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Marked-up Proposed Tech Specs Re Slave Relay Test,Staggered Test Basis,Thermal Power,Trip Actuating Device Operational Test,Unidentified Leakage,Unrestricted Area,Venting,Waste Gas Holdup Sys,Frequency Notation & Operational Modes
	ML20244D810
Person / Time
Site:	Comanche Peak
Issue date:	04/14/1989
From:	; TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC)
To:	;
Shared Package
ML20244D809	List: ML20244D805; ML20244D810;
References
	NUDOCS 8904240123
	Download: ML20244D810 (74)
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, INDEX DEFIN!T!ONS.

SECTION ,

PAGE 1.32 SLAVE RELAY T'fST.............................................. 'l-6 1/33 500 DJFfC AT I O d . . . T. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-b e dJ4!

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1. % STAGGERED TEST BASIS.............

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. . . . ., A 1-6 1.J63YTHERMAL 36 P0WER................................................. 1-6 1.J7 TRIP ACTUATING DEVICE OPERATIONAL TE5T........................ 1-6 35 1.34f UNIDENTIFIED 3

LEAKAGE.......................................... 1-7

1. 8 ' UNRESTRICTED AREA............................................. 1-7 1.# VENTING....................................................... 1-7 .

1.M7 WASTE GAS HOLDUP 5YSTEM....................................... 1-7 TABLE 1.1 FREQUENCY N0TATION...................................... 1-8 TABLE 1.2 OPERATIONAL M0 DES....................................... 1-9 COMANCHE PEAK - UNIT 1 ii

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, INDEX L1MXT!NG CONDITXONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE TABLE 3.3-7 EXPLOSTVE GASEOUS MONITORING INSTRUMENTATION.......... 3/4 3-47 3

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3/4.3.4 TURBINE OVERSPEED PROTECTION.............................. 3/4 3-51 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION Startup and Power Operation.............................. 3/4 4-1 Hot Standby.............................................. 3/4 4-2 .

Hot Shutdown............................................. 3/4 4-4 Col d Shutdown - Loops F i l l ed. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-6 Cold Shutdown - Loops Not filled......................... 3/4 4-7 3/4.4.2 SAFETY VALVES Shutdown............................................... 3/4 4-8 0perating.............................................. 3/4 4-9 3/4.4.3 PRESSURIZER.............................................. 3/4 4-10 3/4.4.4 RELIEF VALVES............................................ 3/4 4-11 3/4.4.5 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems................................ 3/4 4-13 Operational Leakage...................................... 3/4 4-14 TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES...... 3/4 4-16 3/4.4.6 CHEMISTRY................................................ '/4 4-17 TABLE 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS............... 3/4 4 J .

3/4.4,7 SPECIFIC ACTIVITY.. ..................................... 3/4 4-19 FIGURE 3.4-1 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY > 1pCi/ gram DOSE EQUIVALENT I-131.................................... 3/4 4-20 TABLE 4.4-1 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM.................................................. 3/4 4-21 COMANCHE PEAK - UNIT 1 vi

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.- DEFINITIONS CONTAINMENT INTEGRITY p

b 1.7 CONTAINMENT INTEGRITY shall exist when: ,

a. All penetrations required to be closed during accident conditions I 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 csitiens. Wives- th&t Gre_ . TS regJired to be Open to performisur-veillance-test-or for norW 99-060 p.lant tvolutions smy be vpene&ctr arr-intermittent hath ed=4mistrative centrols, oli+"^% e mP* **

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b. All equipment hatches are closed and sealed, +k #"M ""D#'" # .

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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, be' lows, or 0-rings) is 4PERABLE..

CONTROLLED LEAKAGE 1.8 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals.

CORE ALTERATIONS 1.9 CORE ALTERATIONS shall be the movement or manipulation of any component within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe conservative position.

DIGITAL CHANNEL OPERATIONAL TEST 1.10 A DIGITAL CHANNEL OPERATIONAL TEST shall consist of exercising the digital computer hardware using data base manipulation and injecting simulated process data to verify OPERABILITY of alarm and/or trip functions.

005E EOUIVALENT I-131 1.11 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie / gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The tnyroid 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 O Reactor Sites" or Table E-7 of NRC Regulatory Guide 1.109, Revision 1, October 1977.

COMANCHE PEAK - UNIT 1 1-2

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DEFINITIONS MEMBER (5) 0F THE PUBLIC V 1.17 MEMBER (5) 0F THE PUBLIC shall include all persons who are not occupa-tionally associate,d,with the plant. This category does not include employees i of the licensee, its contractors, or vendors. Also excluded from this category l 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 recre-l ational, occupational, or other purposes not associated with the plant.

OFFSITE DOSE CALCULATION MANUAL 1.1B The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radio-active gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm / Trip 5etpoints, and in the conduct of the Environ-mental Radiological Monitorigt am. The ODCM shall also contain (1) the Radioactive Effluent Qedrols and R diological Environmental Monitoring Pro- p grams required by Sectio

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and descriptions of the~information that shouldbeincludedin\he nnual Radio gical Environmental Operating and Semi- 91 # 0 annualRadioactiveEff{uentReleaseReprtsrequiredbySpecifications6.9.1.3 and 6.9.1.4. 6.7. 3 OPERABLE - OPERABILITY 1.19 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s),

and when all necessary attendant instrumentation, controls, electrical power, (Vfm) 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 MDCE - MODE 1.20 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive 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 nuclear characteristics of the reactor core and related instrumentation: !

(1) described in Chapter 14.0 of the FSAR, (2) authorized under the provisions of 10 CFR 50.59, or (3) otherwise approved by the Commission.

PRESSURE BOUNDARY LEAKAGE 1.22 PRES 5URE BOUNDARY LEAKAGE shall be leakage (except steam generator tube leakage) through a nonisolable fault in a Reactor Coolant System component l body, pipe wall, or vessel wall.

1 CDMAN^HE PEAK - UNIT 1 1-4

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1 Page 12 of 81 DEFINITIONS l

PRIMARY PLANT VENTILATION SYSTEM

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1.23 A PRIMARY PLANT VENTILATION SYSTEM shall be any system designed and V installed to reduce gaseous radioiodine 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 particulate 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.

PROCESS CONTROL PROGRAM 1.24 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, tests, and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated IS processing of actual or simulated wet solid wastes will be accomp1' sped f9-ok0 such a way as to assure compliance with 10 CFR Parts 20, 61, and ,

State regulations, round requirements, and other requirem n & g6 ern- .

ing the disposal f radioac 've waste.

PURGE - PURGING Sold 1.25 PURGE or PURGING shall be any controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

QUADRANT POWER TILT RATIO (O) ' ^ '

1.26 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper half excore detector calibrated output to the average of the upper half excore detector calibrated outputs, or the ratio of the maximum lower half excore j

detector calibrated output to the average of the lower half excore detector j calibrated outputs, whichever is greater. With one excore detector inoperable, I the remaining three detectors shall be used for computing the average.

RATED THERMAL POWER 1.27 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3411 KWt.

REACTOR TRIP SYSTEM RESPONSE TIME 1.28 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its Trip 5etpoint at the channel sensor until loss of stationary gripper coil voltage.

REPCRTABLE EVENT 1.29 A REPORTABLE EVENT shall be any of those conditions specified in 10 CFR 50.73.

COMANCHE PEAK - UNIT I 1-5

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DEFINITIONS p SHUTDOWN MARGIN-U 1.30 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subaritical or would be subcritical from its present condition assuming all 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.

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NW 1.31 The SITE B0UNDARY shall be that line bey:nd eH ch the End--is neither o cen:d, im 'ea:cd, nor-ether iiw ce e ^^d h" "" **"a" -

SLAVE RELAY TEST 1.32 A SLAVE RELAY TEST shall be the energization of each slave relay and verification of OPERABILITY of each relay. The SLAVE RELAY. TEST shall include -

a continuity check, as a minimum, of associated testable actuation devices. 7

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1.33 5 4DIFICATION sh fl be the conver) on of wet waste into a form t at

/ n gg meets ipping and bu- al ground requi ements.

50M CE CHECK

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1.34 A SOURC CHECK shall bs .e qualitative p sessment of chafinel response O' uwhen

_ the cha el sensor is e osed to a soured of increased raffloactiv' STAGGERED TEST BASIS 4 A STAGGERED TEST BASIS shall consist of:

33

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. d # HERMAL POWER shall be the total core heat transfer rate to the reactor co nt.

TRR-ACTUATING DEVICE OPERATIONAL TEST A TRIP ACTUATING DEVICE OPERATIONAL TEST shall consist of operating the T p 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.

COMANCHE PEAK - UNIT 1 1-6

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-vmm v Page 14 of 81 DEFINITIONS UNIDENTIFIED LEAKAGE

1. UNIDENTIFIED LEAKAGE shall be all leakege which is not IDENTIFIED LEAKAGE or CONTROLLED LEAKAGE.

UNRESTRICTED AREA

,g33 1.PJ An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUN forMrposesofprotectionofindividualsfromexposuretoradiationand f radioactive materials, or any area within the SITE BOUNDAR laccess to which]

{is not controlled by the licensee ] a#Used for residential quarters or for ,

induu riai, commercial, institutional, and/or recreational purposes.

VENTING

1. k VENTING shall be the controlled process of discharging air or gas from a 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.

WASTE GAS HOLDUP SYSTEM

1. A WASTE GAS HOLDUP SYSTEM shall be any system designed and installed to reduce radioactive gaseous effluents by collecting Reactor Coolant System offgases from the Reactor Coolant System and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the O environment.

O COMANCHE PEAK - UNIT 1 1-7

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.y[ge 15~of S 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.

SR At least once per 9 months.

R At least once per 18 months.

S/U Prior to each reactor startup.

N. A. Not applicable. 75

, Completed prior to each release. O9 -07 5 O

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COMANCHE PEAK - UNIT 1 1-8 l

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--- v Page 21 of 81 SAFETY LIMITS BASES 2.1.2 REACTOR C00LhNT 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, valves and fittings 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 73 criteria and associated Code requirements.

3 The entire RCS is hydrotested at 12 % (3110 psig) of design pressure, to demonstrate integrity prior to initial ope

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O COMAN:HE PEAK - UNIT 1 B 2-2

. _ a nw v umge 22 O V Bil i

APPLICABILITY l

SURVEILLANCE REQUIREMENTS (Continued) 4.0.6.3 Inspection Frequencies - The above required inservice inspections of steam generator tubes shall be performed at the following frequencies:

a. The first inservice inspection shall be performed after 6 Effective Full Power Months (EFPM) and before 12 EFPM and shall include a special inspection of all expanded tubes in all steam generators.

' Subsequent inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection. If two consecutive inspections, not including the preser-vice inspection, result in all inspection results falling into the C-1 category or if two consecutive inspections demonstrate that previously obser'.ed degradation has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of .

Once per 40 months;

b. If the results of the inservice inspection of a steam generator conducted in accordance with Table 4.0-2 at 40-month intervals fall in Category C-3, the inspection frequency shall be increased to at least once per 20 months. The increase in inspection frequency shall apply until the subsequent inspections satisfy the criteria of Specification 4.0.6.3a.; the interval may then be extended to a maximum of once per 40 months; and k c. Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection specified in Table 4.0-2 during the shutdown subsequent to any of the following conditions: TS ff9-

1) Primary-to secondary tubes leak (not including leaks originating o75 from tube-to-tub Specification . p t h ds) in excess of the limits of V, L2, 3.orv. s. 1

2) A seismic occurre Q r t n the Operating Basis Earthquake, or

3) A loss-of-coolant accident requiring actuation of the Engineered 4

Safety Features, or

4) A main steam line or feedwater line break.

O COMANCHE PEAK - UNIT I 3/4 0-5

Attachment to TXX-89169 g;;'2 24;,2 9r F/ArAt. DRArr APPLICABILITY n

() SURVEILLANCE REQUIREMENTS (Continued) 4.0.6.4 Acceptance Criteria

a. As used in this specification:

1) Imperfection means an exception to the dimensions, finish, or contour of a tube from that required by fabrication drawings or specifications. Eddy-current testing indications below 20% of the nominal tube wall thickness, if detectable, may be considered as imperfections;

2) Degradation means a service-induced cracking, wastage, wear, or general corrosion occurring on either inside.or outside of a tube; -

3) Degraded Tube means a tube containing imperfections greater than or equal to 20% of the nominal wall thickness caused by degradation;

4) % Degradation means the percentage of the tube wall thickness affected or removed by degradation;

( 5) Defect means an imperfection of such severity that it exceeds

( tne plugging limit. A tube containing a defect is defective;

6) Plugging Limit means the imperfection depth at or beyond which rs tne tube snall be removed from service and is equal to 40% W H-o n of the nominal tube wall thickness;

7) Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integ-rity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in Specification 4.0.6.3c., above;

8) Tube Insoection means an inspection of the steam generator tube from the point of entry (hot leg side) completely around the U-bend to the top support of the cold leg; and

_ __. rs

Value to be determined in accordance with the recommendations of Regulatory ## 7 6 t Guide 1.121, August 1976. j 1

I r

(.

(

l COMANCHE PEAK - UN:T 1 3/4 0-6 i

__-______-____-___L

p 1 9 l Page 24 of 81 l . i REACTIVITY CONTROL SYSTEMS l

,m CHARGING PUMPS - OPERATING

()

LIMITING CONDITION.FOR OPERATION rs !

3.1.2.4 i g pumps shall be OPERABLE. ype;5 Atleasttwocentrf[uplc APPLICABILITY: MODES 1, 2 ad-ACTION:

With only one charging pump OPERABLE, restore at least two charging pumps to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least 1% ak/k at 200*F within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore at least two charging pumps to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . ,

SURVEILLANCE REQUIREMENTS 4.1.2.4.1 The required centrifugal charging pump (s) shall be demonstrated OPERABLE by testing pursuant to Specification 4.0.5.

4.1.2.4.2 The required positive displacement charging pump shall be demonstrated OPERABLE by testing pursuant to specification 4.1.2.2.c.

('~_}

C 4.1.2.4.3 Whenever the temperature of one or more of the Reactor Coolant System (RCS) cold legs is less than or equal to 350*F, a maximum of two charging pumps shall be OPERABLE, except as allowed by Specification 3.4.8.3.

When required, one charging pump shall be demonstrated inoperable

at least once per 31 days by verifying that the motor circuit breakers are secured in the open position.

An inoperable pump may be energized for testing provided the discharge of the pump has been isolated from the RCS by a closed isolation valve (s) with power removed from the valve operator (s) or by a manual isolation valve (s) secured in the closed position.

- The provisions of Specification 3.0.4 and 4.0.4 are not applicable for entry into MODES 3 and 4 for the charging pump declared inoperable pursuant to Specification 3.1.2.4 provided the charging pump is restored to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after entering MODE 3 or prior to the temperature of one or more of the RCS cold legs exceeding 375*F, whichever comes first.

In Mode 4 the positive displacement pump may be used in lieu of one of the required centrifugal charging pumps.

COMANCHE PEAK - UNIT 1 3/4 1-10

Attachment to TXX-89169 gzn,w F l e t. t. P e t e r ,

TABLE 4.3-1 (Continued)

TABLE NOTATIONS a

0nly if the reactor trip breakers happen to be in the clorsed ard the Control Rod Drive System is capable of rod withdrawal.

b Below P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.

c Below P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.

d Above the P-7 (At Power) Setpoint.

l (1) If not performed in previous 31 days.

(2) Comparison of calorimetric to excore power and N-16 power indication above 15% of RATED. THERMAL POWER. Adjust excore channel and/or N-16 channel gains consistent with calorimetric power if absolute difference of the respective channel is greater than 2%. The provisions of Specification 4.0.4 are not -

applicable to entry into MODE 1 or 2.

(3) Single point comparison of incore to excore AXIAL FLUX DIFFERENCE l above 15% of RATED THERMAL POWER. Recalibrates if the absolute '

difference is greater than or equal to 3%. For the purpose of these surveillance requirements "M", is defined as at least once per 31 EFPD.

The provisions of Specification 4.0.4 are not applicable for entry into MODE 1 or 2.

(4) Neutron and N-16 detectors may be excluded from CHANNEL CALIBRATION.

(5) Detector plateau curves shall be obtained and evaluated. For the Intermediate Range Neutron Flux, Power Range Neutron Fluy and N-16 channels the provisions of Specification 4.0.4 are not applicable for entry into MODE 1 or 2.

(6) Incore - Excore Calibration, above 75% of RATED THERMAL POWER. For the purpose of these surveillance requirements "Q" is defined as at least once per 92 EFPD. The provisions of Specification 4.0.4 are not applic-able for entry into MODE 1 or 2.

(7) Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS.

(8) The surveillance frequency and/or MODES specified for these channels in Table 4.3-2 are more restrictive and therefore applicable.

(9) Quarterly surveillance in MODESa 3 , a4 , and aS shall also include verifica-tion that permissives P-6 and P-10 are in their required state for exist-ing plant conditions by observation of the permissive 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 T-5 rate within a 10-minute period. The fro Mene of Spe d Oco.+;oa 8.o v gs-oss anc. pot s pic n ble. be up +o 1A k eso la llewig w re nter ge:p ,

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TABLE 3.3-2 (Continued)

TABLE NOTATIONS C- a Trip function may be blocked in this MODE below the P-11 (Pressurizer Pressure Interlock) Setpoint. '

b Trip function automatically blocked above P-11 and may be unblocked below P-11 by blocking the Safety Injection on low steam line pressure.

c Not applicable if each affected main steam isolation valve and its associated upstream drain pot isolation valve per steam line is closed.

d The provisions of Specification 4.0.4 are not applicable for entry into MODE 3. i

'The channel which provides a steam generator water level control signal (if one of three specific trip channels is selected to provide input into steam generator water level control) must be placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and maintained in the tripped condition with the exception that the channel may be taken out of the tripped condition for up to'2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to allow ~

testing of redundant channels.

I Not applicable if Preferred Offsite Source Breaker is open, n on ,rypinsh ' C +A

a d"9 Ocd~ke N h Jul"<d ;~v~bh ersn a +.

p.oga t ACTION STATEMENTS Sg.sta J. % /. 2 ACTION 12 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, 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 COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ; however, one channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.2.1, provided the other channel is OPERABLE.

ACTION 13 - With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed until performance of the next required ANALOG CHANNEL OPERATIONAL TEST provided the inoperable channel is placed in the tripped condition within '

I hour.

ACTION 14 - With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed provided the inoper-able channel is placed in the bypassed condition and the Minimum 7hannels OPERABLE requirement is met. One additional channel 1 may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.2.1.

ACTION 15 - With less than the Minimum Channels OPERABLE requirement, opera-tion may continue provided the containment pressure relief valves are closed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and maintained closed.

ACTION 16 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel

( to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 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 COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

COMANCHE PEAK - UNIT 1 3 3/4 3-22 4

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+o c ' eg nc l" . S a '@ +rm -

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s wn o+

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c,- %I. s Ynk%

g y

w. N O

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C 88 6

3 Tg 7

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A 22 2 2 P 8 I -

RT 0 TN 1

/I l MO xm RP . . * /

AT 2i S LE A A.

. .C N AS NN O

I g T

A E 44 R L E B , , ,

P A 33 3 O C

I S , , ,* '

T LE 22 26 N PD l A PO , , ,, l L AM 11 14 A P

R O SE e F MLL k UEB a N MNA t O I NR n I NAEl i

4 T I iP /

- A MCO II 1 2 3 T N

3 E M M C U R L

l R A l T L A S A T N S/ e I LP k EI a G NR t N NT . . n I A i R l i0 A. A. /

O C1 NN 1 1 T

I N

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N i O t I y a T t l A i o I v s l D i y I e A tt v R ci n e

. av o L n oi i y o it t t n i d c a i o t aa l v i c Ro i i t e i t t a t ed n c i e t a e a d D aR V o a l i m R e us t d o -

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. 1 I

C N

U F

R 1

C 2

G C 3

A l9" y>* ' - U* " e l

i Attachment to TXX-89169

~

ag 3 of TABLE 3.3-4 (Continued)

(~')

O TABLE NOTATIONS , y ,g n ,

Gaseous Effluear ke. K~h n #i'03g I

Must satisfy Spee4ficaticr. 3.1h2.1 requir=gA*cmes nt:. - <", 'r k e ODCM- 1 During CORE ALTERATIONS or movement of irradiated fuel within containment.

ACTION STATEMENTS i

ACTION 26 - With the number of OPERABLE channels less than the Minimum Channels OPERABLE requirement, operation may continue provided the containment ventilation valves are maintained closed. The D containment pressure relief valves may only be gpened_in com- "'#N l

__pliance with Specificat__ ion 3.6.1.7 and e.1 3._4_yche lla 4 4**

QC _ ur Et,Ttse s M.a m r. E ~ nr ~ e@. sea m o w n +s ;a Pu 1.1~ e 4 ACTION 27 - With the number of OPER LE channels one 1Ess trian the Minimum de-

"' Channels OPERABLE requirements, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . secure the Contro buA.

Room makeup air supply fan from the affected intake or initiate . .

and maintain operation of the Control Room Emergency Air Cleanup System in emergency r circulation.

ACTION 28 - With the number of OPERABLE channels less than the Minimum k

f'hi[

Channels OPERABLE requirement, comply with the ACTION require-ments of Specification 3.4.5.1.

E L i M

m, s,

n !

P.E!

i COMANCHE PEAK - UNIT 1 3/4 3-40 e

~~~

" ~ ~~ " "'^' ' "

. A'prII II,'~1989 Page'38 of 81

. INSTRUMENTATION EXPLOSIVE GASEOUS MONITORING INSTRUMENTATION LIMITING CONDIT1DN FOR OPERATION TS 7~ 99-3.3.3.4 The exp1 ive gaseous monitoring instrumentation channels shown in e75 Table T.3- shal be OPERABLE with their Alarm / Trip Setpoints set to ensure that the j ' of Specifications 3.11.2.1 are not exceeded.

APPLICABILITY: As shown in Table 3.3-7 ACTION:

al With an explosive gaseous monitoring instrumentation channel Alarm /

Trip Setpoint less conservative than required by the above specifi-cation, declare the channel inoperable and take the ACTION shown in Table 3.3-7. ,

b. With less than the minimum number of explosive gaseous monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3-7. Restore the inoperable instrumentation to OPERABLE status within 30 days and, if unsuccessful prepare and submit a special report to the Commission pursuant to Specification 6.9.2 to explain why this inoperability was not corrected in a timely manner.

, c. The provisions of Specifications 3.0.3 are not applicable.

l SURVEILLANCE REQUIREMENTS h sheaatoT3le3.1-]7/

4.3.3.4 Each explosive gaseous monitoring instrumentation c hall be demonstrated OPERABLE:by perference of tM CHANNEL CliECK, CP"NEL CALIBRATRN rs and ANAL-OfrCHA'NEL OPERATIONAL TEST--at the i, wquenc4es,iwwn in Table 4. 3-3. ep e A + least o"ce per 24 kc"'S by 1 c ' -lcr~ c e oC%

. C H AszNEL C Heck,

b. A 4 leas + ane e per 3/ Ap h y p er % ~m e ef "

ANA LCS C t/ANNEL OPERA r r d^1AL Tn G wol

c. A t leas + es<e pee 92 olays b y per krw 05 '&

CHA Nhl EL L A LI OR ATTON wC ck. s k II I"'l"'Y' #

of skhr) ga.s s u ples la u urdwe t "" *A ' '

M W $4 c fu !e P 'b reLom mea dwl,*rn,s.

COMANCHE PEAK - UNIT 1 3/4 3-46 I

(

9M Q? O L

A S LN T GEO N ONI T E LNTS i

AAAE M M E'

l NiRT R ACE I P U O Q

E R

.~ ~

E '

C N N A O L LI L ET I NA E NR ) )

V LAB 1 1 R i I ( (

U CL Q Q S A C

N m O I

3

= 3 T

A T

N 4 E

- M E U L R B T L A S EK L T N NC I NE D D Ai l G liC N C I

R O

T I

N O

M e

_ v

_ S i

_ U s

_ O o E l p

u S A

G x

- E V E I T m S S e s i

O Yt r L S s o s P y t r X PS i o E U n t D g o i L n M n 0i o l

I r n M e

St o. g n Ai o e G n r g o d y T EM ly x N T I O E S s M Aa . .

U WG a b 7 R T

S t .

I 1

[l' m g>* [xZ - ~> 8

- ll l

'AprIIT 5,'~1989

~

~~ ""*"

, , . , 1

.Page 40 of 81- TABLE 4.3-3 (Continued) -'

TABLE NOTATIONS

(; ~' (1) The CHANNEL CALIBRATION shall include the use of standard as samples in accordance with the manufacturer's recommendations -

TS p- 0, 5 .

9 e

I 4

i O UO I

~

O COMANCHE PEAK - UNIT 1 3/4 3-50

At t a c hm e nt- to T X X-3916 9 --. -

4 of gh h 1

O 75 8 9 - D 7 Ef e,fYM L -

g ;,1~ AbJ 800-780-760-

, 740-720-700- .

680-i /

g 660-

@ 640-E 620-

/' 2

's 600-580-560-540-520-500 ', l ; i ; i j ;

60 100 140 180 220 260 300 340 RCS TEMPERATURE, DEG F FIGUR 3.4-4. PORV SETPOINTS FOR OVERPRESSURE MITIGATION APPLICABLE UP T0,1F EFPY O

COMANCHE PEAK - UNIT 1 3/4 4-29

_ -a

-- ~ - - - -

Attachment to TXX-89169

. -April 14, 1989 Page 42 of 81-j I

4 geg lo um c4 war JG FIGURE 3.4 4 l

PORV Setpoints For Overpressure Mitigation Applicable Up to 16 EFPY

{

600 780 -

760 - ,

740 720 x

9 700 - -

$ \ '

,, 680 f

r.

660 O 640 m

g 620 2

600 +-

550

$60 -- t

$40 520 -

500 - ,

60 100 140 160 220 260 300 340 RCS TEMP [RATURE, Dtc r i

l Attachment to TXX-89169

~

ag 4 of 8 i REACTOR COOLANT SYSTEM

(_) 3/4.4.10 REACTOR COOLANT SYSTEM VENTS LIMITING CONDITION FOR OPERATION 3.4.10 At least one Reactor Coolant System vent path consisting of two vent i valves in series powered from emergency busses shall be OPERABLE and closed at l each of the following locations:

a. Reactor vessel head, and

b. Pressurizer steam space.

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

ACTION: ,

a. With one of the above Reactor Coolant System vent paths inoperable, STARTUP and/or POWER OPERATION may continue provided the inoperable vent path is maintained closed with power removed from the valve actuator of all the vent valves in the inoperable vent path; restore the inoperable vent to OPERABLE status within 30 days, or, be TS in HOT STANDBY wi-gn6h s and in COLD SHUTDOWN within the follow- g9 -o? .f) ing 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . 4g

b. With both Reactor oo an System vent paths inoperable; maintain the inoperable vent paths closed with power removed from the valve actua-tors of all the vent valves in the inoperable vent paths, and restore di. lec::t one of the vent paths to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT STANDBY wit n ifli6u and in COLD SHUTDOWN within the fol-lowing 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . ^

As neyr SURVEILLANCE REQUIREMENTS 4.4.10 Each Reactor Coolant System vent path shall be demonstrated OPERABLE at least once per 18 months by:

a. Verifying all manual isolation valves in each vent path are locked in the open position, -

l

b. Cycling each vent valve through at least one complete cycle of l full travel from the control room, and {

c. Verifying flow through the Reactor Coolant System vent paths during venting.

l J

COMANCHE PEAK - UNIT 1 3/4 4-32

_ _ - - - - . - _ _ - _ _ --- J

Attach ent to TXX-89169 _ _ . _ _ . . . _ _ _ . ... -

~ '

1 g 44 f8 hh [

l EMERGENCY CORE COOLING SYSTEMS

('h SURVEILLANCE REQUIREMENTS 4.5.2 Each ECCS subsystem 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 verifying that the following valves are in the indicated positions with power to the valve operators removed:

Valve Number Valve Function Valve Position 8802 A & B SI Pump to Hot Legs Closed ,

8808 A, B, C, D Accum. Discharge Open* l 8809 A & B RHR to Cold Legs Open

. 8835 SI Pump to Cold Legs Open 8840 RHR to Hot Legs Closed 8806 SI Pump Suction from RWST Open -

8813 SI Pump Mini-Flow Valve Open ,

b. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path that is i.n locked, sealed, or otherwise secured in position, is in its correct position.

c. By a visual inspection which verifies that no loose debris (rags, tresh, clothing, etc.) is present in the containment which could be transported to the containment sump and cause restriction of the pump p) suctions during LOCA conditions. This visual inspection shall be

(- performed:

1) For all accessible areas of the containment prior to establish-ing CONTAINMENT INTEGRITY, and

2) Of the areas affected within containment at the completion of each containment entry when CONTAINMENT' INTEGRITY is established.

d. At least once per 18 months by:

1) Verifying automatic isolation and interlock action of the RHR system from the Reactor Coolant System to ensure that:

a) With a simulated or actual Re oTiiitssyste pressure 75 signal greater than or equalf io psig the int ks 69- b5:L prevent the valves from beingma ene and y97 b) With a simulated or actual Reac orCoojant Sys. tem ssure I signal less than or equal to 750 psig'th'e' int 61ocks will j cause the valves to automatically close. !

(~'}

Surveillance Requirements covered in Specification 4.5.1.1.

'V COMANCHE PEAK - UNIT 1 3/4 5-4 j

.,r 1 9 Page 45 of 81 EMERGENCY CORE COOLING SYSTEMS a SURVEILLANCE REQUIREMENTS 4.5.3.1.1 The ECCS sub qstem shall be demonstrated OPERABLE per the applicable requirements of Sp e Mica 'on 4.5.2.

'b TS 4.5.3.1.2 A m ximu of tw charging pumps shall be OPERABLE except as allowed p-oS9 by Specificati n . 8.,3 When required, one charging pump shall be demon-strated inoper b e ' verifying that the motor circuit breaker is securr;d in the open positib ithin 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after entering MODE 4 from MODE 3 or prior to the temperature of one or more of the RCS cold leg: decreasing below 325 F, whichever occurs first, and at least once per 31 days thereafter.

\

l O \

l l

An inoperable pump may be energized for testing provided the discharge of the pump has been isolated from the RCS by a closed isolation valve (s) with power removed from the valve operator (s) or by a manual isolation valve (s) secured in the closed position.

COMANCHE PEAK - UNIT 1 3/4 5-8

1 Attachment to TXX-89169 Nh I

* - April 1.4 , 1 9 8 9 Page 46 of 81 CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

b. If any periodic Type A test fails to mcet either 0.75 L, or 0.75 Lg ,

the test schedule for subsequent Type A tests shall be reviewed and approved by the Commission. If two conse'cutive Type A tests fail to meet either 0.75 L, or 0.75 Lt, a Type A test shall be performed at least every 18 months until two consecutive Type A tests meet either 0.75 L, or 0.75 Lt at which time the above test schedule may be resumed;

c. The accuracy of each Type A test shall be verified by a supplemental I test which:

1) Confirms the accuracy of the test by verifying that the supple- l mental test result, L c, is in accordance with the appropriate '

. following equation:

lL c- (L,, + L,) l $ 0.25 La or l Lc- (Lg + L )g l 1 0.25 Lt '

where L,,or Ltm is the measured Type A test leakage and L, ,

t is the superimposed leak;

2) Has a duration sufficient to establish accurately the change in

,eb leakage rate between the Type A test and the supplemental test;

{ and 7 3) Requires that the rate at which gas is injected into the contain-

$ ment or bled from the containment during the supplemental test s is between 0.75 L, and 1.25'L,; or 0.75 Lt and 1.25 Lt '

dd. Type B and C tests shall be conducted with gas at a pressure not t less than P,, 48.3 psig, at intervals no greater than 24 months J except for tests involving:

/) 1) Air locks, E 2) Containment ventilation isolation valves with resilient material j seals, Safety Injection Valves as specified in Specification 4.6.1.2.g, jq 3) and f

s

4) Containment Spray Valves as specified in Specification 4.6.1.2.h.

e Air locks shall be tested and demonstrated OPERABLE by the require-

/ )I.

g ments vf Specification 4.6.1.3;

, f. Containment ventilation isolation valves with resilient material e

( seals shall be tested and demonstrated OPERABLE by the requirements of Specification 4.6.1.7.2 or 4.6.1.7.3, as applicable;

g. Safety Injection Valves 1-8802A, 1-8802B, 1-8809A, 1-88098, 1-8818A,

'{ 1- 88188 , 1- 8818C , 1- 8818D , 1- 8819 A , 1- 88198 , 1- 8819C , 1- 88190, 1- 8835 ,

g 1-8840, 1-8841A, 1-8841B, 1-8905A, 1-8905B, 1-8905C, and 1-8905D shall Wo76 be leak teste psig,atinte$withwateratapressurenotlessthan1.1Pa,53.13 rvals no greater than 24 months;

h. Containment Spray Valves 1HV-4776, 1HV-4777, ICT-142, and ICT-145 shall be leak tested with water, at a pressure not less than 1.1 Pa, 53.13 psig, at intervals no greater than 24 months; and

i. The provisions of Specification 4.0.2 are not applicable.

COMANCHE PEAK - UNIT 1 3/4 6-3

Attachment tn TXX-89169 N

i

. April 14, 1989 ,

Page 47 of 81 I CONTAINMENT SYSTEMS 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS i

CONTAINMENT SPRAY SYSTEM l

LIMITING CONDITION FOR OPERATION I

3.6.2.1 Two independent Containment Spray Systems shall be OPERABLE with each Spray System capable of taking suction from the RWST and manually transferring suction to the containment sump.

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

ACTION:

With one Containment Spray System inoperable, restore the inoperable Containment Spray System to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 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 ; restore the inoperable Containment Spray System to OPERABLE status within the next 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS

~

4.6.2.1 Each Containment Spray, System shall be demonstrated OPERABLE:

a. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position;

b. By verifying that in the test mode each train provides a total dis- l charge flow through the test header of greater than or equal to l 6600 gpm at 245 psig with the pump eductor line open when tested pursuant to Specification 4.0.5; i

l

c. At least once per 18 months during shutdown, by:

1) Verifying that each automatic valve in the flow path actuates to its correct position on a Containment Spray Actuation test signal, and 75

2) Verifying that each spray, et automatically on a Containment Spray Actu6 tion Safety n' ction test signal. N-076

( Qe < smut una n 9

d. At least once per 5 yearPhyaerformingan-41 or smoke flow test through each spray header and verifying each spray nozzle is unobstructed.

/

COMANCHE PEAK - UNIT 1 3/4 6-11

Attachment to TXX-89169 [

. . . April 14, 1989 I Page 48 of 81 PLANT SYSTEMS

() AUXILIARY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2 At least three independent steam generator auxiliary feedwater pumps and associated flow paths shall be OPERABLE with:

a. Two motor-driven auxiliary feedwater pumps, each capable of being 1 powered from separate emergency busses, and !

b: One steam turbine-driven auxiliary feedwater pump capable of being powered from two OPERABLE steam supplies.

APPLICABILITY: MODES 1, 2, and 3. -

ACTION:

a. With one auxiliary feedwater pump or associated flow path inoper-able, restore the required auxiliary feedwater pump or associated flow path to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 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 .

7 V b. With two auxiliary feedwater pumps or associated flow paths inoper-able, be in at least HOT STANDBY witin 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in H0T SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c. With three auxiliary feedwater pumps or associated flow paths inop-erable, immediately initiate corrective action to restore at least one auxiliary feedwater pump or associated flow path to OPERABLE s.tatus as soon as possible.

d. With only one OPERABLE steam supply system capable of providing power to the turbine-driven auxiliary feedwater pump, restore the required OPERABLE steam supplies within 7 days 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 .

SURVEILLANCE REQUIREMENTS 4.7.1.2 Each auxiliary feedwater pump and associated flow path shall be demonstrated OPERABLE:

a. At least once per 31 days on a STAGGERED TEST BASIS b . D Hu.A0 84-oM ,

1) Verifying that each motor-driven pump develop a schErgw 7 pressure of greater than or equal to 1372 psi at a flow of (d greater than or equal to 430 gpm; j

COMANCHE PEAK - UNIT 1 3/4 7-3

Attachment to TXX-89169 9 of 1 PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

2) Verifying that the steam turbine-driven pump develops a pressure of greater than or equal to 1338 psid at a flow of greater than or equal to 860 gpm when the secondary steam supply pressure is greater than 532 psig. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3;

3) Verifying that each non-automatic valve in the flow path that is not locked, sealed, or otherwise secured in position is in , ,3 its correct pos tion; and

4) Verifying t eac5 etic valve in the flow path is in the M-0D fully open position whenever the Auxiliary Feedwater System is in standby for auxiliary feedwater automatic initiation or when above 10% RATED THERMAL POWER. ,

i

b. At least once per 18 months during shutdown by: l l

1) Verifying that each automatic valve in the flow path actuates l to its correct position upon receipt of an Auxiliary Feedwater .

j Actuation test signal, and !

2) Verifying that each auxiliary feedwater pump starts as designed automatically upon receipt of an Auxiliary Feedwater Actuation test signal. The provisions of Specification 4.0.4 are not applicable to the turbine driven auxiliary feedwater pump for entry into Mode 3.

COMANCHE PEAK - UNIT 1 3/4 7-4

Attachment to TXX-89169 l a 5 of M o PLANT SYSTEMS

~

3/4 . 7. 4 STATION SERVICE WATER SYSTEM LIMITING CONDITION F"0R OPERATION 3.7.4 At least two independent service water loops shall be OPERABLE.

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

ACTION:

With only one service water loop OPERABLE, restore at least two loops to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 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 COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.7.4 Each service water loop shall be demonstrated OPERABLE:

a. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) servicing safety-related equipment that O is not locked, sealed, or otherwise secured in position is in its correct position; and

b. Atleastonceper18monthsduringshutdown,byverifyingthat[ d

-1) IglLautomatfe-va4ve-s ervicing-s afetyN elated -equipment-asteatW 8')-07 8 to its correc4.-position-ora-Safety-Inject 4opct signal. 2n W ach station service water pump starts automatically on a Safety Injection test signal.

O COMANCHE PEAK - UNIT 1 3/4 7-14

Attachment to TXX-89169 l

- . April 14, 1989 Page 51 of 81 mhh q

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

b. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemir- release in any ventilation zone communi-cating with the system .,,:

1) Verifying that the filtration unit satisfies the in place pene-tration and bypass leakage testing acceptance criteria of less than 0.05% by using the test procedure guidance in Regulatory i Position C.5.a. C.5.c and C.5.d of Regulatory Guide 1.52, Revisions 2, March 1978,* and the emergency filtration unit flow rate is 8000 cfm i 10%, and the emergency pressurization unit flow rate is 800 cfm i 10%;

2) Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accor-dance with Regulatory Position C.6.b of Regulatory Guide 1.52,

i Revision 2, March 1978,* meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revi-sion 2, March 1978,* for a methyl iodide penetration of less than 0.2%; and

3) Verifying an emergency filtration unit flow rate of 8000 cfm

+ 10% and an emergency pressurization unit flow rate of 800 cfm I 10% during system operation when tested in accordance with ANSI N510-1980.

c. After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation, by verifying, within 31 days after removal, that a laboratory analysis of a repre-sentative carbon sample obtained in accord 6nce with Regulator _5 1

Position C.6.b of Regulatory Guide 1.52, Revisip 2, March 19' ggf7 meets the laboratory testing criteria of ReguldtofRegulatoryGuide1.52, or a methyl Revision iodide penetration of less than 0.2%;

d. At least once per 18 months by:

1) Verifying that the total pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 8.0 inches Water Gauge while operating the emergency filtration unit at a flow rate of 8000 cfm i 10%, and is less than 9.5 inches Water Gauge while operating the emergency pressurization unit at a flow rate of 800 cfm i 10%;

2) Verifying that on a Safety Injection, Loss-of-Offsite Power, or Intake Vent-High Radiation test signal, the train automatically ,

switches into the emergency recirculation mode of operation with l flow through the HEPA filters and charcoal adsorber banks; J

3) Verifying that the emergency pressurization unit maintains the control room at a positive pressure of greater than or equal ,

m) Ac.E 450'1-3L4 T$ 1

ANSI N510-1980 shall be used in place of ANSI N510-1975. g _

l

- - Q pstrJ5D W h COMANCHE PEAK - UNIT 1 3/4 7-18 res9 l

Attachment to TXX-89169 h3 g p ag e

. . April 14, 1989 FFV e QJ' J Pa g e 5 2 o f 81-PLANT SYSTEMS .

_m

( ) SURVEILLANCE REQUIREMENTS (Continued)

Revis. ion 2, March 1978,* and verifying the flow rate is 15,000 cfm i 10% per ESF Filtration Unit when tested in accordance with ANSI N510-1980; and

2) Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accor-dance with Regulatory Position C.6.b of Regulatory Guide 1.52, 75 Revision 2, March 1978,* meets the laboratory testing criteria of Regulatory PoM on C.6.a of Regulatory Guide 1.52, 2evi- D #7 sion 2, March 1 98f or a methyl lodide penetration of less than 1.0%.

c. After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation, by verifying, within 31 days after removal, that a laboratory ant. lysis of a repre '

sentative carbon sample obtained in accordance with Regulatory .

Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978,*

meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978,8 for a methyl iodide penetration of less than 1.0%;

d. At least once per 18 months by:

1) Verifying that the total pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 8.5 inches O Water Gauge while operating each ESF Filtration Unit at a flow rate of 15,000 cfm t 10%,

2) Verifying that each ESF Filtration Unit starts on a Safety Injection test signal,

3) Verifying that the heaters dissipate 100 2 5 kW when tested in accordance with ANSI N510-1980, and

4) Verifying that the train maintains the negative pressure envelope of the Auxiliary, Safeguards, and Fuel Buildings at a negative pressure of greater than or equal to 0.05 inch water gauge rela-tive to the outside atmosphere.

e. After each complete or partial replacement of a HEPA filter bank, by verifying that the associated ESF Filtration Unit satisfies the in place penetration and bypass leakage testing acceptance criteria of less than 1.0% in accordance with ANSI N510-1980 for a DOP test aerosol while operating the associated ESF Filtration Unit at a flow rate of 15,000 cfm i 10%; and

f. After each complete or partial replacement of a charcoal adsorber bank, by verifying that the associated ESF Filtration Unit satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 1.0% in accordance with ANSI N510-1980 for a halogenated hydrocarbon refrigerant test gas while operating the associated ESF filtration unit system at a flow rate of 15,000 cfm ! 10%. D ti-of'1

' .t AMH NSY

" ANSI N510-1980 diall be used in place of ANSI N510-1975 0 5di- f %

sJ .

COMANCHE PEAK - UNIT 1 3/4 7-21 _ _

...pr 1989 Page 53 of 81 f

TABLE 3.7-3 AREA TEMPERATURE MONITORING

~ MAXIMUM AREA TEMPERATURE LIMIT (*F)

Normal Abnormal Conditions Conditions

1. Electrical and Control Building Normal Areas 10? 131 Control Room Main Level (El. 830'-0") 80 104 Control Roora Technical Support Area

-(El . 840'-6") 104 104 UPS/ Battery Rooms 104 113 Chiller Equipment Areas 122 131

2. Fuel Building Normal Areas 104 131 Spent Fuel Pool Cooling Pump Rooms 122 131

3. Safegaurds Building TS Normal Areas Motor-Orive.. yAFW, RHR, SI, Containment 104 131 d-dI7 O Spray Pump Rooms RHR Valve and Valve Isolation Tank Rooms 122 122 131 131 RHR/CT Heat Exchanger Rooms 122 131 Diesel Generator Area 122 131 Diesel Generator Equipment Rooms 130 131 Day Tank Room 122 131

4. Auxiliary Building Normal Areas 104 131 CCW, CCP Pump Rooms 122 131 CCW Heat Exchanger Area 122 131 CVCS Valve and Valve Operating Rooms 122 131 Auxiliary Steam Drain Tank Equipment Room 122 131 Waste Gas Tank Valve Operating Room 122 131 l 5. Service Water Intake Structure 127 131

6. Containment Building General Areas 120 129 CRDM Platform 140 149 Reactor Cavity Exhaust 150 175 R.C. Pipe Penetrations 200 209 CRDM Shroud Exhaust 163 172 O

COMANCHE PEAK - UNIT 1 3/4 7-24

A.t t a c h m e nl ._1_o_T_X X. _fL916.9_._ __ ____. ._. _.. . _ . -

g 5 f Nh f l ELECTRICAL POWER SYSTEMS ,,

( ) ' SURVEILLANCE REQUIREMENTS (Continued) TS 6400 - 7 70b WOY shall be loaded to'an indicated 7000 - 500& kW# and during the remaining 22 hours2.546296e-4 days 0.00611 hours 3.637566e-5 weeks 8.371e-6 months of this test, the diesel' generator shall be loaded to an indicated 6300 - 7000 kW# . The generator voltage and frequency shall be 6900 1 690 volts and 60 1 1.2 Hz within 10 seconds after the start signal; the steady-state generator voltage and frequency shall be maintained within these limits during this test. Within 5 minutes after completing this 24-hour test, perform Specification 4.8.1.1.2f.4)b);*

8) Verifying that the auto-connected loads to each diesel generator do not exceed the continuous rating of 7,000 kW;

9) Verifying the diesel generator's capaoility to,:

a) Synchronize with the offsite p; ar source while the -

generator is loaded with its emergency loads upon a simulated restoration of offsite power, b) Transfer its loads to the offsite power source, and c) Be restored to its standby status.

N 10) Verifying that with the diesel generator operating in a test (d mode, connected to its bus, a simulated Safety Injection signal overrides the test mode by: (1) returning the diesel generator to standby operation, and (2) automatically energizing the emergency loads with offsite power;

11) Verifying that the fuel transfer pump transfers fuel from fuel storage tank to the day tank of its associated diesel via the installed lines;

12) Verifying that the automatic load sequence timers are OPERABLE with the interval between each load block within + 10% of its design interval;

13) Verifying that the following diesel generator lockout features prevent diesel generator starting:

This band is meant as guidance to avoid routine overloading of the diesel generator. Momentary load excursions outside this band due to changing bus loads shall not invalidate the test.

If Specification 4.8.1.1.2f.4)b) is not satisfactorily completed, it is not necessary to repeat the preceding 24-hour test. Instead, the diesel generator

- may be operated between 6300 - 7000 kW for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or until operating tempera-(y ture has stabilized before repe6 ting 4.S.1.1.2f.4)b).

COMANCHE PEAK - UNIT 1 3/4 8-7

v ww

Pa" g e 5 5 o f 81 TABLE 4.8-1 i

,,\ DIESEL GENERATOR TEST SCHEDULE N-) .

NUMBER OF FAILUllES IN NUMBER OF FAILURES IN LAST 20 VALID TESTS

LAST 100 VALID TESTS

TEST FREQUENCY 11 54 Once per 31 days

> 2** >5 Once per 7 days f

(_,

" Criteria for determining number of failures and number of valid tests shall be in accordance with Regulatory Position C.2.e of Regulatory Guide 1.108, but determined on a per diesel generator basis.

For the purpose of determining the required test frequency, the previous test failure count may be reduced to zero if a complete diesel overhaul to like new condition is completed, provided that the overhaul, including appropriate post-maintenance operation and testing, is specifically approved by the manufacturer and if acceptable reliability has been demonstrated. The reliability criterion shall be the successful completion of 14 consecutive tests in a single series. Tum.e4fhesetes shall be in accordance with the T3 routine Surveillance Requirements 4.8.1.1. 2a. 4) and 4. 8.1.1. 2a. 5) and fourt testeh-accordancewihthe-184-day-testing-requirement of-Survefliance d g pd Wment--4r8:-Irldf. If this criterion is not satisfied during the first series of tests, any alternate criterion to be used to transvalue the failtse count to zero required NRC approval.

f

- The associated test frequency shall be maintained until seven consecutive failure free demands have been performed and the number of failures in the last 20 valid demands has been reduced to one.

)

p l N..}

COMANCHE PEAK - UNIT 1 3/4 8-9

g 5 of M D. C. SOURCES 13 V SURVEILLANCE REQUIREMENTS (Continued)

b. At least once per 92 days and within 7 days after a battery discharge with battery terminal voltage below 110 volts, or battery overcharge with battery terminal voltage above 150 volts, by verifying that:

1) The parameters in Table 4.8-2 meet the Category B limits,

2) There is no visible corrosion at either terminals or connectors, or the connection resistance of these items is less than 150 x 10 6 ohm, and

3) The average electrolyte temperature of 12 of connected cells is above 70 F. ,

c. At least once per 18 months by verifying that:

1) The cells, cell plates, and battery racks show no visual indication of physical damage or abnormal deterioration,

2) The cell-to-cell and terminal connections are clean, tight, and coated with anticorrosion material,

3) The resistance of each cell-to-cell and terminal connection is b'N less than or equal to 150 x 10 8 ohm, an 55 n5 BWI

4) The battery charger will supply at leas 3 W a peres at 130 volts for at least 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

d. At least once per 18 months, during shutdown, by verifying that the battery capacity is adequate to supply and maintain in OPERA 8LE status all of the actual or simulated emergency loads for the design duty cycle when the battery is subjected to a battery service test; j

e. At least once per 60 months, during shutdown, by verifying that the battery capacity is at least 80% of the manufacturer's rating when subjected to a performance discharge test. Once per 60-month interval this performance discharge test may be performed in lieu of the battery service test required by Specification 4.8.2.1d.; and

f. At least once per 18 months, during shutdown, by giving performance discharge tests of battery capacity to any battery that shows signs of degradation or has reached 85% of the service life expected for the application. Degradation is indicated when the battery capacity drops more than 10% of rated capacity from its average on previous performance tests, or is below 90% of the matiufactilrer's rating.

O V

COMANCHE PEAK - UNIT 1 3/4 8-12 l

_ - - - - - - - 1

v ~ v vw w v

. . April 14, 1989

~ ~

D.O.950NCEY SHUTDOWN

) LIMITING CONDITION FOR OPERATION 3.8.2.2 As a minimum, two 125V D.C. station batteries of one train and at least one associated full-capacity charger for each required battery shall be OPERABLE.

APPLICABILITY: PMDES 5 and 6.

ACTION:

With the required battery train and/or required full-capacity chargers inoperable, immediately suspend all operations involving CORE ALTERATIONS, positive reactivity changes, or movement of irradiated. fuel; initiate correc- 75 tive action to restore the required batter train and full-capacity charger to g_0 4 L OPERABLE status as soon as possible, w1 in 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , depressurize and vent the Reactor Coolant System th h"a 2.9 square inch vent.

g leas 1 SURVEILLANCE REQUIREMENTS _

4.8.2.2 The above required 125V D.C. station batteries and full-capacity q charger shall be demonstrated OPERABLE in accordance with Specification 4.8.2.1.

V O COMANCHE PEAK - UNIT 1 3/4 8=14

v - - - - -w v Fage 58 of $8 3/4.8.3 ONSITE POWER DISTRIBUTION OPERATING C LIMITING CONDITION FOR OPERATION l 3.8.3.1 The following electrical busses shall be energized in the specified manner:

a. Train A A.C. Emergency Busses consisting of:

1) 6900-Volt Emergency Bus IEA1,

2) 480-Volt Emergency Bus 1EB1 from transformer TIEB1, and

3) 480-Volt Emergency Bus 1EB3 from transformer T1EB3.

b.

Train B A.C. Emergency Busses consisting of:

1) 6900-Volt Emergency Bus 1EA2,

2) 480-Volt Emergency Bus 1EB2 from transformer T1EB2, and ,

3) 480-Volt Emergency Bus 1EB4 from transformer TIEB4. I

c. 118-Volt A.C. Instrument Bus 1PC nd 1EC1 energized from its ( 02 7 associated inverter connected to D.C. B~us 1ED1*;

d. 118-Volt A.C. Instrument Bus IPC M and 1EC2 energized from its associated inverter connec D.C. Bus 1ED2*;

Pt.3 ud

e. 118-Volt A.C. Instrument salE energized from its associated inverter connected to D.C. 1ED3*;

O f.

LPC %~.1 118-Volt A.C. Instrument s 1EC energized from its associated inverter connected to D.C. us ED4*;

g. Train A 125-Volt D.C. Busses 1ED1 and IED3 energized from Station Batteries BT1EDI and BTIED3, respectively; and

h. Train B 125-Volt D.C. Busses 1ED2 and 1ED4 energized from Station Batteries BT1ED2 and BTIED4, respectively.

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

ACTION:

a. With one of the required trains of A.C. emergency busses not fully energized, reenergize the trains within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 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 COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

The inverters may be disconnected from,one D.C. bus for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months as necessary, for the purpose of performing an equalizing charge on their asso-ciated battery train provided: (1) their instrument busses are energized, and (2) the instrument busses associated with the other battery train are energized from their associated inverters and connected to their associated f D.C. bus.

COMANCHE PEAK - UNIT 1 3/4 8-15

- - y us 0, o o- - - -- m Page 59 of 81 e

ONSITE POWER DISTRIBUTION

(~'g LIMITING CONDITION FOR OPERATION

%) .

ACTION (Continued) "

b. With one A.C. instrument bus or two A.C. instrument busses (consisting of one 7.5 KVA protection channel and one 10KVA vital bus of the same train) de energized, re-energize the A.C. instrument bus (ses) within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-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 COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ,

c. With one A.C. instrument bus or two A.C. instrument busses (consisting of one 7.5 KVA protection channel and one 10 KVA vital bus of the same train) operating with the associated inverter (s) not connected with the D.C. source (s), or operating with the inverter (s) not supplying the A.C. instrument bus (but with the instrument bus energized from its associated bypass distribution source), energize the A.C. instru -j3 ment bus (ses) from its associate D.C. bus within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in ~

at least HOT STANDBY within the xt 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . he rf e r c ewed,JAssoc k N Taft.j1

d. With one D.C. bus not energized from its associated station Dattery, reenergize the D.C. bus from its associated station battery within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-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 COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

O SURVEliLANCE REQUIREMENTS 4.8.3.1 The specified busses shall be determined energized in the required manner at least once per 7 days by verifying correct breaker alignment and indicated voltage on the busses.

1 O

i COMANCHE PEAK - UNIT 1 3/4 8-16

Attachment to TXX-89169 Pa 6 fo ELECTRICAL EQUIPMENT PROTECTIVE DEVICES SURVEILLANCE REQUIREMENTS (Continued) rotating basis, at least 10% of the circuit' breakers (whichever is greater) of each current rating and performing the following:

a) A CHANNEL CALIBRATION of the associatad protective relays, b) An integrated system functional test which includes simulated automatic actuation of the system and verifying that each relay and associated circuit breakers and control circuits function as designed, and l .

c) For each circuit breaker found inoperable during these functional tests, one or an additional representative sample of at least 10% of all the circuit. breakers of the inoperable type shall also be functionally tested until no ,

more failures are found or all circuit breakers of that type have been functionally tested; and

2) By selecting and functionally testing a representative sample -

of at least 10% of each type 480 V molded case circuit breakers and of lower voltage circuit breakers. Circuit breakers i selected for functional testing shall be selected on a rotating basis. Testing of these circuit breakers shall consist of injecting a current with a value equal to 300% of the pickup of the long-time delay trip element and 150% of the pickup of the short-time delay trip element, and verifying that the circuit breaker operates within the time delay band width for that current specified by the manufacturer. The instantaneous element shall be tested by injecting a current equal to 120% of the pickup value of the element and verifying that the circuit breaker trips instantaneously with no intentional time delay.

Ided case circuit breaker. testing shall also follow this f p&ocedure except that generally no more than two trip elements. f5 e ti.me delay and instantaneous, will be involved. The instanta-f[ neous element for molded case circuit breakers shall be tested gPo'I4 by injecting a current for a frame size of 250 amps or less with 9

Y to erances of +40%, -25% and a frame size of 400 amps or greater

, 6 + 25% and verifying that the circuit breaker trips instanta-rhouslywithnoapparenttimedelay. Circuit breakers found inoperable during functional testing shall be restored to OPERABLE status prior to resuming operation. For each circuit breaker found inoperable during these functional tests, an additional representative sample of at least 10% of all the circuit breakers of the inoperable type shall also be functionally tested until no more failures are found or all circuit breakers of that type have been functionally tested;

b. At least once per 60 months by subjecting each circuit breaker to an inspection and preventive maintenance in accordance with procedures prepared in conjunction with its manufacturer's recommendations.

COMANCHE PEAK - UNIT 1 3/4 8-19

. . myeyy ge, 3psy v - - - - -

-v_w v Page 61 of 81

<- SPECIAL TEST EXCEPTIONS

_g 3/4.10.5 POSITION INDICATION SYSTEM - SHUTDOWN LIMITING CONDITION FOR OPERATION f5 H-b7 $

3.10.5 The limitations of Spec fication 3.1.3.3 may be suspended during the i

performance of individual shutdown and control rod drop time measurements fLrw,1 g uJJmaysurwn" etbs"I"e=&M MJ:"cn M

a. Only one shutdown or control bank is withdrawn from the fully inserted position at a time, and b; The digital rod position indicator is OPERABLE during the withdrawal of the rods.*

APPLICABILITY: MODES 3, 4, and 5 during performance of red drop time measurements e wp y , s ur oe s t tw e . e4 J;p rsJ pn;+w 14t~% w ACTION: cPrAAorArry, With the required digital rod position indicator (s) inoperable or with more than one bank of rods withdrawn, immediately open the Reactor trip breakers.

(

V SURVEILLANCE REQUIREMENTS 4.10.5 The above required digital rod position indicator (s) shall be determined to be OPERABLE within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to the start of and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter during rod drop time measurements by verifying the Demand Position Indication System and the Digital Rod Position Indication System agree:

a. Within 12 stcps when the rods are stationary, and

b. Within 24 steps during rod motion.

T5

This requirement is not applicable during the ibration of the NO Digital Rod Position Indication System provided: (1) K is maintained lessthanorequalto0.95,and(2)onlyoneshutdownof$ontrolrodbankf is withdrawn from the fully inserted position at one time.

O COMANCHE PEAK - UNIT 1 3/4 10-5

Attachment to TXX-89169 i Pa 6 of Na RADIOACTIVE EFFLUENTS GAS STORAGE TANKS LIMITING CONDITION FOR OPERATION 3.11.2.2 The quantity of radioactivity contained in each gas storage tank shall be limited to less than or equal to 200,000 Curies of noble gases (considered as Xe-133 equivalent).

APPLICABILITY: At all times.

ACTION:

a. With the quantity of radioactive material in any gas storage tank exceeding the above limit, immediately suspend all additions of radioactive material to the tank, within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduce the tank g contents to within the limit, a describe the events leading to this condition in the next Semiann lit ioactive Effluent Release Report, .

Wo 75 pursuanttoSpecification6.9..g

b. The provisions of Specifications 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS

.2 The quantity of radioactive material contained in each gas storage tr5 tank s all be determined to be'within the above limit at least once per 92 O (QSy's ons en radioactive materials are being added to the tank.

p.og l

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COMANCHE PEAK - UNIT 1 3/4 11-3

. . Apr 1 9 Page 63 of 81 , 4 REACTIVITY CONTROL SYSTEMS

,m BASES BORATION SYSTEMS (Continued)

With the RCS temperature below 200 F, one Boron Injection System is acceptable without single failure consideration on the basis of the stable .

reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single Boron Injection System becomes inoperable.

The limitation for a maximum of two charging pumps to be OPERABLE and the requirement to verify one charging pump to be inoperable below 350 F provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV.

The limitation for minimum solution temperature of the borated water sources

. are sufficient to prevent boric acid crystallization with the highest allowable '

boron concentration. l The boron capability required below 200 F is sufficient to provide a SHUTDOWN MARGIN of 1% ak/k after xenon decay and cooldown from 200*F to 140*F.

This condition requires either 1,100 gallons of 7000 ppm borated water from the boric acid storage tanks or 7,113 gallons of 2000 ppm borated water from the RWST.

(n)

" As listed below, the required indicated levels for the boric acid storage tanks and the RWST include allowances for required / analytical volume, unusable volume, measurement uncertainties (which include instrument error and tank tolerances, as applicable), system configuration requirements, and other required volume. ,

Tank MODES Ind. Unusable Required Measurement System Other Level Volume Volume Uncertainty Config. (gal)

(gal) (gal) (gal)

RWST 5,6 25% 47,472 7,113 5% of span 57,857 N/A 1,2,3,4 96% 47,472 70,702 5% of span N/A 355,557*

Boric 5,6 10% 3,221 1,100 6% of span N/A N/A TS Acid 1,2,3,4 50% 3,221 15,700 6% of span N/A N/A Storage 8PO7 9 Tank

$ 5 e(u.td1 2.0%

3, W to l00 A d 'f** W Nf The OPERABILITY of one Boron Injection System during REFUELING ensures that this system is available for react.ivity control while in MODE 6.

J LJ

Additional volume required to meet Specification 3.5.4.

COMANCHE PEAK - UNIT 1 B 3/4 1-3

~

~~ "

Page 64 of 81 ]

" 5 POWER DISTRIBUTION LIMITS

()

e ~~

BASES DNB PARAMETERS (Continued)

The additional surveillance requirement sociated with the RCS total flow rate are sufficient to ensure that fie v meas ement uncertainties are 75 limited to 1.8% as assumed in the Impro ed Design rocedure Report for CPSES. '

gh-043 The r m~ t Performance of a precision seconda calor' etric is required to precisely determine the RCS temperature. The transit time flow meter, which uses the N-16 system signals, is then used to accurately measure the RCS flow. Subsequently, the RCS flow detectors (elbow tap differential pressure detectors) are normalized to this flow determination and used throughout the cycle.

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"-- v Page 66 of 81 REACTOR COOLANT SYSTEM BASES i . TS LOW TEMPERATURE OVERPRESSURE PROTECTION (Continued) 4

~ and valve opening, instrument uncertainties, and single failur . o .he tran-sients noted, the resulting pressure will not exceed the nomina) }6 ffective Full Power Years (EFPY) Appendix G reactor vessel NDT limits and- e forces gen-erated due to PORV cycling do not exceed PORV piping and structural limitations.

l To ensure that mass and heat input transients more severe than those assumed ceroot occur, Technical Specifications require the lockout of all safety injectior ;wnps and one charging pump while in MODES 4, 5 and 6 with the reactor vessel b d installed, and disallow start of an RCP if secondary temperature is more thar. LC9F above primary temperature.

Operation below 350'c but greater than 325 F with charging and safety injec-tion pump 3 N ERABLE is allowed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . Given the short time duration that this condition is allowed initiation of both trains of safety injection during this 4-hour time frame due to operator error or a single failure occurring during testing of a redundant channel are not considered to be credible accidents.

Plant specific analysis has shown that the Cold 0' overpressure Mitigation System (COMS) arming ten.perature may be reduced from 350*F to 320'F if the following additional restrictions are met:

V 1. At least one recctor coolant pump must be in operation.

2. Pressurizer level is less than or equal to 92%.

3. The plant heatup rate shall be limited to 60 F in any one hour period.

TheseconditionsappigwheneverthetemperatureofoneormoreoftheRCScold legs is less than 350 F but all RCS cold legs are greater than or equal to 320 F.

When any of the RCS RCS cold leg temperatures drop below 320 F, the original requirements on low temperature operation apply. ,

The Maximum Allowed PORV Setpoint for the LTOPS will be updated based on the results of examinations of reactor vessel material irradiation surveillance specimens performed as required by 10 CFR Part 50, Appendix H, and in accordance with the schedule in Table 4.4-2.

3/4.4.9 STRUCTURAL INTEGRITY -

The inservice inspection and testing programs for ASME Code Class 1, 2, and 3 components ensure that the structural integrity and operational readiness of these components will be maintained at an acceptable level throughout the life of the plant. These programs are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code edition and applicable Addenda as required by 10 CFR 50.55a(g) except where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i).

Components of the Reactor Coolant System were designed to provide access O to permit inservice inspections in accordance with Section XI of the ASME Boiler and Pressure Vessel Code, 1980 Edition. ,

COMANCHE PEAK - UNIT 1 B 3/4 4-14 1

i I

Attachment to TXX-89169 Nh I

April 14, 1989 Page 67 of 81 h 3/4.5 EMERGENCY CORE C00 LING SYSTEMS o

l Q BASES 3/4.5.1 ACCUMULATORS The OPERABILITY of each Reactor Coolant System (RCS) accumulator ensures that a sufficient volume of borated water will be immediately forced into the reactor core through each of the cold legs in the event. the RCS pressure falls l

below the pressure of the accumulators. This initial surge of water into the core provides the initial cooling mechanism during large RCS pipe ruptures. (

Tne limits on accumulator volume, boron concentration and pressure ensure that the assumptions used for accumulator injection in the safety analysis are g met. The required indicated accumulator volumes and pressures include A _ _x !

5 ercent _ measurement uncertaintvJTAe 14cHLG~GERIA vi'* 'es of 31 '7" ~<t bp. l I % a,-s yns..U.a the aa I mal r 1:, us o f G It_9.gs.IIem n

.-4 tS9 7 git.grespe<+:ve y, ein n as

- ihe accumulator power operated isolation valves are considered to NQ;L',,

" operating bypasses" in the context of IEEE Std. 279-1971, which requires tha bypasses of a protective function be removed automatically whenever permissive conditions are not met. In addition, as these accumulator isolation valves fail to meet single failure criteria, removal of power to the valves is required by BTP ICSB 18. This is accomplished via key-lock control board cut-off switches.

(q

'y The limits for operation with an accumulator inoperable for any :ason except an isolation valve closed minimizes the time exposure of the plant,to a LOCA event occurring concurrent with failure of an additional accumulator which may result in unacceptable peak cladding temperatures. If a closed isolation valve cannot be imtrediately opened, the full capability of one accumulator is not available and prompt action is required to place the reactor in a mode where this capability is not required.

3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS The OPERABILITY of two independent ECCS subsystems ensures that sufficient emergency core cooling capability will be available in the event of a LOCA assuming the loss of one subsystem through any single failure consideration.

Either subsystem operating in conjunction with the accumulators is capable of supplying sufficient core cooling to limit the peak cladding temperatures within acceptable limits for all postulated break sizes ranging from the double ended break of the largest RCS cold leg pipe downward. In addition, each ECCS subsystem provides long-term core cooling capability in the l recirculation mode during the accident recovery period. I With the RCS temperature below 350 F, one OPERABLE ECCS subsystem is acceptable without single failure consideration on the basis of the stable l reactivity condition of the reactor and the limited core cooling requirements.

The limitation for a maximum of two charging pumps to be OPERABLE and the requirement to verify one charging pump and all safety injection pumps COMANCHE PEAK - UNIT 1 B 3/4 5-1 i l

l l

l

o - myruu aw),u a ses - - - - - -

Page 68 of 81 %

+ 3/4.6 CONTAINMENT SYSTEMS

/~T 8ASES O

3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 CONTAINMENT INTEGRITY Primary CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the safety analyses. This restric-tion, in conjunction with the leakage rate limitation, will limit the EXCLUSION AREA BOUNDARY radiation doses to within the dose guideline values of 10 CFR 100 during accident conditions.

3/4.6.1.2 CONTAINMENT LEAKAGE The limitations on containment leakage rates ensure that the total ,

containment leakage volume will not exceed the value assumed'in the safety analyses at the peak accident pressure, P,. As an added conservatism, the measured overall integrated leakage rate is further limited to less than or equal to 0.75 L, or 0.75 L t, as applicable, during performance of the periodic test to account for possible degradation of the containment leakage barriers between leakage tests. T5 m be gm, For specific system configurations, credit tp en for a 30-day water Q sealthatwillbemaintainedtopreventcontain(mEWatmosphereleakagethrough V the penetrations to the environment. The following is a list of the containment isolation valves that meet this system configuration and the Maximum Allowed Leakage Rate (MALR) required to maintain the water seal for 30 days.

MALR MALR MALR Valve No. (cc/hr) Valve No. (cc/hr) Valve No. (cc/hr) g 1-8802A 424 819)I __ 114/ h 1-8905A 17 87-07 6 1-8802B 119.0 88J9B Ei9.~0 T-8f03B / / 4y.

1-8809A 77.0 1-8905C C / 43 1-8809B 77.0 t -8819C

\1-8819D

[07. 05D 22 P 1-8818A/#8MA 114 .0 F8835 36 p 1CT-142 .0 4734 1-8818B/st a 6 165 .0 1-8840 2577.0 1CT-145 680.0 't 7 3'l 1-8818C/stMc 207 .0 1-8841A/89056 4 .0 1HV-4776 4680 l 4 '734 Ig. 0 4 7.3 Y i

1-8318D/grHb 114j 1-8841B/ggosc 43 0 1HV-4777 The surveillance testing for measuring leakage rates is consistent with the requirements of 10 CFR 50 Appendix J.

3/4.6.1.3 CONTAINMENT AIR LOCKS The limitations on closure and leak rate for the containment air locks are required to meet the restrictions on CONTAINMENT INTEGRITY and containment

, leak rate. Surveillance testing of the air lock seals provides assurance that

( j the overall air lock leakage will not become excessive due to seal damage LG during the intervals between air lock leakage tests.

COMANCHE PEAK - UNIT 1 B 3/4 6-1

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3,/4. 6.1. 4 INTERNAL PRESSURE The limitations on containment internal pressure ensure that: (1) the j containment structure is prevented from exceeding its design negative pressure rJ l differential of 5 psid with respect to the outside atmosphere, and (2) the i containment peak pressure does not exceed the design pressure of 50 psig W O 7 / l during a LOCA. i The maximum peak pressure expected to be obtained from a LOCA event is 48.3 psig, which is less than design pressure and is consistent with the safety analyses. This valve includes the limit of 1.5 psig for initial positive containment pressure.

3/4.6.1.5 AIR TEMPERATURE The limitations on containment average air temperature ensure that the overall containment average air temperature does not exceed the initial tem-perature condition assumed in the safety analysis for a LOCA or steam line break accident. The average temperature shall be by an adjusted averaging of at least 2 of the measurements made at the listed locations, by fixed or portable instruments with allowance for temperature measurement uncertainty.

3/4.6.1.6 CONTAINMENT STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the containment will be maintained comparable to the original design standards for the life of the facility. Structural integrity is required to ensure that the containment will withstand the maximum pressure of 48.3 psig 'in the event of a LOCA. A visual inspection in conjunction with the Type A leakage tests is sufficient to demonstrate this capability.

3/4.6.1.7 CONTAINMENT VENTILATION SYSTEM i l

The 48-inch and 12-inch containment and hydrogen purge supply and exhaust isolation valves are required to be locked closed during plant operations since these val +es have not been demonstrated capable of closing during a LOCA or steam line break accident. Maintaining these valves locked closed during plant operation ensures that excessive quantities of radioactive materials will not be released via the Containment Ventilation System. To provide assurance that these containment valves cannot be inadvertently opened, the valves are locked closed in accordance with Standard Review Plan 6.2.4 which includes mechanical devices to seal or lock the valve closed, or prevents j power from being supplied to the valve operator. j The use of the Containment Ventilation System during operations is restricted to the 18-inch pressure relief discharge isolation valves (with an effective diameter of 3 inches) since, these venting valves are capable of closing during a LOCA or steam line break accident. Therefore, the Exclusion Area dose guideline of 10 CFR 100 would not be exceeded in the event of an accident during containment venting operation.

COMANCHE PEAK - UNIT 1 B 3/4 6-2 {

I

- _ _ _ ____ - ____ _. k

. . . pr 1 9 Page 70 of 81 c

CONTAINMENT SYSTEMS (3

V , BASES CONTAINMENT VENTILATION SYSTEM (Continued)

Leakage integrity tests with a maximum allowable leakage rate for contain-ment ventilation valves will provide early indication of resilient material seal degradation and will allow opportunity for repair before gross leakage failures could 3velop. Tne 0.60 L leakage limit of Specification 3.6.1.2b.

shallnotbeexceededwhentheleak$geratesdeterminedbytheleakageintegrity tests of thesa valves are added to the previously determined total for all valves,and penetrations subject to Type B and C tests.

3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS 3/4.6.2.1 CONTAINMENT SPRAY SYSTEM -

The OPERABILITY of the Containment Spray System ensures that containment depressurization and cooling capability will be available in the event of a LOCA or steam line break. The pressure reduction and resultant lower contain-ment leakage rate are consistent with the assumptions used in the safety analyses.

The Containment Spray System which is composed of redundant trains, pro- !

p) y vides post-accident cooling of the containment atmosphere. However, the Con-tainment Spray System also provides a mechanism for removing iodine from the {

l containment atmosphere and therefore the time requirements for restoring an J inoperable Spray System to OPERABLE status have been maintained consistent with that assigned other inoperable ESF equipment.

3/4.6.2.2 SPRAY ADDITIVE SYSTEM The OPERABILITY of the Spray Additive System ensures that sufficient Na0H is added to the containment spray in the event of a LOCA. The limits on NaOH volume and concentration ensure a long term pH value of between 8.5 and 10.5 for the solution recirculated within containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and j caustic stress corrosion on mechanical systems and components. The contained . i solution volume limit includes an allowance for solution not usable because of tank discharge Tine iacation or other physical characteristics. These assurp-tions are consistent with the iodine removal efficiency assumed in the safety analyses.

3/4.6.3 CONTAINMENT ISOLATION VALVES The OPERABILITY of the containment isolation valves ensures that the con-tainment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressuri-zation of the containment and is consistent with the requirements of General Design Criteria 54 through 57 of 10 CFR 50 Appendix A. Containment iso-lation within the time lir:its specified for those isolation valves designed to T5 f-)

(' close automatically ensures that the release of radioactive material to the en-vironment will be consistent with the assumptions used in the analyses for a LOCA.

g l--IA)SC W ] \

COMANCHE PEAK - UNIT 1 B 3/4 6-3 l

l AktachmenttoTXX-89169 I

- April 14, 1989 Page 71 of 81

~

f INSERT TO BASES 3/4.6.3 i o GDC-57 penetrations are provided with at least one remote manual containment isviation valve outside containment and closed systems inside containment.

The closed system inside containment ensures that the containment atmosphere

! will be isolated from the outside environment. Therefore, GDS-57 isolation valves, outside containment, are not subject to the requirements of Specification 3/4.6.3.

l \

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pr 1 9 Page 72 of 81 e

PLANT SYSTEMS ,

i V BASES 3/4.7.7 CONTROL ROOM HVAC SYSTEM The OPERABILITY of the Control Room HVAC System ensures that: (1) the control room ambient air temperature does not exceed the allowable temperature per 3/4 7.10 for continuous-duty rating for the equipment and instrumentation cooled by this system, and (2) the control room will remain habitable including temperature for operations personnel during and following all credible accident conditions. Operation of the system with the heaters operating to maintain low humidity using automatic control for at least 10 continuous hours in a 31-day period is sufficient to reduce the buildup of moisture on the adsorbers and HEPA filters. The OPERABILITY of this system in conjunction with control room design provisions is based on limiting the radiation exposure to personnel occupying the control room to 5 rems or less whole body, or its equivalent.

This limitation is consistent with the requirem nts eneral Design Criter- T3 ion 19 of 10 CFR 50 Appendix A. ANSIN510-1980wi1behsedTsma- ocedural 8"pos'7 guide for surveillance testing, g g g yo 3/4.7.8 PRIMARY PLANT VENTILATION SYSTEM - ESF FILTRATION UNITS The OPERABILITY of the ESF Filtration Units ensures that radioactive materials leaking from the ECCS equipment within the safeguards and auxiliary O buildings following a LOCA are filtered prior to reaching the environment.

b These filtration units also ensure that radioactive materials leakage from within the fuel building are filtered prior to reaching the environment.

Operation of the ESF filtration units with the heaters operating to mLintain l low humidity using automatic control for at least 10 continuous hours in a !

31-day period is sufficient to reduce the buildup of moisture on the adsorbers g l and HEPA filters. The operation of the ESF filtration units and the resultant effect on off

dosage calculations was assumed in the safety analyses. ANSI Pd N510-19 g w111 be used as a cedural guide for surveillance testing.

a a ,t A MT N 509 -t no i T eMat4ve-prejpLursJnyvMo of the Auxiliary, Safeguards and Fuel {

Buildings is the portions of these buildings which is exhausted post accident '

to ensure that potential ECCS leakage is filtered.

3/4.7.9 SNUBBERS All snubbers are required OPERABLE to ensure that the structural integrity of the Reactor Coolant System and all other safety-related systems is main-tained during and following a seismic or other event initiating dynamic loads. l Snubbers are classified and grouped by design and manufacturer but not by size. For example, mechanical snubbers utilizing the same design features of the 2-kip,10-kip and 100-kip capacity manuf actured by Company "A" are of the same type. The same design mechanical snubbers manufactured by Company "B" for the purposes of this Technical Specification would be of a different type, as would hydraulic snubbers from either manufacturer.

O y/

COMANCHE PEAK - UNIT 1 B 3/4 7-5 l

Apr 1 9 M Page 73 of 81 3/4.11 RADIOACTIVE EFFLUENTS BASES 3/4.11.1 LIQUID EFFLUENTS 3/4.11.1 LIQUID HOLDUP TANKS The tanks listed in this specification include all those unprotected outdoor tanks both permanent and temporary that are not surrounded by liners, dikes, or walls capable of holding the tank contents and that do not have tank overflows and surrounding area drains connected to the Liquid Radwaste Treatment System.

Restricting the quantity of radioactive material contained in the speci-fied tanks provides assurance that in the event of an uncontrolled release of the tank's contents, the resulting concentrations would be less than the limits of 10 CFR 20, Appendix B, Table II, Column 2, at the nearest potable water supply and the nearest surface water supply in an UNRESTRICTED AREA. ,

3/4.11.2 GASEOUS EFFLUENTS 3/4.11.2.1 EXPLOSIVE GAS MIXTURE This specification is provided to ensure that the concentration of poten-tially explosive gas mixtures contained in the WASTE GAS HOLDUP SYSTEM is maintained below the flammability limits of hydrogen and oxygen. Automatic control features are included in the system to prevent the hydrogen and oxygen O concentrations from reaching these flammability limits. These automatic control features include isolation of the source of hydrogen and/or oxygen.

Maintaining the concentration of hydrogen and oxygen below their flammability limits provides assurance that the releases of radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of 10 CFR Part 50 Appendix A.

3/4 11.2.2 GAS STORAGE TANKS The tanks included in this specification are those tanks for which the quantity of radioactivity contained is not limited directly or indirectly by another Technical Specification. Restricting the quantity of radioactivity contained in each gas storage tank prgrige ssurance that in the event of an Tg uncontrolled release of the tank's contents, th esulting whole body exposure ' ^

to a MEMBER OF THE PUBLIC at the nearest will n exceed 0.5 rem. This is WOI consistent with Standard Review Plan 11.3, Branch Technica osition ETSB 11-5,

" Postulated Radioactive Releases Due to a Waste Gas System Le or Failure," in NUREG-0800, July 1981.

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Attachment to TXX 89169 pp%f g ygp {

a . - April 14, 1989 Pa ge 75 o f 81 ADMINISTRATIVE CONTROLS UNIT STAFF (Continued)

b. At least cae licensed Operator shall be in the control room when fuel it il the reactor. In addition, while the unit is in MODE 1, 2, 3, er 4, at least one licensed Senior Operator shall be in the 75 control room; y

c. A Radiation Protection Technic aka d a Chemistry Technician

shall be on site when fuel is in the a tor;

d. All CORE ALTERATION 5 shall be observed and directly supervised by either a licensed Senior Operator or licensed Senior Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation;

e. A site Fire Brigade of at least five members

shall. be maintained on site at all times. The Fire Brigade shall not include the Shift

Supervisor and the two other memoers of the minimum shift crew necessary for safe shutdown of the unit and any personnel required for other essential functions during a fire emergency;

f. Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safety-related functions (e.g., licensed Senior Operators, licensed Operators,.

Radiation Protection Technicians, auxiliary operators, and key maintenance personnel).

The amount of overtime worked by unit staff members performing safety-related functions shall be limited in accordance with the NRC Policy Statement on working hours (Generic Letter No. 82-12); and l g. The Shift Operations Manager shall hold a Senior Reactor Operator l

license.

1

The Radiation Protection and the Chemistry Technicians and Fire Brigade composition may be less than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , in order to accommodate unexpected absence, provided immediate action is taken to fill the required positions.

COMANCHE PEAK - UNIT 1 E-2

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. s ipril 147T98 9' 1

Page 76 of 81 Nh f 7 ADMINISTRATIVE CONTROLS 6.2.3 INDEPENDENT SAFETY ENGINEERING GROUP (ISEG)

FUNCTION 6.2.3.1 The ISEG shall function to examine unit operating characteristics, NRC issuances, industry advisories, Licensee Event Reports, and other sources of unit design and operating experience information, including units of similar design, which may indicate areas for improving unit safety. The ISEG shall make detailed recommendat i for y vistd,procedu TS cations, mainta.idnce activi e.s',Aonoperations activities,gorequipment 6therweansmodifi-of improving unit safety to t Wa Ar-esider.t, " clear -Operatto .

ggg COMPOSITION "% U d ~i"~' ~

l 6.2.3.2 The ISEG shall be composed of at least five, de 1cated, full-time engineers located on site. Each shall have a bachelor's degree in engineering or related science and at least 3 years professional level experience in his field.

RESPONSIBILITIES 6.2.3.3 The ISEG shall be responsible for maintaining surveillance of unit activities to provide independent verification" that these activities are performed correctly and that human errors are reduced as much as practical. !

RECORDS 6.2.3.4 Records of activities performed b the ISEG shall be prepared, main- D tained, and forwarded each calendar month o " ice PresidatrNuc4 eat-0peret4ons. 8h075 6.2.4 SHIFT TECHNICAL ADVISOR N "'"' 'h*

6.2.4.1 The Shift Technical Advisor (STA) shall provide advisory technical support to the Shift Supervisor in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit.

The Shift Technical Advisor shall have a bachelor's degree or equivalent in a scientific or engineering discipline and shall have received specific training in the response and analysis of the unit for transients and accidents, and in unit design and layout, including the capabilities of instrumentation and controls in the control room.

6.3 UNIT STAFF QUALIFICATIONS 6.3.1 Each member of the unit staff shall meet or exceed the minimum qualifica-tions of ANSI-N18.1-1971 for comparable positions, except for the Radiation Protection Manager ** who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975, for a Radiation Protection Manager. The licensed Operators and Senior Operators shall 'also meet or exceed the minimum qualifica-tions of the supplemental requirements specified in Sections A and C of Enclo-sure 1 of the March 28, 1980 NRC letter to all licensees. (Prior to meeting the qualifications of ANSI N18.1-1971, technicians and maintenance personnel may be permitted to perform work in the specific task (s) for which qualification has been demonstrated.)

Not responsible for sign-off function.

- Until the Radiation Protection Manager meets all qualification per R.G.I.8, September 1975, an individual who meets all those qualifications shall support the Radiation Protection Manager.

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Attachment to TXX-89169 f a rv P9ep WFW(

. April 14, 1989 Page 77 of 81

. ADMINISTRATIVE CONTROLS RESPONSIBILITIES (Continued)

j. Review of the Emergency Plan and shall submit re' commended changes to the ORC;

k. Review of changes to the PROCESS CONTROL PROGRAM, the OFFSITE DOSE I CALCULATION MANUAL, and Radwaste Treatment Systems;

1. Review of any accidental, unplanned or uncontrolled radioactive l

release including the preparation of reports covering evaluation, recommendations, and disposition of the corrective action to prevent recurrence and the forwarding of these reports to the Vice

. President, Nuclear Operations, and to the ORC;

m. Review of Unit operatAons to detect potential hazards to nuclear safety; ,

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n. Investigations or analysis of special subjects as re_questeCby the TS Chairman of the ORC or the Vice President,-Nuclear'OperatTons; gpos t l' voi,a sua ~; ud o ReviN.hh~s71re'FrotectM arid implement 4ng-procedureR s and._s ubstta Lof-re comme nde<Lchang e s-to-the-OR C ;-a nds ;

W/st/wm/~mnsJ l p. Review of the Technical Requirements Manual and revision thereto.

6.5.1.7 The 50RC shall:

a. Recommend in writing to the designated manager (see Specification 6.5.3) approval or disapproval of items considered under Specifica-tion 6.5.1.6a through e prior to their implementation; b ., Render determinations in. writing with regard to whether or not each item considered under Specification 6.5.1.6a. through e. constitutes an unreviewed safety question; and

c. Provide written notification within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to the Executive Vice President-Nuclear Engineering and Operations and the Operations Review Committee of disagreement between the 50RC and the designated manager (see Specification 6.5.3); however, the Vice President, Nuclear Operations shall have responsibility for resolution of such disagreements pursuant to Specification 6.1.1.

RECORDS

6. 5.1. 8 The 50RC shall mi.intain written minutes of each SORC meeting that, at a minimum, document the results of all 50RC activities performed under the responsibility provisions of these Technical Specifications. Copies shall be provided to the Vice President-Nuclear Operations and the Operations Review Committee.

COMANCHE PEAK - UNIT 1 6-7

Attachment to TXX-89169

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(~ g g etes> yh y' g

. ' April 14, 1989 Page 78 of 81 ADMINISTRATIVE CONTROLS _

TECHNICAL REVIEW AND CONTROLS (Contirtued)

b. Proposed tests and experiments which affect plan't nuclear safety shall be prepared, reviewed, and approved. Each such test or experi-ment shall be reviewed by a qualified individual / group other than the individual / group which prepared the proposed test or experiment.

Proposed test and experiments _shall be approved before implementation by the Manager, Plant,0peration'sddividualsysponsible for con- ~

ducting such revi ws/shall be members of the Nucleaf Operations Staff previously desi ated by t L Vice Presiden t J ear 0perations; gesp.,8 ble be t@ Ge;rw . g

c. Proposed change or moc1T1 cations to plant nuclear safe y-related structures, sys ms d components shall be eviewed as designated N00 by the Vice Presipen engineering and4ee't m:t ha. J ach such modification shall be reviewed by a qualified individual / group meet-ing the experience h uire N18. - 971, Section 4.6 .

other than the individual / gents Af ANSgF5Ifp wnich des gned the modification, s q but who which may be from the pamee modification designed (ns.qanizallon Indiv1 uals/geoups a the individual / group responsible for conduct g juch reviews shall be previou ly designated by the Vice Presi ntdgngineering:ndConstwctic., Proposed s difica-tions to p nt uclear safety-related structures, systems nd components lh al be ap roved b the Manager, Plant 0 eratio s prior +

to implementh ion, g w k,-+L g.g ,

d. Individuals resp ns ble for reviews performed in accoJ44 e with the O requirements of Spe(cificR ions E 5 1 1a1T + 6 h ib, shall be members of the Nuclear Operations Management staff previously designed by the Vice President, Nuclear Operations. Each such review shall include a determination of whether or not additional cross-disciplinary review is necessary. If deemed necessary, such review shall be done in accordance with the appropriate qualification -

requirements;

e. Each review shall include a determination of whether or not an unreviewed safety question is involved. For items involving unreviewed safety questions, NRC approval shall be obtained prior to the Manager, Plant Operations, approval for implementation; and

f. The Security Plan and Emergency Plan, and implementing procedures, shall be reviewed at least once per 12 months. Recommended changes to the implementing procedures shall be approved by the Vice President, Nuclear Operations. Recommended changes to the Plans shall be reviewed pursuant to the requirements of Specifications 6.5.1.6 and 6.5.2.8 and approved by the Vice President, Nuclear Operations. NRC approval shall be obtained as appropriate.

6.5.3.2 Records of the above activities described in 6.5.3.1 shall be provided to the Vice President, Nuclear Operations, 50RC, and/or ORC as necessary for required reviews.

O -

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m - --- -

ADMINISTRATIVE CONTROLS A

V PROCEDURES AND PROGRAMS (Continued)

f. OFFSITE DOSE CALCULATION MANUAL implementation;

g. Quality Assurance for ef fluent and environmental monitoring; h.

W Fire ProtectyProgram implementation and Lg, s.,-e m ea ts t%mcJ r5 Technical 5 0F 1.

j implementation, g9-o t, 0 6.8.2 Each procedure and administrative policy of Specification 6.8.1 above, and changes thereto, shall be reviewed and approved prior to implementation as set forth in Specification 6.5 above.

6.8.3 The following programs shall be established, implemented, and maintained:

a. Primary Coolant Sources Outside Containment A program to reduce leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. The systems include the post accident recirculation portions of the Containment Spray System, Safety Injection System, Chemical and Volume Control System, RHR System, and RCS Sampling System (Post Accident Sampling System portion only). The program shall include the following:

l 1) Preventive maintenance and periodic visual inspection require-ments, and

2) Integrated leak test requirements for each system at refueling i cycle intervals or less.

b. In-Plant Radiation Monitoring A program which will ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions. This program shall include the following:

'1) Training of personnel,

2) Procedures for monitoring, and

3) Provisions for maintenance of sampling and analysis equipment.

c. Secondary Water Chemistry l A program for monitoring of secondary water chemistry to inhibit steam generator tube degradation and low pressure turbine disc stress corrosion cracking. This program shall include: )

1) Identification of a sampling schedule for the critical variables

(

and control points for these variables, l

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- ' Page 80 of 81

, ADMINISTRATIVE CONTROLS 6.13 PROCESS CONTROL PROGRAM (PCP) ,

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6 13.1 ine 7CP-shal' be-approvedly the-cms 4an_ prier-to-implementation &9-o75

's*(2bicenn;- SEtist hanges to the PCP:

a. Shall be documented and records of reviews performed shall be retained as required by Specification 6.10.3o. This documentation shall contain:

1) Sufficient information to support the change together with the appropriate analyses or evaluation justifying the change (s) and

2) A determination that the change will maintain the overall conformance of the solidified waste product to existing requirements of Federal, State, or other applicable regulations. .

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b. Shall become effective after review and acceptance by the 50RC and the approval of the Vice President, Nuclear Operations.

6.14 0FFSITE DOSE CALCULATION MANUAL (ODCM)

TS h.-14rl d he @ @ Shall he apDPOVed by MMhsiGr. priGr tG implement 4t4Q 99 cy ik emaa "itht plangestotheODCH:

/~'N a. Shall be documented and records of reviews performed shall be

() retained as required by Specification 6.10.30. This documentation shall contain:

1) Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change (s) and rs

2) A_f ermination that the change will maintain the level f rad active effluent control required by 10 CFR 2 0 CFR 90, 10 CFR 50.36a, and Appendix I to 10 CF and no dversely impact the accuracy or reliability uent, dose, or setpoint calculations,

b. Shall become effective after review and acceptance by the 50RC and the approval of the Vice President, Nuclear Operations.

c. Shall be submitted to the Commission in the forin of a complete, j legible copy of the entire ODCM as a part of or concurrent with the i Semiannual Radioactive Effluent Release Report for the period of the report in which any change to the 00CM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g., month / year) the change was implemented.

i COMANCHE PEAK - UNIT 1 6-23

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, ADMINISTRATIVE CONTROLS / ,

6.15 MAJOR ANGES TO LIQUID, GASEOUS, AND SOLID RADWASTE TREATMENT YSTEMS* #7 C 6.15.1 Licensee \-

(liquid, gaseous,(nitiatedmajorchangestotheRadwasteTreatmentSystems solid):

a. Shall be reported to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the valuation was reviewed by t'he 50RC. The discussion of each chang shall contain:

N A summary sof the evaluation that led to the determination that 1) the change could be made in accordance with 10 CFR 50.59;

2) Sufficient de ' led information to totall support the reason !

for the change w thout benefit of additi nal or supplemental information;

3) A detailed description of the equipme , components, and processes 5 involved and the inte ces with oth r plant s'ystems;

4) An evaluation of the chagge, which shows the predicted releases ofradioactivematerialsinliquipandgaseouseffluentsand/or quantity of solid waste thd predicted in the License appication

( differandfrom those previously amendments thereto;

5) An evaluation of the change, ich shows the expected maximum I exposures to a MEMBER OF THE UB IC in the UNRESTRICTED AREA

(' and to the general populati tha differ from those previously

\ estimated in the License a. licati and amendments thereto;

6) A comparison of the predi ted release of radioactive materials, in liquid and gaseous ef luents and in olid waste, to the actual releases for th period prior to hen the change is to be made;

7) An estimate of the e posure to plant operati g personnel as a result of the chan ; and

8) Documentation of he fact that the change was r iewed and ,

found acceptable by the 50RC. '

b. Shall-become effect ve upon review and acceptance by the ' ORC.

N N

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Licensees may cho e to submit the information called for in this Specification

( as part of the a ual FSAR update.

COMANCHE PEAK - NIT 1 6-24

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ML20244D810

Text

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