(8)
Deleted l
(9) Quarterly surveillance in MODES 3*,
4*, and 5* shall also include verifica-tion that permissives P-6 and P-10 are in theirJrequired state for existing plant conditions by observation of the pennissive annunciator window.
Quarterly surveillance shall include verification of the Boron Dilution Alarm Setpoint of less than or equal to an increase of twice the count rate within a 10-minute period.
l CALLAWAY - UNIT 1 3/4 3-12 Amendment No. 19, 28
TABLE 4.3-1 (Continued)-
TABLE NOTATIONS (10). Setpoint verification is not required.
l (11) Following maintenance or adjustment of the Reactor trip breakers, the TRIP ACTUATING DEVICE OPERATIONAL TEST shall include independent verifi-cation of the Undervoltage and Shur t trips.
(12) At least once per 18 months during shutdown, verify that on a simulated Boron Dilution Doubling test signal the normal CVCS discharge valves will
~~lose and the. centrifugal charging pumps suction valves from the RWST will open within 30 seconds.~~
(13) CHANNEL CALIBRATION shall include the RTD bypass loops flow rate.
i (14) Each channel shall be tested at least every 92 days on a STAGGERED TEST BASIS.
(15) The surveillance frequency and/or MODES specified for these channels in Table 4.3-2 are more restrictive and, therefore, applicable.
(16) The. TRIP ACTUATING DEVICE OPERATIONAL TEST shall independently verify the OPERABILITY of the Undervoltage and Shunt Trip circuits for the Manual Reactor Trip function. The test shall also verify the OPERABILITY of the Bypass Breaker trip circuit.
CALLAWAY - UNIT 1 3/4 3-12a Amendment No.19,28
3/4.5 ENERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS LIMITING CONDITION FOR OPERATION i
3.5.1 Each Reactor Coolant System accumulator shall be OPERABLE with.
a.
The isolation valve open and power removed, b.
A contained borated water volume of between 6061 and 6655 gallons, l
A boron concentration of between 1900 and 2100 ppm, and c.
d.
A nitrogen cover pressure of between 602 and 648 psig.
APPLICABILITY: MODES 1, 2, and 3*.
ACTION:
With one accumulator inoperable, except as a result of a closed a.
isolation valve, restore the inoperable accumulator to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , b.
With one accumulator inoperable due to the isolation valve being closed, either immediately open the isolation valve or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
SURVEILLANCE REQUIREMENTS 4.5.1.1 Each accumulator shall be demonstrated OPERABLE:
a.
At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by:
1)
Verifying, by the absence of alarms, the containeu borated water volume and nitrogen cover pressure in the tanks, and 2)
Verifying that each accumulator isolation valve is open.
~~Pressurizer pressure above 1000 psig.~~
CALLAWAY - UNil 1 3/4 5-1 Amendment No. 28
-j
EMERGENCY CORE COOLING SYSTEMS I
I SURVLILLANC.E_ REQUIREMENTS (Continued) b.
At'least once per 31 days and within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after each solution volume increase of greater than or equal to 70 gallons by verifying the boron concentration of the accumulator solution; and l
c.
At least once per 31 days when the RCS pressure is above 1000 psig
~~by verifying that the circuit breaker supplying power to the isola-tion valve operator is open.~~
4.5.1.2 Each accumulator water level and pressure channel shall be demonstr-ated OPERABLE at least once per 18 months by the performance of a CHANNEL CALIBRATION.
1
)
CAllAWAY - UNil 1 3/4 5-2 l
l
3/4.2 p0WER DISTRIBUTION LIMITS
(
BASES
\\
The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency)
(1) maintaining the minimum DNBR in the core at or above the safety l
events by:
analysis DNBR limits during normal operation and in short-term transients, and l
(2) limiting the fission gas release, fuel pellet temperature, and claddingIn addition, mechanical properties to within assumed design criteria.
the peak linear power density during Condition I events provides assurance that i
the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200 F is not exceeded.
The definition of certain hot channel and peaking factors as used in these specifications are as follows:
Heat Flux Hot Channel Factor, is defined as the maximum local heat flux F(Z) g 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; and Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of N
F the integral of linear power along the rod with the highest integrated AH power to the average rod power.
l 3/4.2.1 AXIAL FLUX DIFFERENCE The limits on AXIAL FLUX DIFFERENCE (AFD) assure that the Fg(Z) upper bound 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.
The Target flux difference is determined at equilibrium xenon conditions.
full-length rods may be positioned within the core in accordance with their respective insertion limits and should be inserted near their normal position j
The value of the target flux for steady-state operation at high power levels.
difference obtained under these conditions divided by the fraction of RATED THERMAL POWER is the target flux difference at RATED THERMAL POWER for the Target flux differences for other THERMAL associated core burnup conditions.
POWER levels are obtained by multiplying the RATED THERMAL POWER value by the The periodic updatino of the target appropriate fractional THERMAL POWER level.
flux difference value is necessary to reflect core burnup considerations.
The limits on /XIAL FLUX DIFFERENCE (AFD) are given in Specification 3.2.1.
One mode is Normal Operation, where the Two modes of operation are permissible.
The AFD limit for this i
applicable AFD limit is defined by Specification 3.2.1.a.
I mode of operation is a +3, -12% target band about the target flux difference.
After extended load following maneuvers, the AFD limits may result in restric-tions in the maximum allowed power to quarantee operation with Fg(Z) less than its limiting value. To prevent this occurrence, another operating mode which j
,CALLAWAY - UNIT 1 B 3/4 2-1 Amendment No. JE, 28 j
4 POWER DISTRIBUTION LIMITS BASES 3/4.2.1 AXIAL FLUX DIFFERENCE (Continued) restricts the AFD to a relatively small target band and does not allow signif-icant changes in power level has been defined. This mode is called RESTRICTED AFD OPERATION, which restricts the. AFD to a 13% target band about the target flux difference and restricts power levels to between APLND and either APLRAFD0 or 100% of RATED THERMAL POWER, whichever is less.
Prior to entering RESTRICTED AFD OPERATION, a 24-hour waiting period at a power level (12%) above APLND and below that allowed by Nonnal Operation is necessary.
During this time period load changes and control rod motion are restricted to that allowed by the RESTRICTED AFD OPERATION procedure. After the waiting period, RESTRICTED AFD OPERATION is permitted.
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 will 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 reouced significance.
Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitor Alarm. The computer deter-mines 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 H0T 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 desi minimum DNBR are not exceeded, and 2) gn limits on peak local power density and in the event of a LOCA the peak fuel clad temperature will not exceed the 2200*F ECCS acceptance criteria limit.
I CALLAWAY - UNIT 1 B 3/4 2-2 Amendment No. JE, 28 l
I
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1 1 i
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+20%
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l INDICATED AXtAL PLUX OlFFERENCE i
FIGURE S 3/4.2-1 TYPICAL INDICATED AXIAL FLUX DIFFERENCE VERSUS THERMAL POWER CALLAWAY - UNIT 1 3 3/4 2-3 Amendment No. 15 L =
POWER DISTRIBUTION LIMITS BASES 1
3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR AND NUCLEAR ENTHALPY RISE h0T CHANNEL FACTOR (Continued)
Each of these'is measurable but will normally only be determined period-l ically as specified in Specifications 4.2.2 and 4.2.3.
This periodic surveil-lance is sufficient to ensure that the limits are maintained provided:
Control rods in a single group move together with no individual a.
rod insertion differing by more than + 12 steps, indicated, from the group demand position, b.
Control rod banks are sequenced with overlapping groups as described in Specification 3.1.3.6.
The control rod insertion limits of Specification 3.1.3.6 are c.
maintained.
d.
The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.
N aboveaNmaintained.willbemaintainedwithinitsligitsprovidedconditionsa.throughd.
F The relaxation of F as a function of THERMAL POWER allows changes in the radial power shape fh all permissible rod insertion limits.
When an Fg measurement is taken, an allowance for both experimental error 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.
N When F is measured, (i.e., inferred), no additional allowances are necessaryprNrtocomparisonwiththelimitsofSection3.2.3.
An error allow-ance of 4% has been included in the limits of Section 3.2.3.
Margin between the safety analysis DNBR limits (1.42 and 1.45 for the Optimized fuel thimble and typical cells, respectively, and 1.61 and 1.69 for the VANTAGE 5 thimble and typical cells) and the design DNBR limits (1.32 and 1.34 for the Optimized fuel thimble and typical cells and 1.32 and 1.33 for the VANTAGE 5 thimble and typical cells, respectively) is maintained. A fraction of this margin is utilized to accommodate the transition core DNBR penalty (2% for Optimized fuel,12h% for. VANTAGE 5 fuel) and the appropriate fuel rod bow DNBR penalty (less than 1.5% per WCAP-8691, Rev.1). The margin between design and safety analysis DNBR limits of 7% ft ?ptimized fuel and 18% fdr VANTAGE 5 fuel includes greater than 3% margin (
Optimized fuel and 4%
margin for VANTAGE 5 fuel for plant design flexibility.
The hot channel factor Fh(z) is measured periodically and increased by a cycle and height dependent power factor appropriate to either Normal Operation or RESTRICTED AFD OPERATION, W(z)h(or W(z)E00,W(z)N0 to provide assurance that the limit on the hot channel factor, z),is CALLAWAY - UNIT 1 B 3/4 2-4 Amendment No. JE, 28
a 1
l l
l 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) of normal operation transients and was determined from expected power control maneuvers over the full range of burnup conditions in the core.
W(z)ggpon accounts for the more restrictive operating limits required by RESTRICTED AFD OPERATION which result in less severe transient values. The W(z) functions are provided in the Peaking Factor Limit Report per Specification 6.9.1.9.
Provisions to account for the possibility of decreases in margin to the FQ(z) limit during intervals between surveillance are provided. Any decrease in the minimum margin to the F (z) limit compared to the minimum margin determined 0
from the previous flux map is determined by comparing the ratio of:
maximum F((z) over z K(z) taken from the current map to the same ratio from the previous map.
The ratios to be compared from the two flux maps do not need to be calculated at identical z locations.
Increases in this ratio indicate that the minimum margin to the F (z) limit has decreased and that additional penalties must be applied to the Q
measured F (z) to account for further decreases in margin that could occur Q
before the next surveillance. More frequent surveillance may also be substi-tuted for the additional penalty.
3/4.2.4 QUADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that the radial power distri-bution 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 associ-ated with the indicated power tilt.
The 2-hour time.. allowance forr. 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 misal.ign.ed control rod.
In the event such action does not correct the tilt,-the margio'for. uncertainty.on F :~is-re' instated by reducing Q
the maximum allowed power :by 3% forceach. 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 symetric thimbles. The two sets of four symmet-ric thimbles is a unique set of eight detector locations. These locations are C-8 E-5, E-ll, H-3, H-13, L-5, L-11, N-8.
CALLAWAY - UNIT 1 B 3/4 2-5 Amendment No. JE,28
=
_ = _ _
POWER DISTRIBUTION LIMITS BASES-3/4.2.5 DNB PARAMETERS The limits on the DNB-related parameters assuro that each of the param-eters 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 indicated Tavg value of 592.6*F and the indicated pressurizer pressure value of 2220 psig correspond to analytical limits of 595.2 F and 2202 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 uncer-tainty 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 measurement uncertainty for the RCS total flow rate is based upon perform-ing_a precision 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, j
i I
j
'CALLAWAY - UNIT 1 B 3/4 2-6 Amendment No. J5, 28
t e
ADMINISTRATIVE CONTROLS PEAKING FACTOR LIMIT REPORT value for APL D ()as required) shall be established for at least each reload 6.9.1.9 The W(z functions for Normal and RESTRICTED AFD OPERATION and the H
core and shall be maintained available in the Control Room. The limits shall be established and implemented on a time scale consistent with normal proce-dural changes.
The analytical methods used to generate the W(z) functions and APLND shall be those previously reviewed and approved by the NRC*.
If changes to these l
methods are deemed necessary, they will be evaluated in accordance with 10 CFR 50.59 and submitted to the NRC for review and approval-prior to their use if the change is determined to involve an unreviewed safety question or if such a change would require amendment of previously submitted documentation.
A report containing the W(z) functions, as a function of core height (and burnup, if applicable) and APLND shall be provided to the NRC Document Control Desk with copies to the Regional Administrator and the Resident Inspector within 30 days of their implementation.
SPECIAL REPORTS 6.9.2 Special Reports shall be submitted to the Regional Administrator of the NRC Regional Office within the time period specified for each report.
6.10 RECORD RETENTION In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following recorda shall be retained for at least the minimum period indicated.
6 W.1 The following records shall be retained for at least 5 years:
a.
Records and logs of unit operation covering time interval at each power level; b.
Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety;
{
c.
All REPORTABLE EVENTS; j
d.
Records of surveillance activities; inspections and calibrations l
required by these Technical Specifications; l
~~WCAP-8385', " Power Distribution Control-and Load Following Procedures,"~~
WCAP-9272-A, " Westinghouse Reload Safety Evaluation Methodology," and WCAP-19216-P-A, " Relaxation of Constant Axial Offset Control / FQ Surveillance Technical Specification."
CALLAWAY - UNIT 1 6-21 Amendment No.
28
l 1
ADMINISTRATIVE CONTROLS
_RECORDRETENTION(Continued) e.
Records of changes made to the procedures required by Specification l
6.8.1; i
f.
Records of radioactive shipments; g.
Records of sealed source and fission detector leak tests and results; and h.
Records of annual physical inventory of all sealed source material i
of record.
6.10.2 The following records shall be retained for the duration of the unit Operating License:
a.
Records and drawing changes reflecting unit design modifications made to systems and equipment described in the Final Safety Analysis Report; b.
Records of new and irradiated fuel inventory, fuel transfers and assembly burnup histories; c.
Records of radiation exposure for all individuals entering radiation control areas; d.
Records of gaseous and liquid radioactive material released to the environs; Records of transient or operational cycles for those unit components e.
identified in Table 5.7-1; f.
Records of reactor tests and experiments; g.
Records of training and qualification for current members of the unit staff; h.
Records of in-service inspections performed pursuant to these Technical Specifications; i.
Records of quality assurance activities required by the~QA Program; j.
Records of reviews performed for changes made to procedures or f
equipment or reviews of tests and experiments pursuant to j
10 CFR 50.59; I
k.
Records of meetings of the ORC and the NSRB; 4
1.
Records of the service lives of all hydraulic and mechanical snubbers required by Specification 3.7.8 including the date at which the service life commences and associated installation and maintenance records; m.
Records of secondary water sampling and water quality; and n.
Records of analysis required by the Radiological Environmental 1
Monitoring Program that would permit evaluation of the accuracy j
of the analysis at a later date. This should include procedures effective at specified times and QA records showing that these procedures were followed.
CALLAWAY - UNIT 1 6-22 Amendment No.
28
_ _ _ _ _ -

ML20236A687

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