Text

Forwards Proposed marked-up Changes to NRC Second Draft Tech Specs,Reflecting Util Tech Spec Review Program
	ML20195G764
Person / Time
Site:	Comanche Peak
Issue date:	06/23/1988
From:	Counsil W; TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC)
To:	; NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
	TXX-88512, NUDOCS 8806280131
	Download: ML20195G764 (500)
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Ref # 10CFR50.36 7UELECTRIC June 23, 1988 William G. Counsil Executne the Presulent U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555

SUBJECT:

COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)

DOCKET N0. 50-445 PROPOSED CHANGES TO THE NRC SECOND DRAFT TECHNICAL SPECIFICATIONS REF: C. I. Grimes letter to W. G. Counsil dated March 2, 1988 Gentlemen:

n The referenced letter provided CPSES with a copy of the NRC's second draft

( ,/ Technical Specifications. TV Electric has reviewed this document and proposed changes are hereby provided in the attachments to this letter. Print-ready figures for this package are not included and will be provided at a later date.

A Technical Specification review program is in progress at CPSES. As such, the changes in Attachments 1 through 17 are reflective of this review program

as of this date. In some cases, sufficient design reviewed information is not available at this time to fully validate certain portions of the Technical Specifications. This information has been "bracketed". Validation against design reviewed documents will be completed prior to issuance of the CPSES Unit 1 operating license.

Also as part of our review program, and in an effort to enhance the Technical Specifications from the users perspective, certain specifications or portions of specifications have been relocated from the CPSES Technical Specifications

, to the programs described in Section 6.8 of the CPSES Technical Specifications. Information which falls into the above category can be identified in the following attachments by the word "Relocate".

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8806280131 880623 I l

PDR ADOCK 05000445 l '

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.i Pf] TXX-88512 Page 2 of 2 Finally, the attached Technical Specification changes are being tracked with identification numbers by TV Electric. An "ID#" is used to correlate the specific change to the applicable detailed description. If you require-additim 11 information regarding our tracking system, contact this office.-

Very truly yours, M .

W. G. Counsil RWH/grr Attachments c - Mr. R. D. Martin, Region IV (1 copy)

Resident Inspectors, CPSES (3 copies)

Mr. R. F. Warnick, NRC (1 copy)

Mr. J. H. Wilson, NRC (5 copies)

Mr. Bob Giardina, NRC (1 copy)

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IXX-88512 AllACHM(HT1 PAGE 1 0F 13 k

COMANCHE PEAK STEAM ELECTRIC STATION TECHNICAL SPECIFICATION 1

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IXXC512 ATTAC M NT 1 CPSES Technical Specifications f 4E 2 0F 13 NRC Draft 2 Markup Section 1 (j Change ID# Justification For Change Change #0001. Replaced the reference to table 3.6-1 with the Technical Specification Improvement Program (TSIP) since the table 3.6-1 has been relocated to the TSIP and any exception required during plant operations is annotated in the TSIP. This change is similar to that Licensed at Seabrook, Vogtle and Shearon Harris.

Change #0002. Added new definition for the Radioactive Effluent and Environmental Monitoring Manual (REEMM) to be consistent with what has been required by the NRC in other Licensing Dockets and to be consistent with the same reasoning of having a definition for the Offsite Dose Calculation Manual (00CM). The REEfH is described in section 6.15, referenced in section 3/4.11 and incorporates the requirements of sections 3/4.11 (in part) and 3/4.12. This change is similar,to that Licensed at Millstone.

Change #0600. The Primary Plant Ventilation System (PPVS) definition has been changed to eliminate the sentence that states the PPVS. does not include ESF Cleanup Systems. For the CPSES design the PPVS includes the ESF filtration units along with the normal operation filtration units. The testing requirements on the PPVS are included in Specification 3/4.7.8, PPVS-ESF FILTRATION

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t SECTION 1.0 DEFINITIONS O

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O COMANCHE PEAK - UNIT 1 1-0

IXX-88512 ATTACHMENT 1 f PAGE 4 CF 13 , ,,

. 1. 0 DEFINITIONS V, -

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

ACTION 1.1 ACTION shall be that part of a Technic.1 Specification which prescribes remedial measuaes required under designated conditions.

ACTUATION LOGIC TEST

1. 2 An ACTUATION LOGIC TEST shall be the application of various simulated input combinations in conjunction with each possible interlock logic state and verification of the required logic output. The ACTUATION LOGIC TEST shall include a continuity check, as a minimum, of output devices.

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

( AXIAL' FLUX DIFFERENCF

1. 4 AXIAL FLUX DIFFdRENCE shall be the difference in normalized flux signals between the top and bottom halves of a four section excore neutron detector.

CHANNEL CALIBRATION 1.5 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel such that it responds within the required range and accuracy to known values of input. The CHANNEL CALIBRATION shall encompass the entire channel including the sensors and alarm, interlock and/or trip functions and may be performed by any series of sequential, overlapping, or total channel steps such that the entire channel is calibrated.

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

I l O COMANCHE PEAK - UNIT 1 1-1

TXX-88512 AfiACHMENT1 59 PAGE 5 0F 13 ,af DEFINITIONS' b

(/ CONTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when:

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

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

2) Closed by manual valves, blind f'langes, or deactivated automatic valves secured in their closed positions, except as provided in kt he b '

b. All equipment hatches are closed and sealed, ID I: 0001 I

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

/7 CONTROLLED LEAKAGE l

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

l l CORE ALTERATIONS 1.9 CORE ALTERATI0HS shall be the movement or manipulation of any component M

within the reactor pressure vessel with the vessel head removed and fuel in I

the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe conservative position.

1

. DIGITAL CHANNEL OPERATIONAL TEST 1.10 A DIGITAL CHANNEL OPERATI0hAL 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.

! DOSE EQUIVALENT 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 thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, "Calculation of Distance Factors for Power and Test

Reactor Sites" or Table E-7 of NRC Regulatory Guide 1.109, Revision 1, October 1977.

~J l COMANCHE PEAK - UNIT 1 1-2

TXX-88512 ATTACHMENT 1 PAGE 6 0F 13 DEFINITIONS !

p E - AVERAGE DISINTEGRATION ENERGY 1.12 E shall be the average (weighted in proportion to the concentration of each radionuclide in the sample) of the sum of the average beta and gamma energies per disintegration (MeV/d) for the radionuclides with a halflife greater than ter. (10) einutes in the sample.

ENGINEERED SAFETY FEATURES RESPONSE TIME 1.13 The ENGINEERED SAFETY FEATURES (ESF) RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF Actuation Setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable.

EXCLUSION AREA BOUNDARY 1.14 The Exclusion Area Boundary, used for establishing effluent release limits is shown in Figure 5.1-1.

FREQUENCY NOTATION 1.15 The FREQUENCY NOTATION specified for the performance of Surveillance

, Requirements shall correspond to the intervals defined in Table 1.1.

IDENTIFIED LEAKAGE 1.16 IDENTIFIED LEAKAGE shall be:

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

b. Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of Leakage Detection Systems or not to be PRESSURE B0UNDARY LEAKAGE, or j c. Reactor Coolant System leakage through a steam generator to the Secondary Coolant System.

MASTER RELAY TEST l 1.17 A MASTER RELAY TEST shall be the energization of each master relay and verification of OPERABILITY of each relay. The MASTER RELAY TEST shall include a continuity check of each associated slave relay.

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l COMANCHE PEAK - UNIT 1 1-3

TXX-88512 ATTACHMENT 1 T"T PAGE 7 OF 13

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OEFINITIONS O . MEMBER (S) 0F THE-PUBLIC 1.18 HEMBER(S) 0F THE PUBLIC shall include all persons who are not occupa-tionally associated with the plant. This category does not include employees of the licensee, its contractors, or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries.

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

OFFSITE DOSE CALCULATION MANUAL 1.19 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm / Trip Setpoints, and in the conduct of the Environ-mental Radiological Monitoring Program.

OPERABLE - OPERABILITY 1.20 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, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its function (s) are also capable of performing their related support function (s).

OPERATIONAL MODE - MODE 1.21 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 1.22 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 1.23 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube leakage) through a nonisolable fault in a Reactor Coolant System component body, pipe wall, or vessel wall.

O COMANCHE PEAK - UNIT 1 1-4 I l

TXX-88512 l

ATTACHMENT 1 l PAGE 8 0F 13 {W DEFINITIONS- -

PRIMARY PLANT VENTILATION SYSTEM 1.24 A PRIMARY PLANT VENTILATION SYSTEM shall be any system designed and 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 particulates 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. E @ r:d [

Safety features Atac:pheric Cl;;nup Sy:t::: Or: not con:idered t b e " ! " ""' IM 0600 FLANT VENTILATION SYSTE". cc;p;n:nt:.

l PROCESS CONTROL PROGRAM 1.25 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 rolid radioactive wastes baseo on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71 and Federal and State regulations, burial ground requirements, and other require-ments governing the disposal of radioactive waste.

PURGE - PURGING 1.26 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

{n l required to purify the confinement.

QUADRANT POWER TILT RATIO 1.27 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 detector calibrated output to the average of the lower half excore detector calibrated outputs, whichever is greater. With one excore detector inoperable, the remaining three detectors shall be used for computing the average.

x45ERT RATED THERMAL POWER 10 1: 0002 i.29

+.-24 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3411 MWt. .

REACTOR TRIP SYSTEM RESPONSE TIME i.3o H9 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its Trip Setpoint at the channel sensor until loss of stationary gripper coil voltage.

REPORTABLE EVENT

1. 3 %

+:-30 A REPORTABLE EVENT shall be any of those conditions specified in 10 CFR 50.73.

COMANCHE PEAK - UNIT 1 1-5

TXX-88512 ATTAClelENT 1

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PAGE 9 0F 13 ,

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i l' RA_D10 ACTIVE EFFLUENT AND ENVIRONMENTAL MONITORING MANUAL 1.2% - The ~ RADI0 ACTIVE EFFLUENT AND ENVIRONMENTAL MONITORING MANUAL- (REEffi) shall outline the effluent and environmental sampling and analysis program.

used to determine the concentration of radioactive materials in those pathways which lead to the radiation exposures to MEMBER (S) 0F THE PUBLIC from routine station operation.

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IXX-88512 AilACHMENT1 PAGE 10 0F 13 hy DEFINITIONS SHU TDOWN MARGIN - -

1.32.

31 SHUTOOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical 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.

SLAVE RELAY TEST i.33 1-}e 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.

SOLIDIFICATION i.39 i-30 SOLIDIFICATION shall be the conversion of wet wastes into a form that meets shipping and burial ground require 6ents.

SOURCE CHECK 135 1-3+ A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity.

p STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of:

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 l component at the beginning of each subinterval.

l THERMAL POWER I.31 1-3fr THERMAL POWER shall be the total core heat transfer rate to the reactor coolant.

TRIP ACTUATING DEVICE OPERATIONAL TEST l.38 1-W A TRIP ACTUATING DEVICE OPERATIONAL TEST shall consist of operating the Trip Actuating Device and verifying OPERABILITY of alarm', interlock and/or I trip functions. The TRIP ACTUATING DEVICE OPERATIONAL TEST shall include I

adjustment, as necessary, of the Trip Actuating Device such that it actuates at the required setpoint within the required accuracy.

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COMANCHE PEAK - UNIT 1 1-6

IXX 88512 ATTACHMENT 1 PAGE 11 0F 13 DEFINITIONS UNIDENTIFIED LEA 1(AGE

t. M 4r48 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE or CONTROLLEO LEAKAGE.

UNRESTRICTED AREA t A C>

1 An UNRESTRICTED AREA shall be any area at or beyond the EXCLUSION AREA BOUNDARY for purposes of protection of individuals from exposure to radiation and radioactive materials, or any area within the EXCLUSION AREA BOUNDARY,

. access to which is not controlled by the licensee, and used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes.

VENTING i.91

-h40 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 HOLOUP SYSTEM t .9 2.

h41 A WASTE GAS HOLDUP SYSTEM shall be any system designed and installed to 7 reduce radioactive gaseous effluents by collecting Reactor Coolant System (h 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 environment.

O COMANCHE PEAK - UNIT 1 1-7 1

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TXX-88:12 AliACHIENT 1 PAGE 12 OF 13 y TABLE 1.1 y

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

O 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.

A At least once per 12 months. ,-

R At least once per 18 months.

S/U Prior to each reactor startup.

N.A. Not applicable.

P Completed prior to each release.

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COMANCHE PEAK - UNIT 1 1-8 i

IXX-08512 AliACMENT I PAGE 13 OF 13 TABLE 1.2 i OPERATIONAL MODES yD REACTIVITY % RATED AVERAGE COOLANT MODE CONDITION, K,7f , THERMAL POWER

TEMPERATURE

1. POWER OPERATION > 0.99 > 5% > 350*F

2. STARTUP > 0.99 5 5% > 350*F

3. HOT STANOBY < 0.99 0 > 350*F

4. HOT SHUT 00WN < 0.99 0 350*F > T

> 200*F avg

5. COLD SHUTOOWN < 0.99 0 1 200 F

6. REFUELING ** 1 0.95 0 1 140*F

Excluding decay heat.

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

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t O COMANCHE PEAX - UNIT 1 1-9 l

TXX-88512 AllACHMEHi2 PAGE 1 0F 20 4

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COMANCHE PEAX STEAM ELECTRIC STATION TECHNICAL SPECIFICATION 2.0 lO

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'W ATTACMMuli 2 CPSES Technical Specifications

,- PAE 2 0F 20 NRC Oraft'2 Markup Section 2 b

J Change 10# Justification For Change 0012 Deleted the > and < from the Trip Setpoints for P-10 for clarificatioi. In setting the P-10 bistable with the Standard Technical Specifications would require that the setpoint would have to.be exactly 10% with no deviation which is physically impossible.~ By making this change the Trip Setpoint would be assumed nominal and would be set as close as possible to 10%. This change is similar to that Licensed at Callaway and Wolf Creek.

0922- The parentheses around "Tc-Tc " are deleted to be consistent with the Westinghouse scaling calculations.

These parentheses were added during the January 11, 1988 meeting on-site for urknown reasons.

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111-88512 ATTACHMfMT2 PAGE 3 0F 20

'l SECTION 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS COMANCHE PEAX - UNIT 1 2-0 4

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

IXX-88512 AliACHMENT2 PAGE 4 0F 20 {j'(

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2. 0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFErY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temperature (Tavg) shall not exceed the limits shown in Figure [2.1-1.

APPLICABILITY: MODES 1 and 2.

ACTION:

Whenever the point defined by the combination of the highest operating loop average temperature and THERMAL POWER has exceeded the appropriate pressurizer pressure line, be in HOT STAND 8Y within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and comply with the require-ments of Specification 6.7.1.

REACTOR COOLANT '"EM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig.

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

ACTION:

MODES 1 and 2:

Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANDBY with the Reactor Coolart Sys im pressure within its limit within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and comply with the requiremt.nts of Specification 6.7.1, MODES 3, 4 and 5:

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

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COMANCHE PEAK - UNIT 1 2-1

IIX-88512 ir ATTACHMENT 2 '. ' "l PAGE 5 0F 20

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i FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT COMANCHE PEAK - UNIT 1 2-2 ,

111-88512 ATTACHMENT 2 FAGE 6 0F 20

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SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS

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

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

ACTION:

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

b. With the Reactor Trip ystem Instrumentation or Interlock Setpoint less ce;servative than the value shown in the Allowable Values column of Table 2.2-1, either:

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

( 2. Dec1 r' the channel inoperable and apply the applicable ACTION

b' statement requirement of Specification 3.3.1 until the channel is restored to OPERABLE status with its setpoint adjusted -

consistent with the Trip Setpoiat value.

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

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

. COMANCHE PEAK - UNIT 1 2-3 I

TABLE 2.2-1 n

o 3.,55 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS

~g3

g SENSOR

==

3m m

TOTAL ERROR y FUNCriONAL UNIT AJLOWANCE(TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE

] 1. Manual Reactor Trip N.A. N.A. N.A. N.A. N.A.

j g 2. Power Range, Neutron Flux . 4 l [ a. High Setpoint 7. 5 4.56 0 <109% of RTP* #

$111.2% of RTP*

., b. Low Setpoint 8.3 4.56 0 <27.2% of RTP" 125% of RTP*

3. Power Range, Neutron Flux, 1.6 0.5 0 <5% of RTP* with <6.3% of RTP* with

High Positive Rate i time constant i time constant

_2 seconds _2 seconds

4. Power Range, Neutron Flux, 1.6 0.5 0 <5% of RTP* with <6.3% of RTP* with y High Negative Rate i time constant i tinc ...nstant a

>2 seconds >2 seconds a

5. Intermediate Range, 17.0 8.4 0 $25% of RTP* $31% of RTP*

Neutron Flux

6. Source Range, Neutron Flux 17.0 10.0 0 $105 cps $1.4 x 105 cps

7. Overtemperature N-16 6.4 4.7' 1. 8 See Note 1 See Note 2

. 8. Overpower N-16 4.0 1.91 1.3 1112% 1114.5%

9. Pressurizer Pressure-Low 8.8 2.'81 1. 5 >1910 psig >l896 psig

10. Pressurizer Pressure-High _

7. 5 4.96 0.5 <E385psig L

399 psig D

n h3

RTP = RATED THERHAL PDWEP 7 '}

(, ,J

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TABLE 2.2-1 (Continued)

O ,,_

R ACTOR TRIP SYSTEM INSTPUMENTATION TRIP SETPOINTS . E35

.gn 5

E TOTAL SENSDR ERROR

$5 3~

g FUNCTIONAL UNIT ALLOWANCE (TA) Z (5) TRIP SETPOINT 2- ALLOWABLE VALUE x 11. Pressurizer Water Level-High 5.0 2.18 1.5 $92% of instrument

' $93.8% of instrument span span g s .

p 12. Reactor Coolant Flow-tow 2.5 1.31 0.6 290% of loop 188.8% of loop design flow ** design flow **

13. Steam Generator Water 8.8 7.08 1. 5 Level - Low-Low 143.4% of narrow 142.1% of narrow range instrument range instrument span span

14. Undervoltage - Reactor 7. 7 0 0 >(4830 volts- >4781 volts-Coolant Pumps Each bus each bus y 15. Uaderfrequency - Reactor 4.4 0 0 157.2 Itz un Coolant Pumps 157.1 Itz

16. Turbine Trip

a. Low Trip System Pressure N.A. (-

N.A. N.A. 145 psig

, 2}3psig

b. Turbine Stop Valve N A. N.A. N.A. 11% open 11% open Closure

. 17. Safety Injection Input N.A. N.A. N.A. N.A. N.A.

from ESF

- Loop design flow = 95,700 gps. C y

M

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N N ,]ol TABLE 2.2-1 (Continued) 3" gj g c3 REACTOR TRIP SYSTEM INSTRUNENTATION TRIP SETP0iNTS C E d:

E 5 SENSOR

BE E5 TOTAL LIC ERROR m FUNCTIONAL UNIT ni ALLOWANCE (TA) Z

- (S) TRIP SETPOINT ALLOWABLE VALUE ,

$E 18. Reactor Trip System

. Interlocks c ' '

35 a. Intermediate Range N.A. N.A. N.A. >l x 10 80 amps

-d ->6 x 10 88 amps

~

Neutron Flux, P-6

b. Low Power Reactor Trips Block, P-7

?) P-10 input N.A. N.A. N.A. ( 10% of RTP* $12.2% of RTP* IB : 0012

2) P-13 input N.A. N.A. N.A. $10% RTP* Turbine $12.2% RTP* Turbine first Stage Pres- First Stage Pressure n3 sure Equivalent Equivalent S

c. Power Range Neutron N.A. N.A. N.A.

Flux, P-8

$48% of RIP * $50.2% of RTP*

d. Power Range Neutron N.A. N.A. f;. A.

10% of RTP* $7.8% of RTP*

. Flux, P-10

19. Reactor Trip Breakers N.A. N.A. N.A N.A. N.A.

20. Automatic Trip and Interlock N.A. N.A. N.A. N.A. N.A.

Logic -

RIP = RAIE0 IHERMAL POWER E

V TABLE 2.2-1 (Continued) khh.

r-. Ekk

%5~

j TABLE NOTATIONS z 3" 2

NOTE 1: Overtemperature N-16 o

h N = K,-K2 [$!T-T*f]+K (P-P8) 3 - f, (aq) DhM

. 1*t2 5[ g s

e

Measured N-16 Power by ion chambers, Where: N =

" = Cold leg temperature *F, T

g

.T* = SS9.6*F, Reference T at RATED THERMAL POWER, c

K, = 1.069 -

K2 = 0.00948/*F, b

f{ 3

= The function generated by the lead-lag compensator for measured Tc '

ti, 12 = Time constants utilized in the lead-lag compensator for T , t, = 10 s, c

and T 2 = 35 0.000494/psig,

. K3 =

ri

G -

E O TABLE 2.2-1 (Continued) 3$5 8

TABLE NOTATIONS (Continued) $hh g NOTE 1: (Continued) *ER' m .

g~

P = Pressurizer pressure, psig.

7 P1 = 2235 psig (Nominal RCS operating pressure),

E S = Laplace transform operator, s-8, * -

e and f (aq) is a function of the indicated difference between top and bottom halves of i

detectors of the power-range neutron ion chambers; with gains to be selected based on measured instrument response during plant STAR 7UP tests such that:

(i) for q t

Ab between -35% and +10%, f (aq) = 0, where tq and q b are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and qt*4b is t tal THERMAt POWER in percent of RATED THERMAL POWER, y (ii) for each percent that the magnitude of qt ~9 b exceeds -35%, the N-16 Trip Setpoint shall be automatically reduced by 1.25% of its value at RATED THERMAL POWER, and (iii) for each percent that the magnitude of q t gbexceeds +10%, the N-16 Trip Setpoint shall be automatically reduced by 1.55% of its value at RATED THERMAL POWER.

NOTE 2: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 1.4%

. of span.

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TXX-88512 ATTACHMEH1 2 PAGE 12 0F 20 '

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1 BASES FOR SECTION 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS NOTE The BASES contained in succeeding pages sumarize the reasons for the Specifications in Section 2.0, but in accordance with 10 CFR 50.36 are not part of these Technical Specifications.

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TXX-88512 :

ATTACHMENT 2 PAGE 13 0F 20 P ,

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2.1 SAFETY LIMITS

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V BASES -

2.1.1 REACTOR CORE The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is pre-vented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface tempera-ture is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate b0iling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. The local DNB heat flux ratio (DNBR) is defined as the ratio of the heat flux that would cause ONB at a particular core location to the local heat flux and is inoicative of the margin to ONB. DNBR is not a directly measurable parameter during operation and therefore THERMAL POWER and reactor coolant temperature and pressure have been related to ONBR through the W-3 correlation. The W-3 DNS correlation has been developed to predict the ONB flux and the location of DNB for axially uniform and nonuniform heat flux distributions.

The minimum value of the ONBR during steady-state operation, normal operational transients, and anticipated transients is limited to 1.30. This value corresponds to a 95% probability at a 95% confidence level that DNB will not occur and is chosen as an appropriate margin to ONB for all operating conditions.

The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum ONBR is no less than 1.30, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid.

These curves are based on a nuclear enthalpy rise hot channel factor, F H' of 1.55 and a reference cosine with a peak of 1,55 for axial power shape. An allowance is included for an ircrease in F"" 0 at reduced power based on the l expression:

F = 1. 55 (1+ 0. 2 (1-P)]

Where P is the fraction of RATED THERMAL POWER.

These heat flux conditions are more limiting than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance is within the limits of the ft (a!) function of the Overtemperature N-16 trip. When the axial power imbalance is not within the tolerance, the axial power imbalance effect on the Overtemperature N-16 trips will reduce the Setpoints to provide protection v consistent with core Safety Limits.

l l COMANCHE PEAK - UNIT 1 8 2-1

IXI-88512 AliACHMENT2 l PAGE 14 0F 20 r* I t

SAFETY LIMITS , ,

t3ASES l

2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System (RCS) from overpressurizatinn 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%

TheSafetyLimitof2735psigisthereforeco(2735psig)ofdesignpressure.

nsistent with the design criteria and associated Code requirements.

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

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IIX-88512 AliACHMENT2 PAGE 15 0F 20 2.2 LIMITING SAFETY SYSTEM SETTINGS n

b BASES -

2.2.1 REACTCE TRIP SYSTEM INSTRUMENTATION SETPOINTS The Reactor Trip Setpoint Limits specified in Table 2.2-1 are the nominal values at which the Reactor trips are set for each functional unit. The Trip Setpoints have been selected to ensure that the core and Reactor Coolant System are prevented from exceeding their safety limits during normal operation and design basis anticipated operational occurrences and to assist the Engi-neered Safety Features Actuation System in mitigating the consequences of accidents. The setpoint for a Reactor Trip System or interlock function is considered to be adjusted consistent with the nominal value when the "as measured" setpoint is within the band allowed for calibration accuracy and instrument drift.

To accommodate the instrumer.t drift assumed to occur between operational tests and the accuracy to which setpoints can be measured and calibrated, Allowable Values for the Reactor Trip Setpoints have been specified in Table 2.2-1. Operation with setpoints less conservative than the Trip Set-point but within the Allowable Value is acceptable since an allowance has been made in the safety analysis to accommodate this error. An optional provision has been included for determining the OPERABILITY of a channel when its Trip Setpoint is found to exceed the Allowable Value. The methodology of this l

option utilizes the "as measured" deviation from the specified calibration l point for rack and sensor components in conjunction with a statistical combin-ation of the other uncertainties of the instrumentation to measure the process variable and the uncertainties in calibrating the instrumentation. In Equa-tion 2.2-1, Z + R + 5 < TA, the interactive effects of the errors in the rack and the sensor, and the "as measured" values of the errors are considered. Z, as specified in Table 2.2-1, in percent span, is the statistical summation of I errors assumed in the analysis excluding those associated with the sensor and

. rack drift and the accuracy of their measurement. TA or Total Allowance is the difference, in percent span, between the Trip Setpoint and the value used in the analysis for Reactor trip. R or Rack Error is the "as measured" devia-tion, in percent span, for the affected channel from the specified Trip Set-point. 5 or Sensor Error is either the "as measured" deviation of the sensor from its calibration point or the value specified in Table 2.2-1, in percent span, from the analysis assumptioris. Use of Equation 2.2-1 allows for a sensor drift factor, an increased rack drift factor, and provides a threshold value for REPORTA8LE EVENTS.

The methodology to derive the Trip Setpoints is base.d upon combining all of the uncertainties in the channels. Inherent to the determination of the Trip Setpoints are the magnitudes of these channel uncertainties. Sensors and other instrumentation utilized in these channels are expected to be capable of l

operating within the allowances of these uncertainty magnitudes. Rack drift in excess of the Allowable Value exhibits the behavior that the rack has not met its allowance. Being that there is a small statistical chance that this will happen, an infrequent excessive drift is expected. Rack or sensor drift, in excess of the allowance that is more than occasional, may be indicative of

( more serious problems and should warrant further investigation.

l COMANCHE PEAK - UNIT 1 B 2-3 l

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'XX-88512 AllACMENT 2 PAGE 16 0F 20 j

LIMITING SAFETY SYSTEM SETTINGS 8-> -

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D BASES

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REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS (Continued)

The various Reactor trip circuits automatically open the Reactor trip breakers whenever a condition monitored by the Reactor Trip System reaches a preset or calculated level. In addition to redundant channels and trains, the design approach provides a Roactor Trip System which monitors nuinerous system variables, therefore providing Trip System functional diversity. The functional capability at the specified trip setting is required for those anticipatory or diverse Reactor trips for which no direct credit was assumed in the safety analysis to enhance the overall reliability of the Reactor Trip System. The Reactor Trip System initiates a Turbine trip signal whenever Reactor trip is initiated. This prevents the insertion of positive reactivity that would otherwise result from excessive Reactor Coolant System cooldown and thus avoids unnecessary actuation of the Engineered Safety Features Actuation System.

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

't Power Range, Neutron Flux G

In each of the Power Range Neutron Flux channels there are two independent bistables, each with its own trip setting used for a High and Low Range trip setting. The Low Setpoint trip provides protection durir.g subcritical and low power operations to mitigate the consequences of a power excursion beginning from low power, and the High Setpoint trip provides protection during power operations to mitigate the consequences of a reactivity excursion from.all power levels.

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

Power Range. Neutron Flux, High Rates

. The Power Range Positive Rate trip provides protection against rapid flux increases which are characteristic of a rupture of a control rod drive housing.

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

The Power Range Negativ Rate trip provides protection for control rod drop accidents. At high power a single or multiple rod drop accident could cause local flux peaking which could cause an unconservative local DNBR to exist. The Power Range Negative Rate trip will prevent this from occurring by tripping the reactor. No credit is taken for operation of the Power Range Negative Rate trip for those control rod drop accidents for which DNBRs will be greater than 1.30.

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IXX-88512 AliACHMENT2 5m.

PAGE 17 0F 20 m LIMITING SAFETY SYSTEM SETTINGS A

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BASES Intermediate and Source Ranae, Neutron Flux The Intermediate and Source Range, Neutron Flux trips provide core protection during reactor startup to mitigate the consequences of an uncon-trolled rod cluster control assembly bank withdrawal from a subcritical condition. These trips provide redundant protecti.on to tne Low Setpoint trip of the Power Range, Neutron Flux channels. In addition, the Source Range Neutron Flux trip provides similar protection during shutdown operations with the reactor trip breakers closed and the rod control system capable of control rod withdrawal. The Source Range channels will initiate a Reactor trip at about 105 counts per second unless manually blocked when P-6 becomes active. The Intermediate Range channels will initiate a Reactor trip at a current level equivalent to approximately 25% of RATED THERMAL POWER unless manually blocked when P-10 becomes active.

Overtemocrature N-16 The Overtemperature N-16 trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power dis-tribution, provided that the transient is slow with respect to piping transit delays from the core to the N-16 detectors, and pressure is within the range O

V between the Pressurizer High and Low Pressure trips. The setpoint is auto-matically varied with: (1) coolant temperature to correct for temperature induced changes in density and heat capacity of water and includes dynamic compensation for piping delays from the core to the cold leg temperature detectors, (2) pressurizer pressure, and (3) axial power distribution. With normal axial power distribution, this Reactor trip limit is always below the core Safety Limit as shown in Figure 2.1-1. If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the Reactor trip is automatically reduced according to the notations in Table 2.2-1.

Overpower N-16

^

The Overpower N-16 trip provides assurance of fuel integrity (e.g., no fuel pellet melting and less than 1% cladding strain) under all possible over-power conditions, limit 9 the required range for Overtemperature trip, and pro-vides a backup to the High Neutron Flux trip. The Overpower N-16 trip

. provides protection to mitigate the consequences of various size steam breaks as reported in WCAP-9226, "Reactor Core Response to Excessive Secondary Steam Releases."

Pressurizer pressure In each of the pressurizer pressure channels, there are two independent bistables, each with its own trip setting to provide for a High and Low Pres-sure trip thus limiting the pressure range in which reactor operation is permitted. The Low Setpoint trip protects against low pressure which could lead to DNB by tripping the reactor in the event of a loss of reactor coolant pressure.

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TH 88512 ATTACHMENT 2 PAGE 18 0F 20 LIMITING SAFETY SY& TEM SETTINGS f%

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8ASES Pressurizer Pressure (Continued)

On decreasing power the Low Setpoint trip is automatically blocked by P-7 (a power level of approximately 10% of RATED THERMAL POWER with turbine first stage chamber pressure at approximately 10% of full power equivalent); and on increasing power, automatically reinstated by P-7.

The High Setpoint trip functions in conjunction with the pressurizer relief and rafety valves to protect the Reactor Coolant System against system overpressure.

Pressurizer Water Level The Pressurizer Water Level-High trip is provided to prevent water relief through the pressurizer safety valves. On decreasing power the Pres-surizer High Water Level trip is automatically blocked by P-7 (a power level of approximately 10% of RATED THERMAL POWER with a turbine first stage chamber pressure at approximately 10% of full power equivalent); and on increasing power, automatically reinstated by P-7.

,f q Reactor Coolant Flow i i C/ The Reactor Coolant Flow-Low trip provides core protection to prevent DNB l by mitigating the consequences of a loss of flow resulting from the loss of one or more reactor coolant pumps.

On increasing power above P-7 (a power level of approximately 10% of RATED THERMAL POWER or a turbine first stage chamber pressure at approximately 10% of full power equivalent), an automatic Reactor trip will occur if the flow in more than one loop drops below 90% of nominal full loop flow. Above P-8 (a power level of approximately 48% of RATED THERMAL POWER) an automatic l Reactor trip will occur if the flow in any single loop drops below 90% of l nominal full loop flow. Conversely, on decreasing power between P-8 and %e-l P-7 an automatic Reactor trip will occur on low reactor coolant flow in more than one loop and below P-7 the trip function is automatically blocked.

j Steam Generator Water Level The Steam Generator Water Level Low-Low trip protects the reactor from loss of heat sink in the event of a sustained steam /feedwater flow mismatch I

resulting from loss of normal feedwater. The specified setpoint provides allowances for starting delays of the Auxiliary Feedwater System.

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TXX-88512 AfiACHMENT2 PAGE 19 0F 20 '

LIMITING SAFETY- SYSTEM SETTINGS 6 BASES -

I Undervoltage and Underfrequency - Reactor Coolant Pump Busses The Undervoltage and Underfrequency Reactor Coolant Pump Bus trips pro-vide core protection against DNB as a result of complete loss of forced coolant flow. The specified setpoints assure a Reactor trip signal is generated before the Low Flow Trip Setpoint is reached. Time delays are incorporated in the Underfrequency and Undervoltage trips to prevent spurious Reactor trips from momentary electrical power transients. For undervoltage, the delay is set so that the time required for a signal to reach the Reactor trip breakers following the simultaneous trip of two or more reactor coolant pump bus circuit breakers shall not exceed 1.2 seconds. For underfrequency, the delay is set so that the time required for a signal to reach the Reactor trip breakers after the Underfrequency Trip Setpoint is reached shall not exceed 0.3 second.

On decreasing power the Undervoltage and Underfrequency Reactor Coolant Pump Bus trips are automatically blocked by P-7 (a power level of approximately 10%

of RATED THERMAL POWER with a turbine first stage chamber pressure at approximately 10% of full power equivalent); and on increasing power, reinstated automatically by P-7.

Turbine Trip A Tu'rbine trip initiates a Reactor trip. On decreasing power the Reactor l trip from the Turbine trip is automatically blocked by P-7 (a power level of approximately 10% of RATED THERMAL POWER); and on increasing power, reinstated automatically by P-7.

Safety Injection input from ESF If a Reactor trip has not already been generated by the Reactor Trip System instrumentation, the ESF automatic actuation logic channels will initiate a Reactor trip upon any signal which initiates a Safety Injection. The ESF instrumentation channels which initiate a Safety Injection signal are shown in Table 3.3-3.

I l

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IIX 'J512 l AliACHMENT2 PAGE 20 0F 20 QMITINGSAFETYSYSTEMSETTINGS l

BASES I Undervoltage and Underfrequency - Reactor Coolant Pumo Busses (Continued)

Reactor Trip System Interlocks

, The Reactor Trip System interlocks perform the following functions:

P-6 On increasing power P-6 allows the manual block of the Source Range trip (i.e. , prevents premature block of Source Range trip), provides a backup block for Source Range Neutron Flux doubling, and deener-gizes the high voltage to the detectors. On decreasing power, Source Range Level trips are automatically reactivated and high voltage restored.

P-7 On increasing power P-7 automatically enables Reactor trips on low flow in more than one reactor coolant loop, reactor coolant pump bus undervoltage and underfrequency, turbine trip, pressurizer low pressure and pressurizer high level. On decreasing power, the above listed trips are automatically blocked.

P-8 Onincreasingpower,P-8automaticallyenablestheReactortrip/on low flow in one reactor coolant loop. On decreasing power, the P-8 automatically blocks the reactor trip on low flow in one reactor coolant loop.

P-10 On increasing power, P-10 allows the manual block of the Intermediate Range trip and the Low Setpoint Power Range trip; and automatically blockstheSourceRangetripandde[energizestheSourceRangehigh voltage power. On decreasing power, the Intermediate Range trip and the Low Setpoint Power Range trip are automatically reactivated.

Provides input to P-7.

P-13 Turbine first stage chamber pressure provides input to P-7.

COMANCHE PEAK - UNIT 1 8 2-8 ,

1X1-88512 AliACHME%T 3 PACE 1 0F 12 O .

COMAtlCHE PEAK STEAM ELECTRIC STATION TECHNICAL SPECIFICATION 3/4.0 O

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l ATTAcm a r 3 CPSES Technical Spec'ifications PAGE 2 0F 12 NRC Draft 2 Markup Section 3/4.0 4

l' Q Change 10# Justification For Change 0014 Added the "exception clause" to Specifications 4.0.2 and 4.0.4 for consistency as it is already contained in-3.0.3 l and 3.0.4. All of these S)ecifications are excluded from applicability throughout tie Technical Specifications at one point or another.

l 0923 This clarification, relating to voluntary entry into ACTION Statements is to make it clear that even though voluntary entry into the allowed outage times are routinely required for surveillance and maintenance activities, the shutdown portion of the ACTION Statement j should not be used for operational convenience.

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IXX-88512 . .

. AliACHMENT 3 PAGE 3 0F 12 SECTIONS 3.0 AND 4.0 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS O<

O COMANCHE PEAK - UNIT 1 3/4 0-0

l TXX-88512 I AliACMENT3 i PAGE 4 Of 12 l 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.0 APPLICABILITY d,

LIMITING CONDITION FOR OPERATION l

1

3. 0.1 Compliance with the Limiting Conditions for Operation contained in the i succeeding specifications is required during the OPERATIONAL MODES or other i conditions specified therein; except that upon failure to meet the Limiting I Conditions for Operation, the associated ACTION requirements shall be met.

3.0.2 Noncompliance with a specification shall exist when the requirements of the Limiting Condition for Operation and associated ACTION requirements are not met within the specified time intervals. If the Limiting Condition for Operation is restored prior to expiration of the specified time intervals, completion of the ACTION requirements is not required.

3.0.3 When a Limiting Condition for Operation is not met, except as provided in the associated ACTION requirements, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months action shall be initiated to place the unit in a MODE in which the specification does not apply by placing it, as applicable, in:

a. At least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,

b. At least HOT SHUT 00WN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and

c. At least COLD SHUTOOWN within the subsequent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

i i*

\ Where corrective measures are completed that permit operation under the ACTION requirements, the action may be taken in accordance with the specified time limits as measured from the time of failure to meet the Limiting Condition for Operation. Exceptions to these requirements are stated in the individual specifications.

This specification is not applicable in MODE 5 or 6.

3.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made when the conditions for the Limiting Conditions for Operation are not met and the associated ACTION requires a shutdown if they are not met within a specified time interval. Entry into an OPERATIONAL MODE or specified condition may be made in accordance with ACTION requirements when conformance to them permits c ntinued operation of the facility for an unlimited period of time. This provision shall not prevent passage through or to OPERATIONAL MODES as required to comply with ACTION requirements. Exceptions to these requirements are stated in the individual specifications.

O COMANCHE PEAK - UNIT 1 3/4 0-1

I TXX 88512 AllACHMENT3 PAGE 5 0F 12 APPLICA8ILITY *~' ~

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SURVEILLANCE REQUIREMENTS 4.0.1 Surveillance Requirements shall be met during the OPERATIONAL MODES or other conditions specified for individual Limiting Conditions for Operation unless otherwise stated in an individual Surveillance Requirement.

4.0.2 Each Surveillance Requirement shall be performed within the specified time interval with:

a. A maximum allowable extension not to exceed 25% of the surveillance interval, but

b. The combined time interval for any three consecutive surveillance intervals shall not exceed 3.25 times the specified surveillant:e interval. I Emperu c te h e-v ~h m. SL ded la N O'M SMd cd 'b 5 - ID 1: 0014 4.0.3 Failure to perform a Surveillance Requirement within the allowed sur- l veillance interval, defined by Specification 4.0.2, shall constitute noncom-pliance with the OPERABILITY requirements for a Limiting Condition for Operation. The time limits of the ACTION requirements are applicable at the time it is identified that a Surveillance Requirement has not been performed.

The ACTION requirements may be delayed for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to permit the com-pletion of the surveillance when the allowable outage time limits of the ACTION requirements are less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Surveillance Requirements do not i

O,I have to be performed on inoperable equipment, 4.0.4 Entry into an OPERATIONAL N00E or other specified condition shall not be made unless the Surveillance Requirement (s) associated with the Limiting Condition for Operation has been performed within the stated surveillance interval or as otherwise specified. This provision shall not prevent passage through or to OPERATIONAL MODES as required to comoly with ACTION requirarments. l Euepa m to tWe repu*-h u s W eJ a M .ndN.JM c p m c @ m e. W 00H 4.0.5 Surveillance Requirements for inservice inspection and testing of ASME l

l Code Class 1, 2, and 3 components shall be applicable as follows:

a. Inservice inspection of ASME Code Class 1, 2, and 3 components and t

inservice testing of ASME Code Class 1, 2, and 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50.55a(g), except whert specific written relie/ has been

, granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i);

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IIX48512 ATTACHMENT 3 PAGE 6 0F 12

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APPLICABILITY

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SURVEILLANCE REQUIREMENTS (Continued)

b. Surveillance intervals specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda for the inservice inspection and testing activities required by the ASME Boiler and Pressure Vessel Code and applicable Addenda shall be applicable as follows in these Technical Specifications:

ASME Boiler and Pressure Vessel Required frequencies for Code and applicable Addenda performing inservice terminology for inservice inspection and testing inspection and testing activities activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days

c. The provisions of Specification 4.0.2 are applicable to the above required frequencies for performing inservice inspection and testing activities;

(')N

(

d. Performance of the above inservice inspection and testing activities shall be in addition to other specified Surveillance Requirements; and

e. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any Technical Specification.

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(N COMANCHE PEAK - UNIT 1 3/4 0-3 ,

IXX-88512 '

. AITACHMEWI3 PAGE 7 Of 12

(- 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS M- -

3/4.0 APPLICABILITY BASES Specification 3.0.1 through 3.0.4 establish the general requirements applicable to Limiting Conditions for Operation. These requirements are based on the re-quirements for Limiting Conditions for Operation statec: in 10 CFR 50.36(c)(2):

"Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility.

When a limiting condition for operation. of a nuclear roactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specification until the condition can be met."

Specification 3.0.1 establishes the Applicability statement within each indi-vidual specification as the requirement for when (i.e., in which OPERATIONAL MODES or other specified conditions) conformance to the Limiting Conditions for Operation is required for safe operation of the facility. The ACTION requirements establish those remedial measures that must be taken within speci-fied time limits when the requirements of a Limiting Condition for Operation are not met.

There are two basic types of ACTION requirements. The first specifies the remedial measures that permit continued operation of the facility which is not further restricted by the time limits of the ACTION requirements. In this p case, conformance to the ACTION requirements provide an acceptable level of V safety for unlimited continued operation as long as the ACTION requirements continue to be met. The second type of ACTICN requirement specifies a time limit in which conformance to the conditions of the Limiting Condition for Operation must be met. This time limit is the allowable outage time to restore an inoperable system or component to OPERAGLE status or for restoring paramsters within specified limits. If these actions are not completed within the allow-able outage time limitsg a shutdown is required to place the facility in a MODE or condition in which the specification no longer applies. It is.not intended that the shutdown. CTION requirements be used as an operational con- IDI 0923 venience which permits (rou ne) v0 krtary r;;cval of a system (s) or compo-l nent(s) from service in lisu..of other alternatives \that would not result in redundant systems or components being inoperable. \_e.degen of ec deced.

\Qo ch ton M 6e CO M G The specified time limits of the ACTION requirements are applicaole from tne point in time it is identified that a Limiting Condition 3e Operation is not met. The time limits of the ACTION requirements are als' 3iplicable when a system or component is removed from service for surveillance testing or investi-gation of operational problems. Individual specifications may include a speci-find time limit for the completion of a Surveillance Requirement when equipment is removed from service. In this case, the allowable outage time limits of the ACTION requirements are applicable when this limit expires if the surveillance has not been completed. When a shutdown is required to comply with ACTION requirements, the plant may have entered a MODE in which a new specification becomes applicable. In this case, the time limits of the ACTION requirements would apply from the point in time that the new specification becomes applicable if the requirements of the Limiting Condition for Operation are not met.

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TXX88512

. ATTACHMENT 3 PAGE 8 0F 12 APPLICABILITY ,

BASES Specification 3.0.2 establishes that noncompliance with a specification exists when the requirements of the Limiting Condition for Operation are not met and the associated ACTION requirements have not been implemented within the speci-fied time interval. The purpose of this specification is to clarify that (1) implementation of the ACTION requirements within the specified time interval constitutes compliance with a specification and (2) completion of the remedial measures of the ACTION requirements is not required when compliance with 2 Limiting Condition of Operation is restored within the time interval specified in the associated ACTION requirements.

Specification 3.0.3 establishes the shutdown ACTION requirements that must be imolemented when a Limiting Condition for Operatior, is not met and the condi-tion is not specifically addressed by the associated ACTION requirements. The puroose of this specification is to delineate the time limits for placing the unit in a safe shutdown MODE when plant operation cannot be maintained within the limits for safe operation defined by the Limiting Conditions for Operation and its ACTION requirements. It is not intended to be used as an operational convenience which permits (routine) voluntary removal of redundant systems or components from service in lieu of other alternatives that would not result in redundant systems or components being inoperable. One hour is allowed to pre-pare for an orderly shutdown before initiating a change in plant operation.

This time permits the operator to coordinate the reduction in electrical genera-

, [sv) ,

tion with the load dispatcher to ensure the stability and availability of the electrical grid. The time limits specified to reach lower MODES of operation permit the shutdown to proceed in a controlled and orderly manner that is well within the specified maximum cooldown rate and within the cooldown capabilities of the facility assuming only the minimum required equipment is OPERABLE.

This reduces thermal stresses on components of the primary coolant system and the potential for a plant upset that could challenge safety systems under con-ditions for which this specification applies.

If remedial measures permitting limited continued operation of the facility under the provisions of the ACTION requirements are completed, the shutdown may be terminated. The time limits of the ACTION requirements are applicable from the point in time there was a failure to meet a Limiting Condition for l

Operation. Therefore, the shutdown may be terminated if the ACTION require-l

ments have been met or the time limits of the ACTION requirements have not

! expired, thus providing an allowance for the completion of the , required actions.

The time limits of Specification 3.0.3 allow 37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months for the plant to be in the COLD SHUTDOWN MODE when a shutdown is required during'the POWER MODE of coeration. If the plant is in a lower MODE of operation when a shutdown is required, the time limit for reaching the next lower MODE of operation applies.

However, if a lower MODE of operation is reached in less time than allowed, j

the total allowable time to reach COLD SHUTDOWN, or other applicable MODE, is not reduced. For example, if HOT STANOBY is reached in 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , the time allowed to reach HOT SHUTDOWN is the next 11 hours1.273148e-4 days 0.00306 hours 1.818783e-5 weeks 4.1855e-6 months because the total time to l

O reach HOT SHUTDOWN is not reduced from the allowable limit of 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months .

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IIX-C'512

. ATTACHMRT3 PAGE 9 0F 12 APPLICABILITY j BASES Therefore, if remedit.1 measures are completed that would permit a return to ;

POWER operation, a penalty is not incurred by having to reach a lower MODE of l operation in less than the total time allowed.

The same principle applies witti regard to the allowable outage time limits of the ACTION requirements, if compliance with the ACT.10N requirements for one specification results in entry into a MODE or condition of operation for another specification in which the requirements of the Limiting Condition for Operation are not met. If the new specification becomes applicable in less time than specified, the differen:e may be added to the allowable outage time limits of the secend specification. However, the allowable outage time limits of ACTION requirements for a higher MODE of operation may not be used to extend the allowable outage time that is applicable when a Limiting Condition for Operation is not met in a lower MODE of operation.

The shutdown requirements of Specification 3.0.3 do not apply in MODES 5 and 6, because the ACTION requirements of individual spr.:ifications define the remedial measures to be taken.

Specification 3.0.4 establishes limitations on MODE changes when a Limiting Condition for Operation is not met. It precludes placing the facility in a ;

higher MODE of operation when the requirements fur a Limiting Condition for -

O' ( Operation are not met and continued noncompliance to these conditions would result in a shutdown to comply with the ACTION requirements if a change in MODES were permitted. The purpose of this specification is to ensure that facility operation is not initiated or that higher MODES of operation are not entered when enrective action is being taken to obtain compliance with a speci-fication by restoring equipment to OPERABLE status or parameters to specified

, limits. Compliance with ACTION requirements that permit continued operation of the facility for an unlimited period of time provides an acceptable level

l. of safety for continued operation without regard to the status of the piant i before or after a MODE change. Therefor 0, in this case, entry into an OPERATIONAL MODE or other specified condition may be made in accordance with the provisions of the ACTION requiremtnts. The provisions of this specification should not, however, be interpreted av endorsing the failure to exercise good practice in restoring systems or components to OPERABLE status before plant i startup.

I When a shutdowa is required to comply with ACTION requirements, the provisions i of Specification 3.0.4 do not apply because they would delay placing the facil-ity in a lower MODE of operation.

l Specifications 4.0.1 through 4.0.5 establish the genera equiremants applicable to SurveTilence Requirements. Inise requirements are based on the Surveillance Requirments stated in the Code of Federal Regulations,10 CFR 50.36(c)(3):

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TXX-88512 ATTACHMENT 3 PAGE 10 Of 12 APPLICABILITY ,

BASES "Surveillance requir -

cs are requirements relating to test, calibration, or inspection to ensure ;. the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting condition:, of operation will be met."

Specification 4.3.1 establishes the requirement that surveillances must be performed during the OPERATIONAL MODES or other conditions for which the re-quirements of the Limiting Conditions for Operation apply unless otherwise stated in an individual Surveillance Requirement. The purpose of this specifi-cation is to ensure that surveillances are performed to verify the operational status of systems and compo.,ents and that parameters are within specified limits i to ensure safe optration of the facility when the plant is in a MODE or other specified conditic~ for which the associated Limiting Conditions for Operation are applicable. S.:rveillance Requirements do not have to be performed when the facility is in an OPERATIONAL MODE for which the requirements cf +he asso-ciated Limiting Condition for Operation do not apply unless otherwise specified.

The Surveillance Requirements associated with a Special Test Exception are only applicable when the Special Test Exception is used as an allowable excep-tion to the requirements of a specification.

s Specification 4.0.2 establishes the conditions under which the specified time interval for Surveillance Requirements may be extended. Iterr a. permits an i

allowable extension of the normal surveillance interval to facilitate surveil-lance scheduling and consideration of plant operating conditions that may not be suitable for conducting the surveillance, e.g. , transiene conditions or other ongoing surveillance or maintenance activities. Item b. limits the use of the provisions of item a. to ensure that it is not used repeatedly to extend the surveillance interval beyond that specified. The limits of Specification 4.0.2 are based on engineering judgment and the recognition that the most prob-able result of any particular surveillance being performed is the verificaion of conformance with the Surveillance Requirements. These provisions are suf-ficient to ensure that the reliability ensured through surveillance activities is not significantly degraded beyond that obtained from the specified surveil-lance interval.

Decification 4.0.3 establishes the failure to perforra a Surveillance Require-merit within the allowed surveillance interval, defined by the provisions of Spacification 4.0.2, as a condition that constitutes a failure to meet the OPERABILITY requirements for a Limiting Condition for Operation. Under the provisions of this specification, systems and components are assemed to be l OPERABLE when Surveillance Requirements have been satisfactorily performed I within the specified time interval. However, nothing in this provision is to be construed as implying that systems or components are OPERABL6 wbsn they are i found or known to be inoperable although still meeting the Surveillance Require-This specification also clarifles that the fC90N requirements are ments.

applicable when Surveillance Requirements have not been cornpleted within the l

' 3 allowed surveillance interval and that the time limits of the ACTION require-ments apply from the point in time it is identifi!d that a surveillance has not been performeo and not at the time that the allowed surveillance interval COMANCHE PEAK - UNIT 1 is 3/4 0-4 .

T w 88512 ATTACHMENT 3

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PAGE !! 0F 12

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FY APPLICABILITV BASES was exceeded. Completion of the Surveillance Requirement within the allowable outage time limits of the ACTION requirements restores compliance with the requirements of Specification 4.0.3. However, this does not negate the fact that the failure to have performed the surveillance within the allowed surveil-lance interval, defined by the provisions of Specification 4.0.2, was a viola-tion of the OPERABILITY requirements of a Limiting Condition for Operation that is subject to enforcement action. Further, the failure to perform a sur-v dllance within the provisions of Specification 4.0.2 is a violation of a Technical Specification requirement and is, therefore, a reportable event under the requirements of 10 CFR 50.73(a)(2)(i)(B) because it is a condition pro-hibited by the plant's Techr.ical Specifications.

If the allowable outage time limits of the ACTION requirements are less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or a shutdown is required to comply with ACTION requirements, e.g.,

Specification 3.0.3, a 24-hour al U wance is provided to permit a delay in implementing the ACTION requirement.s. This provides an adequate time limit to complete Surveillance Requirements that have not been performed. The purpose of this allowance is to permit the comp!etion of a surveillance before a shutdown is required to comply with ACTION requirenents or before other remedial measures would be required that may preclud completion of a surveil-lance. The basis for this allowance includes consid: ration for plant condi-tions, adequate planning, availability of personnel, the time required to perform the surveillance, and the safety significance of the delay in complet-( ing the required surveillance. This provision also provides a time limit for the completion of Surveillance Requirements that become applicable as a consequence of MODE changes imposed by ACTION requirements and for completing Surveillance Requirements that are applicable when an exception to the requirements of Specification 4.0.4 is allowed. If i surveillance is not completed within the 24-hour allowance and the Surveillance Requirements are not met, the time limits of the ACTION requirements are ap;.licable at the time that the surveillance is terminated.

Surveillance Requirements do not have to be performed on inoperable equipment because the ACTION requirements define the remedial measures that apply.

However, the Surveillance Requirements have to be met to demonstrate that inoperable equipment has been restored to OPERABLE status.

Specification 4.0.4 establishes the requirement that all applicable surveil-lances must be met before entry into an OPERATIONAL MODE or other condition of operation specified in the Applicability statement. The purpose of this speci-fication is to ensure that system and component OPERABILITY requirements or parameter limits are met before entry into a MODE or condition for which these systems and components ensure safe operation of the facility. This provision applies to changes in OPERATIONAL MODES or cther specified conditions associated with plant shutdown as well as startup.

Under the provisions of this specification, the applicable Surveillance p Requirements must be performed within the specified surveillance interval to

( ensure that the. Limiting Conditions for Operation are met during initial plant startg or following a plant outage.

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IXX-68512 i ATTACHMENT 3 FAGE 12 0F 12 I

p APPLICABILITY V

BASES i When a stut is required to comply with ACTION requirements, the provisions of Specification 4.0.4 do not apply because this would delay placing the facility in a lower MODE of operation.

Specification 4.0.5 estab!'shes the requirement that inservice inspection of ASME Code Class 1, 2, ano 3 components and inservice testing of ASME Code Class 1, 2, and 3 pumps and valves shall be performed in accordance with a periodically updated version of Section XI of the ASME Boiler and Pressure Vessel Code and Addenda as required by 10 CFR 50.55a. These requirements apply except when relief has been provided in writing by the Commission.

This specification includes a clarification of the frequencies for performing the inservice inspection and testing activities required by Section XI for the ASME Boiler and Pressure Vessel Code and applicable Addenda. This clari-fication is provided to ensure consistency in surveillance intervals through- ,

out the Technical pecifications and to remove any ambiguities relative to the l frequencies for performing the required inservice in:pection and testing i activities, l Under the terms of this specification, the more restrictive requirements of the Technical Specifications take precedence over the ASME Boiler and Pressure Vessel Code and applicable Addenda. The requirements of Specification 4.0.4 to perform surveillance activities before entry into an OPERATIONAL MODE or .

I other specified condition takes precedence over the ASME Boiler and Pressure Vessel Code provision which allows pumps and valves to be tested up to one week after return to normal operation. The Technical Speci.fication definition of OPERABLE does not allow a grace period before a component, that is not capable of performing its specified function, is declared inoperable and takes precedence over the ASME Boiler and Pressure Vessel Code provision which allows a valve to be incapable of performing its specified function for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months before being declared inoperable.

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IXX-88512

)

AffACHRENT4 PAGE i 0F 29 O

r COMANCHE FEAX STEAM ELECTRIC STATION l

TECHNICAL SPECIFICATION 3/4.1 0

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TXX88512 ATTACH s i 4 CPSES Technical Specifications PAGE 2 of 29 NRC Draft 2 Markup Section 3/4.1 Change 10# Justification For Change

-(G'T 0037 This change is to address multiple immovedle, but 0038 trippable, control rods, she Standard Technical 0454 Specifications as currently written (Rev. 4) do not recognize the fact that in this situation the control rods would still perform their safety function and because more than one control rod is immovable *.he plant is forced to repair the failure or be in Hot Shutdown in six hours. An action this drastic is unnecessary since the proposed changes to this specification adequately address the safety requirements with regards to immovable and misaligned control rods.

The Bates have been modified to reflect this change. This change is similar to that Licensed at South Texas, Vogtle, Seabrook and Millstone. This change was also recommended for incorporation into the STS by Westinghouse via letter NS-NRC-84-2990 dated December 21, 1984.

0041 This change is made to allow testing to prove Operability',

0042 after maintenance has been performed, on the Digital Rod Position Indicators (DRPI). The Standard Technical Specification does not recognize that if there is a

. problem with the DRPI (and maintenance is performed on the (N indicators that would require the rod to be pulled to

(,') prove Operability) there is no way to achieve this retest requirement under the present STS. The special conditions required by the change were taken from Specification 3.10.5 (LC0 notation) which will prevent an inadvertent criticality.

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TXX-11512 \

AliACHMOT 4 l PA E 3 0F 29 CPSES Technical Specification: !

NRC Oraft 2 Markup Section 3/4.1 (n) v Change ID# Justification For Change 1 0454 See 10# 0037 0924 The current note tied to this Surveillance Requirement is changed to ensure there is no confusion as to what is required to isolate the pump from the RCS. Without this change, it could be interpreted that the pump discharge valve has to be shut which is one way to accomplish isolation but not the only way. This change was approved in the meeting on 1/11/88 for Specification 3.5.3.1 and was ove, looked for this Specification.

0925 This surveillance requirement is being deleted to return the specification to Standard Technical Specifications (STS). This change was originally proposed in the October 1987 submittal to show compliance with the LC0. Th:s requirement is not in STS or other Licensed plants since the pump is sufficiently tested per ASME Section XI and the water source for this testing is the RWST, therefore~

the required flow path is verified by design.

0926 Added footnote to ensure the connection is made between the ECCS specification, 3.5.2 and this specification.

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l TXX 88512 !

AliACHi;EK!4 , ,

l PAGE 4 0F 29- - F l

p. 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUT 00WN MARGIN - T,y GREATER THAN 200'F LIMITING CONDITION FOR OPERAT'ON 3.1.1.1 The SHUTOOWN MARGIN shall be greater than or equal to 1.6% Ak/k.

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

ACTION:

With the SHUTDOWN MARGIN less than 1.6% ok/k, immediately initiate and con-tinue boration at greater than or ; qual to 30 gpm of a solution containing greater than or equal to 7,000 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored.

SURVEILLANCE REQUIREMENTS - 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1.6% ak/k:

a. Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after detection of an inoperable control rod (s) and i

at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereaf ter while the rod (s) is inoperable.

If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be verified acceptable with an' increased allowance for the withdrawn worth of the immovable or untrippable control rod (s);

b. When in MODE 1 or MODE 2 with K,ff greater than or equal to 1 at least coce per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that control bank withdrawal is

! within the limits of Specification 3.1.3.6;

c. When in MODE 2 with K,ff less than 1, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to achieving reactor criticality by verifying that the predicted I critical control rod position is within the limits of Specification j 3.1.3.6; ,

d. Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors of Specifica-tion 4.1.1.1.le, below, with the control banks at the maximum inser-tion limit of Specification 3.1.3.6; and

See Special Test Exceptions Specification 3.10.1.

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IXX-88512 ATTACHMENT 4 PAGE 5 0F 29 REACTIVITI CONTit0L SYSTEMS O SURVEILLANCE REQUIREMENTS (Continued)

e. When in MODE 3 or 4, at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by consideration of the following factors:

1) Reactor Coolant System boron concentration,

, 2) Control rod positien,

3) Reactor Coolant System average temperature,

4) Fuel burnup based on gross thermal energy generation,

5) Xenon concentration, and

6) Samarium concentration.

4.1.1.1.2 The overall core reactivity balance shall be compared to predicted values to demonstrate agreement within i 1% Ak/k at least once per 31 Effective Full Power Days (EFPD). This comparison shall consider at least those factors stated in Specification 4.1.1.1.le., above. The predicted reactivity values shall be adjusted (normalized) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 EFP0 O( after each fuel loading.

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IXX-88512 ATTACHliENT4 y PAGE 6 Of 29 .

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REACTIVITY CONTROL SYSTEMS 1 SHUT 00WN MARGIN - T,y LESS THAN OR EQUAL TO 200*F LIMITING CONDITION FOR OPERATION 3.1.1.2 The SHUTDOWN MARGIN shall be greater than or equal to 1% ak/k.

APPLICABILITY: MODE 5.

ACTION:

With the SHUT 00WN MARGIN less than 1% ak/k, immediately initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7,000 ppm boron or equivalent until the requirsd SHUTDOWN MARGIN is restored.

SURVEILLANCE REQUIREMENTS 4.1.1.2 The SHUTOOWN MARGIN shall be determined to be greater than or equal to 1% ok/k:

a. Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after detection of an inoperable control rod (s) and at O least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ',hereafter while the rod (s) is inoperable, b( If the inoperable control rod is immovable or untrippable, the SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod (s); and

b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by consideration of the following factors:

1) Reactor Cuolant System boron concentration,

2) Control rod position,

3) Reactor Coolant System average temperature,

. 4) Fuel burnup based on gross thermal energy generation,

5) Xenon concentration, and

6) Samarium concentration.

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, IXX-89512 ATTACHMENT 4 PAGE 7 0F 29 7s REACTIVITY CONTROL SYSTEMS t \

D MODERATOR TEMPERATURE COEFFICIENT I

LIMITING CONDITION FOR OPERATION 3.1.1. 3 The moderator temperature coefficient (MTC) shall be:

a. Less positive than 0 ok/k/*F for the &il rods withdrawn, beginning of cycle life (BOL), hot zero THERMAL POWER condition; and

b. Less negative than -4.0 x 10 4 ok/k/ F for the all rods withdrawn, end of cycle life (EOL), RATED THERMAL POWER condition.

APPLICABILITY: Specification 3.1.1.3a. - MODES 1 and 2* only**.

Specification 3.1.1.3b. - MODES 1, 2, and 3 only**.

ACTION:

a. With the MTC more positive than the limit of Specification 3.1.1.3a.

above, operation in MODES 1 and 2 may proceed provided:

1. Control rod withdrawal limits are established and maintained sufficient to restore the MTC to less positive than 0 ok/k/'F

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within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

These withdrawal limits shall be in addition to the insertion limits of Specification 3.1.3.6;

2. The control rods are maintained within the withdrawal limits established above until a subsequent calculation verifies that the MlC has been restored to within its limit for the all rods withdrawn condition; and

3. A Special Report is prepared and submitted to the Commiss' ion, pursuant to specification 6.9.2, within 10 days, describing the value of the measured MTC, the interim control rod withdrawal limits, and the predicted average core burnup necessary for restoring the positive MTC to within its limit for the all rods withdrawn condition.

b. With the MTC more negative than the limit of Specification 3.1.1.3b.

above, be in HOT SHUT 00WN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

With K,ff greater than or equal to 1.

- See Special Test Exceptions Specification 3.10.3.

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

TXX 88512

. AllACHMENT4 PAGE 8 0F 29 REACTIVITY CONTROL SYSTEMS O

SURVEILLANCE REQUIREMENTS 4.1.1.3 The MTC shall be determined to be within its limits during each fuel cycle as follows:

a. The MTC shall be measured and compared to the BOL limit of Specifi-cation 3.1.1.3a., above, prior to initial operation above 5% of RATED THERMAL POWER, after each fuel loading; and

b. The MTC shall be measured at any THERMAL POWER and compared to -3.1 x 10 4 ok/k/'F (all rods withdrawn, RATED THERMAL POWER condition) within 7 EFPD after reaching an equilibrium boron concentration of 300 ppm. In the event this comparison indicates the MTC is more negative than -3.1 x 10 4 Ak/k/'i, the MTC shall be remeasured, and compared to the EOL MTC limit of Specification 3.1.1.3b., at least once per 14 EFP0 during the remainder of the fuel cycle.

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IXX-88512 AITACHMENT4 PAGE 9 0F 29 REACTIVITY CONTROL SYSTEMS '

L 'IMUM TEMPERATURE FOR CRITICALITY LIMITING CONDITION FOR OPERATION 3.1.1.4 shall be greater The Reactor than Coolant or equalSystem lowest operating loop temperature (Tavg) to 551'F.

APPLICABILITY: .% DES 1 and 2* #.

ACTION:

With a Reactor Coolant System operating loop temperature (T,yg) less than 551 F, restore T,yg to within its limit within 15 minutes or be in HOT STANDBY within the next 15 minutes.

SURVEILLANCE REQUIREMENTS 4.1.1.4 The Reactor Coolant System temperature (T**9) shall be deter. mined to be greater than or equal to 551'F:

a. Within 15 minutes prior to achieving reactor criticality, and

b. At least once per 30 minutes when the reactor is critical and the Reactor Coolant System T,yg is less than 561 F with the T,yg-T ref Deviation Alarm not reset.

With K,ff greater than or equal to 1.

See Special Test Exceptions Specification 3.10.3.

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IXX-88512 AliACHMENT4 PAGE 10 0F 29 REAC1'IVITY CONTROL SYSTEMS 3/4.1.2 BORATI'ON SYSTEMS FLOW PATH - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.1 As a minimum, one of the following boron injection flow paths shall be OPERABLE and capable of being powered from an OPERABLE emergency power source:

a. A flow path from the boric acid storage tanks via either a boric acid transfer pump or a gravity feed connection and a charging pump to the Reactor Coolant System if thi boric acid storage tank in Specification 3.1.2.5a. is OPERABLE, or

b. The flow path from the refueling water storage tank via a centrifugal charging pump to the Reactor Coolant System if the refueling water storage tank in Specification 3.1.2.5b. is OPERABLE.

APPLICABILITY: MODES 5 and 6.

ACTION:

With none of the above flow paths OPERABLE or capable of being powered from an OPERABLE emergency power source, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

! SURVEILLANCE REQUIREMENTS

( 4.1.2.1 At least one of the above required flow paths shall ce demonstrated l OPERABLE:

a. At least once per 7 days by verifying that the temperature of the flow path is greater than or equal to 65'F when a flow path from the boric acid storage tanks is used, and

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

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TXX-88512 AITACHMENT 4 PAGE 11 Of 29 v

REACTIVITY CONTROL SYSTEMS

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FLOW PATHS - OPERATING LIMITING CONDITION FOR OrJRATION 3.1.2.2 At least two of the following three boron injection flow paths shall be OPERABLE:

a. The flow path from the boric acid storage tanks via either a boric acid transfer pump or a gravity feed connection and a charging pump to the Reactor Coolant System (RCS), and

b. Two flow paths from the refueling water storage tank via centrifugal charging pumps to the RCS.

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

ACTION:

With only one of the above required boron injection flow paths to the RCS OPERABLE, restore at least two boron injection flow paths to the RCS 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 SHUT 00WN 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 flow paths 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 REQUIRcMENTS l

4.1.2.2 At least two of the above required flow paths shall be demonstrated OPERABLE:

a. At least once per 7 days by verifying that the temperature of the flow path from the boric acid storage tanks is greater than .e equal to 65 F when it is a required water source;

b. 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;

c. At least once per 18 months by verifying that the flow path required l by Specification 3.1.2.2a. delivers at least 30 gpm to the RCS; and

d. At 1::st One: p:r 18 :onthsr-by-verffyhg-that-the flew path-requi-ced--

-by Sp;cif kation 3.1. 2.2b. S cap b1: Of deMver4ng-at-least-120-gpe-h

! ICI 0925

A maximum of two charging pumps shall be OPERABLE whenever the temperature of one or more of the RCS cold legs is less than or equal to 350'F. 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 $ ith power removed from the g O- valve operato$)or by a manual isolation valv$) secured in the closed position. g COMANCHE PEAK - UNIT 1 3/4 1-8 ,

a IXX-88512 AllACHMENT4 PAGE 12 Of 29 Q

ku REACTIVITY CONTROL SYSTEMS CHARGING PUMP - SHUTDOWN

,L3 ITING CONDITION FOR OPERATION 3.1.2.3 At least one charging pump in the boron injection flow path required by Specification 3.1.2.1 shall be OPERABLE and capable of being powered from an OPERABLE emergency power source.

APPLICABILITY: MODES 5 and 6.

ACTION:

With no charging pump OPERABLE or capable of being powered from an OPERABLE emergency power source, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

l SURVEILLANCE REQUIREMENTS lO 4.1.2.3.1 At least once per 92 days the above required charging pump shall be demonstrated OPERABLE by verifying that the flow path required by Specifica-tion 3.1.2.la is :apable of delivering at least 30 gpm to the RCS; or 4.1.2.3.2 At least once per 92 days by verifying that the flow path required by Specification 3.1.2.lb is capable of delivering at least 120 gpm to the RCS.

4.1.2.3.3 A maximum of two charging pumps shall be OPERABLE, one charging pump

, shall be der,onstrated inoperable

at least once per 31 days, except when the l reactor vessel head is removed, by verifying that the motor circuit breakers l are secure'l in the open position.

l l

l

An inoperable pump may be energized for testing provided the discharge of the pump has been isolated from tne RCS by a closed isolation valv4bith power removed from the valve operatorf)or by a manual isolation valve')p secured in the closed position.

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IDI 0924 l

COMANCHE PEAK - UNIT 1 3/4 1-9 ,

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IXX-88512 ATTACHMENT 4 PAGE 13 0F 29

,e REACTIVITY CONTROL SYSTEMS y CHARGING PUMPS '- OPERATING LIMITING CONDITION FOR OPERATION ce d ri. k el 3.1.2.4 At least two charging 3 pumps shall be OPERABLE.

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

ACTION:

With only one charging pump OPERABLE, restore at least !.wo 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 itiiin the next

, 7 days or be in COLD SHUT 00WN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS O)

L 4.1.2.4.1 The required centrifugal charging pump (s) shall be demonstrated OPERABLE by testing pursuant to Specification 4.1.2.3.2. 4.c r 4.1.2.4.2 The required positive displacement charging pump shall be #

demonstrated OPERABLE by testing pursuant to Specification 4.1.2.3.1.

4 0.C 1'

,4.1.2.4.3 Whena.ver 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, 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 g the pump has been isolated from the RCS by a closed isolation valve 6hith 8 power removed from the valve operato(t)or by a manual isolation valv@ secured 3 in the closed position, r* me bi 4 6h p c .. L c n d .p w e m.J p- e may a s.eJ k I;ew o f vn c es tu espid c.maci Lwi c.uqiq mg. m ora COMANCHE PEAK - UNIT 1 3/4 1-10 ,

__ y ..

. IXX-88512 ATTACHMENT 4 PAGE 14 0F 29 REACTIVITY C]f/ROL SYSTEMS

('--) 80 RATED WATdR SOURCE - SHUTDOWN LIMITING :0NDITION FOR OPERATION 3.1.2.5 As a minimum, one of the following borated water sources shall be OPERABLE:

a. A boric acid storage tank with:

1) A minimum contained borated water volume of [6385] gallons,

([Later]% of span), when using the boric acid transfer pump.

2) A minimum contained borated water volume of 15,123 gallons

([Later]% of span), when using the gravity feed connection,

3) A minimum boron concentration of 7000 ppm and

4) A minimum solution temperature of 63*F.

b. The refueling water storage tank (RWST) with:

1) A minimum contained borated water volume of 101,120 gallons,

([Later]% of span),

() 2) A minimum b'oron concentration of 2000 ppm and A minimum solution temperature of 40*F.

3)

APPLICABILITY: HOCES 5 and 6.

ACTION:

With no borated water source OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes,

. SURVEILLANCE REQUIREMENTS 4.1.2.5 The above required borated water source shall be demonstrated OPERABLE:

a. At least once per 7 days by:

1) Verifying the boron concentration of the water,

2) Verifying the contained borated water volume, and

3) Verifying the boric acid storage tank solution temperature when it is the source of borated water.

b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature when it is the source of borated water and the outside air temperature is O less than 40*F.

COMANCHE PEAK - UNIT 1 3/4 1-11 .

IXX-88512 ATTACHMEWI 4 PAGE 15 0F 29 REACTIVITY CONTROL SYSTEMS BORATED WATER SDURCES - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.6 As a minimum, the following borated water source (s) shall La OPERABLE as required by Specification 3.1.2.2:

a. A boric acid storage tank with:

1) A minimum contained borated water volume of [22,870] gallons,

([Later]% of span),

2) A minimum boron concentration of 7000 ppm, and

3) A minimum solution temperature of 65*F.

b. The refueling water storage tank (RWST) with:

1) A minimum contained borated water volume of [479,900] gallons,

([Later]% of span),

2) A boron concentration between 2000 ppm and [2200] ppm,

) 3) A minimum solution temperature of 40*F, and

4) A maximum solution temperature of 120 F.

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

ACTION:

a. With the boric acid storage tank inoperable and being used as one of the above required borated water sources, restore the tank 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 borated to a SHUTOOWN MARGIN equivalent to at least 1% ak/k at 200*F; restore the ooric acid storage tank 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 .

b. With the RWST inoperable, restore the tank to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .'nd in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

O COMANCHE PEAK - UNIT 1 3/4 1-12 ,

1 TXX-88512 ATTACHMENT 4 PAGE 16 0F 29

~ ~

REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS 4.1.2.6 Each borated water source shall be demonstrated OPERABLE:

a. At least once per 7 days by:

1) Ver'fying the boron concentration in the water,

2) Verifying the contained borated water volume of the water source, and

3) Verifying the boric acid storage tank solution temperature when it is the source of borated water.

b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature when the outside air temperature is either less than 40 F or greater than 120*F.

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O COMANCHE PEAK - UNIT 1 3/4 1-13 .

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. TXX-88512 AllACHM(Hi4 PAGE 17 0F 29

_ . 1 REACTIVITY CONTROL SYSTEMS

,.)

3/4.1.3 MOVABLE CONTROL ASSEMBLIES i

GROUP HEIGHT LIMITING CONDITION FOR OPERATION 3.1.3.1 All shutdown and control rods shall be OPERABLE and positioned within i i 12 steps (indicated position) of their group step counter demand position.

APPLICABILITY: MODES 1* and 2*. ,

ACTION:

a. With one or more rods inoperable due to being immovable as a result of excessive friction or mechanical interference or known to bs untrippable, determine that the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and be in HOT ,

STANOBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

10 I: 0037

b. W!th cr: th:n :n: fu? ?-?:ngth r:d !c:p:r:b?: : -!::?!gn d 're.-

the gr0up step-counteeHf: :nd po:iti:n by ::r: then i 12 :t:p:

(indicated petitien), be 4. ug7 37agggy $ ggt 7 g 3 ;7;,

b. -er- With one full-length rod trippable but inoperable due to causes other than addressed by ACTION a., above, or misaligned from its group step counter demand height by more than i 12 steps (indicated position), POWER OPERATION may continue provided that O' within 1 hour:

1. The rod is restored to OPERABLE status within the above alignment requirements, or

2. The rod is declared inoperable and the remainder of the rods in the group with the inoperable rod are aligned to within i 12 steps of the inoperable roo while maintaining the rod sequence and insertion limits of Figure 3.1-1. The THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation, or

3. The rod is declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied. POWER OPERATION may then continue provided that:

a) A reevaluation of each accident analysis of Table 3.1-1 is performed within 5 days; this reevaluation shall confirm that the previously analyzed results of these accidents remain valid for the duration of operation under these conditions; b) The SHUT 00WN MARGIN requirement of Specification 3.1.1.1 is oetermined at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ;

See Special Test Exceptions Specifications 3.10.2 and 3.10.3.

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

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TXX-88512 l

' l ATTACHMENT 4 PAGE 18 0F 29 REACTIVITY CONTROL SYSTEMS M)

LIMITING CONDITION FOR OPERATION ACTION (Continued) c) A power distribution map is obtained from the movable incoredetectorsandF(Z)andFhareverifiedtobe 9

within their limits within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; and d) The THERMAL POWER level is reduced to less than or equal to 75% of RATED THERMAL POWER within the next hour and within the following 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months the Figh Neutron Flux Trip Setpoint is reduced to less than or equal to 85%

of RATED THERMAL POWER.

ND

10 h 0038 i

O SURVEILLANCE REQUIREMENTS l

! 4.1.3.1.1 The position of each rod shall be determined to t'e within the group demand limit by verifying the individual rod positions at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the rod position deviation monitor is inoperable, then verify the group positions at least once pt.r 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

4.1.3.1.2 Each rod not fully inserted in the core shall be determined to be OPERABLE by movement of at least 10 steps in any one direction at least once per 31 days.

[

. c. With more than one rod trippable but inoperable due to causes other than l addressed by ACTION a above, POWER OPERATION may continue provided that:

1. Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , the remainder of the rods in the bank (s) with the

inoperable rods are aligned to within i 12 steps of the inoper-able rods while maintaining the rod sequence and insertion limits l of Figure 3.1-1. The THERMAL POWER level shall be. restricted pursuant to Specification 3.1.3.6 during subsequent operation, and

2. The inoperable rods are restored to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

d. With more than one rod misaligned from its group step counter demand

' height by more than i 12 steps (indicated position), be in H0T STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

COMANCHE PEAK - UNIT 1 3/4 1-15 ,

TXX-88512 AliACHMENT 4 PAGE 19 0F 29 TABLE 3.1-1 l ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT OF AN INOPERABLE R00

Rod Cluster Control Assembly Insertion Characteristics

Rod Cluster Control Assembly Misalignment Decrease in Reactor Coolant Inventory l Inadvertent opening of a pressurizer safety or relief valve Break in instrument lire or other lines from reactor coolant pressure boundary that penetrate containment Steam generator tube rupture Loss of coolant accidents resulting from a speccrum of postulated piping breaks within the reactor coolant pressure boundary Increases in Heat Removal by the Secondary System (steam system piping failure)

Spectrum of Rod Cluster Control Assembly Ejection Accidents O

COMANCHE PEAK - UNIT 1 3/4 1-16

IXX 88512 AliACHMENT 4 '

PAGE 20 0F 29 REACTIVITY CONTROL SYSTEMS 0

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m e

POSITION INDICATION SYSTEMS - OPERATING LIMITING CONDITION FOR OPERATION l

3.1.3.2 The Digital Rod Position Indication System and the Oemand Position Indication System shall be OPERABLE and capable of determining the control rod positions within 2 12 steps.

APPLICABILITY: MODES 1 and 2.

ACTION:

a. With a maximum of one digital rod position indicator per bank inoperable either:

1. Determine the position of the nonindicating rod (s) indirectly by the movable incore detectors at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and immediately after any motion.of the nonindicating rod which exceeds 24 steps in one direction since the last determination of the rod's position, or

2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

I b. With a maximum of one demand position indicator per bank inoperable either:

1. Verify that all digital rod position indicators for the affected bank are OPERABLE and that the most withdrawn rod and the least withdrawn rod of the bank are within a maximum of 12 steps of each other at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , or

2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

SURVEILLANCE REQUIREMENTS 4.1.3.2 Each digital rod position indicator shall be determined to be OPERABLE by verifying that the Oemand Position Indication System and the Digital Rod Position Indication System agree within 12 steps at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the rod position deviation monitor is inoperable, then compare the Demand Position Indication System and the

. Digital Rod Position Indication System at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

0 V

COMANCHE PEAK - UNIT 1 3/4 1-17

IXX-88512 ATTACHMENT 4 .

PAGE 21 Of 29 REACTIVITY"CONTROL SYSTEMS

) POSITION INDICATION SYSTEM - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.3.3 One digital rod position indicator (excluain2 demand position indication) shall be OPERABLE and capable of determining the control rod position within i 12 steps for each shutdown or control rod not fully inserted.

APPLICABILITY: MODES 3a

4" *, and 5* M to :: 0041 1

ACTION:

With less than the above required position indicator (s) OPERABLE, immediately open the Reactor Trip System breakers.

SURVEILLANCE REQUIREMENTS I)(

'w 4.1.3.3 Each of the above required digital rod position indicator (s) shall be l determined to be OPERABLE by verifying that the digital rod position indicators agree with the demand position indicators within 12 steps when exercised over the full-range of rod travel at least once per 18 months.

l l

l This requirement is not applicable while verifying Digital Rod Posi ion Indication System OPERABILITY following. maintenance provided:

maintained less than or equal to~0.99, and (2)or only control one shutd 1 rod bank is withdrawn ..

from the fully inse~rted position at one time.

.~ -

m

WiththeReactorTripSystembreakersintheclosedpositionj

-**See Special Test Exceptions Specification 3.10.5; and g) > T 9tG.R.T ID I: 0042 (d

COMANCHE PEAK - UNIT 1 3/4 1-18

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IXX-88512 1 ATIACHMENT4 PAGE 22 Of 29 REACTIVITY CONTROL SYSTEMS

(

R00 OROP TIME LIMITING CONDITION FOR OPERATION 3.1.3.4 The individual (shutdown and control) rod drop time from the fully withdrawn position shall be less than or equal to 2.4 seconds from beginning of decay of stationary gripper coil voltage to dashpot entry with:

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

b. All reactor coolant pumps operating.

APPLICABILITY: MODES 1 and 2.

ACTION:

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

SURVEILLANCE REQUIREMENTS 4.1.3.4 The rod drop time of rods shall be demonstrated through measurement prior to reactor criticality:

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

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

c. At least once per 18 months.

O COMANCHE PEAK - UNIT 1 3/4 1-19 ,

It!-88512 AlfACHMENT4 PAGE 23 0F 29 REACTIVITY CONTROL SYSTEMS

~

SHUT 00WN ROD INSEJYION LIMIT l

LIMITINGCONDI0NFOROPERATION 3.1. 3. 5 All shutdown rods shall be fully withdrawn. APPLICA61LITY: H0 DES 1* and 2* "*. ACTION: With a maximum of one shutdown rod not fully withdrawn, except for surveillance testing pursuant to Specification 4.1.3.1.2, within 1 hour either:

a. Fully withdraw the rod, or

b. Declare the rod to be inoperable and apply Specification 3.1. 3.1.

U ( SURVEILLANCE REQUIREMENTS 4.1.3.5 Each shutdown rod shall be determined to be fully withdrawn: , a. Within 15 minutes prior to withdrawal of any rods in Control 1 Bank A, 8, C, or 0 Juring an approach to reactor criticality, and j b. At least once per 12 hours thereafter. 1

        *See Special Test Exceptions Specifications 3.10.2 and 3.10.3.
       **With K,ff greater than or equal to 1.

COMANCHE PEAK - UNIT 1 3/4 1-20 n

l TXX-88512 4 . ATTACHMENT 4 i PAGE 24 0F 29 p REACTIVITY CONTROL SYSTEMS o CONTROL ROD INSERTION LIMITS l LIMITING CONDITION FOR OPERATION 3.1.3.6 The control banks shall be limited in physical insertion as shown in Figure 3.1-1. APPLICABILITY: MODES 1* and 2* **. ACTION: With the control banks inserted beyond the above insertion limits, except for surveillance testing pursuant to Specification 4.1.3.1.2:

a. Restore the control banks to within the limits within 2 hours, or

b. Reduce THERMAL POWER within 2 hours to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the bank pcsi-tion using the above figure, or

c. Be in at least HOT STANDBY within 6 hours.

SURVEILLANCE REQUIREMENTS , 4.1.3.6 The position of each control bank shall be determined to be within the insertion limits at least once per 12 hours except during time intervals when the rod insertion limit monitor is inoperabl'e, then verify the individual rod positions at least once per 4 hours.

          'See Special Test Exceptions Specifications 3.10.2 and 3.10.3.
        **With K,ff greater than or equal to 1.

COMANCHE PEAK - UNIT 1 3/4 1-21 e

l TXX-88512 ATTACHHENT 4 PAGE 25 0F 29~ o ; DR\F 9 1 O l l l l I l FIGURE 3.1-1 ROD BANK INSERTION LIMITS VERSUS THERMAL POWER i COMANCHE PEAK - UNIT 1 3/4 1-22

i

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IXX-88512 ATTACHMENT 4 FAli26CF29

 .            3/4.1 REACTIVITY CONTROL SYSTEMS                                                         s I. -
         . BASES                  _

3/4.1.1 BORATION COVfROL 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN A sufficient SOUT00WN MARGIN ensures that: (1) the reactor can be made subcritical from all operating conditions, (2) the reactivity transients asso-ciated with postulated accident conditions are controllable within acceptable limits, and (3) the re6cu r will be maintained sufficiently suberitical to preclude inadvertent criticality in the shutdown condition. SHUTOOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS horon concentration, and RCS T,yg. The most restrictive condition occurs at EOL, with T gyg at no loao operating temperature, and is associated with a postulated steam line break accident and resulting uncon trolled RCS cooldown. T.n the analysis of this accident, a minimum SHUTOOWN MARGIN of 1.6% ak/k is required to control the reactivity transient. Accordingly, the SHUTOOWN MARGIN requirement is based upon this limiting condition and is consistent with FSAR safety analysis assumptions. With T,.,g p less than 200'F, the reactivity transients resulting from a postulated steam ( line break cooldown are minimal and a 1% ak/k SHUTDOWN MARGIN provides

        ;    adequa<,e protection.

3/4.1.1.3 M00ERATOR TEMPERATURE COEFFICIENT The limitations o.1 moderator temperature coefficient (MTC) are provided to ensure that the value of this coefficient remains within the limiting condition assumed in the FSAR accident t.nd transient analyses. The MTC values of this specification are applicable to a specific set of plant conditions; accordingly, verification of MTC values at conditions other than those explicitly stated will require extrapolacion to those conditions in order to permit an accurate comparison. The most negative M1C, value equivalent to the most positive moderator density coefficient (MOC), was obtained by increments' S r:orrecting the MDC used in ths FSAR analyses to n W orl opera 'n3 condit..as. These ccrrection. P v ' COMANCHE PEAX - UNIT 1 & 3/4 1-1 ,

TXX-88512

ATIACHMENT4 PAGE 27 0F 29 [] V REACTIVITY CONTROL SYSTEMS U V... - BASES MODERATOR TEMPERATURE COEFFICIENT (Continued) involved subtracting the incremental change in the HOC associated with e core conlition of all rods inserted (most positive MDC) to an all rods withdrawn condition and, a conversion for the rate of change of moderator density with temperature at RATED THERMAL POWER conditions. This value of the MDC was then transformed into the limiting MTC value -4.0 x 10 4 ak/k/*F. The MTC value of -3.1 x 10 4 ak/k/ F represents a conservative value (with corrections for burnup and soluble baron) at a core condition of 300 ppm equilibrium t,oron concentration and is obtained by making these corrc::tions to the limiting MTC value of -4.0 x 10 4 ak/k/ F. The Surveillance Requirements for measurement of the MTC at the beginning and near the end of the fuel cycle are adequate to confirm that the MTC remains within its limits since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup. 3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY This specificativa ensures that the reactor will not be made critical V with the Reactor Coolant System average temperature less than 551'F. This limitation is required to ensure: (1) the moderator temperature coefficient is within it analyzed temperature range, (2) the trip instrumentation is within its normal operating range, (3) the pressurizqr is capable of being in an OPERABLE status with a steam bubble, and (4) the reactor vessel is above its minimum RT temperature. NDT 3/4.1.2 BORATION SYSTEMS The Boron Injection System ensures that negative reactivity control is available deing each mode of facility operation. The components required to perform th?s function include: (1) borated water sources, (2) charging pumps, (3) separate flow paths, (4) boric acid transfer pumps, and (5) an emergency power st.pply from OPERABLE diesel generators. With the R;S cverage temper-ature above 200*F, a minimum of two boron injection flow paths are required to ensure single functional capability ir. the event an assumed failure renders one of the flow paths inoperable. The boration capability of either flow path is sufficient to provide a SHUT 00WN MARGIN from expected operating conditions of 1.6T, ak/k after ytnan dec.ay and a.coldown to 200'F. The maximum expected boration capability requirement occurs at EOL from full power equilibrium xenon conditions and requires [22.870] gallons of 7000 ppm borated water from the boric acid storage tanks or f.475;900] gallons of 2000 ppm borated water from the refueling water storage O- tink (RWST). 'd COMANCHE PEAK - UNIT i B 3/4 1-2 ,

                                                       .                                               1
                                                                                ,                      l 4            TH 88512                                                                                   '

Ai!ACHMENT4 PAGE 28 CF 29 j 7 REACTIVITY CONTROL SYSTEMS

          - BASES i

. 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 i 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. 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 [6,385) gallons of 7000 ppe borated water from the boric acid storage tanks or (101,120] gallons of 2000 ppm borated water from the RWST.

The contained water volume limits include allowance for water not available because of discharge line location and other physical characteristics. The limits on contained water volume and boron concentration of the RWST are also consistent with Specification 3.5.4. The OPERAli!LITY of one Boron Injection System during REFUELING ens.ures that this system is available for reactivity control while in MODE 6. 3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that: (1) acceptable power distri-bution limits are maintained, (2) the minimum SHUTOOWN MARGIN is maintained, and (3) the. potential effects of rod misalignment on associated accident analyses are limited. OPERABILITY of the control rod positicn indicators is required to determine control rod positions and thereby ensure compliance with the control rod aligament and insertion limits. Verification that the Digital Rod Position Indicator agrees with the demanded position within i 12 steps at 24, 48, 120, and 228 steps withdrawn for the Control Banks and 18, :'10, and 223 steps with-drawn for the Shutdown Banks provides assurances that the Digital Rod Position Indicator is operating correctly over the full range of indication. Since the Digital Rod Position Indication System does not indicate the actual shutdown rod position between 18 steps and 210 steps, only pair- 'n the indicated ranges are picked for verification of agreement with deme.. ed position. COMANCHE PEAK - UNIT 1 8 3/4 1-3 '

TXX-88512 AliACHMEN1 4 PAGE 29 0F 29 REACTIVITY CONTROL SYSTEMS BASES MOVABLE CONTROL ASSEMBLIES (Continued) The ACTION statements which permit limited variations from the basic requirements are accompanied by additional restrictions which ensure that the original design cr:teria are met. Misalignment of a rod requires measurement l of peaking factors and a restriction in THERMAL POWER. These restrictions pro- ! vide assurance of fuel rod integrity during continued operation. In addition, those safety analyses affected by a misaligned rod are reevaluated to confirm that the results remain valid during future operation. 4 g,QS qR,T 10 8: 0454 The maximum rod drop time restrictidn is consistent with the assumed rod drop time usec in the safety analyses. Measurement with T,yg greater than or equal to 551*F and with all reactor coolant pumps operating ensures that the measured drop times will be representative of insertion times experienced during a Reactor trip at operating conditions. Control rod positions and OPERABILITY of the rod position indicators are required to be verified on a nominal basis of once per 12 hours with more fre-t quent verifications required if an automatic monitoring channel is inoperable. These verification frequencies are adequate for assuring that the applicable l 6COs are s.tisfied. g (']. v- ( 1 l For Specification 3.1.3.1 ACTIONS b and e it is incumbent upon the )lant to verify the trippa'oility of the inoperable control rod (s). This may >e by 1 verification of a control system failure, usually electrical in nature, or i

          *             ,that the failure is associated with the control rod stepping mechanism.          In the event the plant is unable to verify the rod (s) trippability, it must be assumed to be untrippable and thus fall under the requirements of action a.

Assuming a controlled shutdown from 100% RATED THERMAL POWER, this allows approximately four hours for this verification. O COMANCHE PEAX - UNIT 1 B 3/4 1-4 , j

TIX-88512 ATTACHMENT 5 Pt.GE 1 0F 25 O COMANCHE PEAX STEAM ELECTRIC STATION TECHNICAL SPECIFICATION 3/4.2 O V O .

   ,n             - .-... --- , ,.. ,, ,,.,--        - - , , , , , ,      - . . -_ , , - , - - -   -, --..,,. , ,

                ..             ..             =                                                           .

i IXX 88512 CPSES Technical Specifications NRC Oraft 2 Markuo Section 3/4.2 Change 10# Justification For Change 0047 The LCO has been written in terms of Ffg to simplify the expression and to eliminate R. The actual value of the constant in the F dependsonwhethermeasurementuncertaina[Hexpression es are included or not. The Standard Technical Specification value does include uncertainties. Handling in this manner E ever, presents an inconsistency with the BASES for Specificatior 2.1.1-(uses the 1.55 analysis value) and FSAR pg. 4.3-27 (1.55 value). To accomplish the same objective, the LC0 has been written using the analysis value of 1.55, and a requirement has been added to Surveillance Requirement 4.2.3.2 to increase the measured F by 4% to bound worst case measurement uncertainties.$H This chat,ge is similar to that '.icensed at , Farley, Byron and Vogtle. The above change eliminates the need for the LC0 to address the 4% incore measurement uncertainty. In addition, reference to the 0.1% error for feedwater venturi fouling has been eliminated. This error is insignificant for CPSES as presented in the Imoroved Thermal Design Report (ITDR). The staff has previously concurred (SER Supplement 12, Section 4.4) that no error hg. need be assumed for venturi fouling. 0049 Figure 3.2-3 is deleted since it provides no meaningful information and complicates the implementation of the specification. The limits of flow and R for CPSES, as well as other recent vintage Westinghouse plants, are constants: Flow not less than 389,700 gpm and R not greater than one (1). On earlier Westinghouse plants the R limits were not constant, but rather varied with core burnup. This dictated the need for'a figura to show either the Burnup (Rod Bow Penalty) or the resulting R values. Since the CPSES R limit is unity (1), the expression for the Hot Channel Factor can be simplified and stated directly as F[H < 1.55 [1.0 + 0.2 (1.0-P)] This change is similar to that Licensed at Seabrook, Vogtle, South Texas, Byron and Millstone. 0050 All requirements dealing with RCS Flow have been moved to the DNB Specification. This change is made A sinceRCSFlowRatesandFQHarebothfactorsthat deal with core thermal analysis and the a)proach to O Departure From Hucleate Boiling (DNB.S. T1e limits for each one, however, have na direct bearing on one another. Factors involved with DNB analysis can be more conveniently separated into 2 main categories: l

TXF"$12 AliA m i 5 CPSES Technical Specifications PAGE 3 of 25 NRC Draft 2 Nrkup Section 3/4.2 Chance 10# Justification For Chance 0050 (cont.) (1) Fuel Rod parameters and (2) Coolant / Moderator (flow channel) parameters. The later category consists of Coolant Temperature, Coolant Pressure and Coolant Flow. While these parameters can vary from one coolant channel to another, the macroscopic parameters of RCS T(avg), Pressurizer Pressure and RCS total flow rate provide operator indication of the coolant channel conditions and all are directly under operator control. The same is not true of power related hot channel and peaking factors, including F H. These factors cannot be readily determined on a real-time basis, and cannot be as readily changed or corrected if deviations from required limits have occurred. While reducing reactor power reduces the affect of these parameters on approach to DNB, it does not alter the factors, since they are ratios. For these reasons, it is more logical to address RCS flow with the other operator controlled DNB parameters in Specification 3.2.5. The associated Action Statement is more conservative and the allowed outage time is consistent with that of a primary operator controlled parametei. This change will simplify and make it much easier to determine the proper and conservative corrective measures specified by the Action Statements in the event of an out of limit flow condition. This change is similar to that

                       !.icensed at Seabrook, Vogtle and South Texas.

The change from 31 EFPD to 31 day frequency is to coincide with the frequencies of Specification 4.2.5, which is also consistent with non-core burnup related parameters. Addition of the requirement to conauct this verification of RCS flow by plant computer indication or by voltage measurement is consistent with the assumptions of the CPSES Improved Thermal Design Report (ITOR) (See SER Supplement 12, Section 4.4) te obtain the re gired accuracy. SeparationoftheRCSFlcsrequirementsandtheF[ Hand the proposed deletion of Figure 3.2-3 and R values eliminate any need for or reference to these in the Surveillance (see Separate write-ups). Requirement for calibration of measurement instrumentation prior to the performance of the calorimetric flow measurement has been moved to the requirement for flow rate determination by precision heat balance (Specification 4.2.3.5) which is the calorimetric. It is more appropriata to include the calibration requirements along with the most closely associated Surveillance which T is the heat balance or calorimetric. J

! IXX-14$12 l AllACHENT5 CPSES Technical Specifications i PA K 4 of 25 NRC Oraft 2 Markup Section 3/4.2 q Chanae ID# Lustification For Chance 0050 (cont.) The requirement to normalize the RCS Flow Channels (Elbow Tap D/P instruments) has been added to ensure that the specified calibration includes this normalization following RCS flow rate determination by precision heat balance. This is required by the ITDR t: obtain the stated 1.8% accuracy and is addressed by Supplement 12 of the SER, Section 4.4 as accepted b; the NRC. The time period for this calibration has been revised to 90 days prior to performance of the calorimetric which is consistent with the instrument drift (uncertainties) assumed in the CPSES ITDR. This report used 90 days to allow more time for possible off-site calibration of the required instruments. This report has been reviewad and approved by the NRC in SER Supplement 12, Section 4.4 for the determination of the 1.8% total uncertainty. The specific instruments of concern have been specified, consistent with the ITDR to avoid future mis , interpretations. 0051 This change from the Standard Technical Specifications is made to simplify the Quadrant Power Tilt Ratio (QPTR) Action Statements. The new Action Statements better A address the true concern for quadrant power tilts which is U to ensure that core peaking factors remain within their limits between flux maps. This change meets the purpose of this Specification by detecting gross changes in core power distribution between monthly incore flux maps. This change is similar to that Licensed at Seabrook and South Texas. 0052 This change is made since the Technical Specification 3.3.3.2 is being relocated to the CPSES Technical Specification Improvement Program.

TU-88512

   .          ATTACW.ENT5 PAGE 5 0F 25
                            ^       '

3/4.2 POWER DISTRIBUTION LIMITS , r i 3/4.2.1 AXIAL FLUX DIFFERENCE LIMITING CONDITION FOR OPERATION

3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within the following target band (flux difference units) about the target flux difference:

a. 1 5% for core average accumulated burnup of less than or equal to 3000 MWD /MTU; and

b. + 3%, -12% for core average accumulated buinup of greater than 3000 HlD/MTV.

The indicated AFD may deviate outside the above required target band at greater than or equal to 50% but less than 90% of RATED THERMAL POWER provided the indi-cated AFD is within the Acceptable Operation Limits of Figure 3.2-1 and the cumu-lative penal *.y deviation time does not exceed I hour during the previous 24 hours. The indicated AFD may deviate outside the above required target band at greater than 15% but less than 50% of RATED THERMAL POWER provided the cumulative penalty deviation time does not exceed I hour during the previous 24 hours. APPLICABILITY: MODE 1, above 15% of RATED THERMAL POWER.* ACTION:

a. With the indicated AFD outside of the above required target band and

( with THERMAL POWER greater than or equal to 90% of RATED THERMAL POWER, within 15 minutes either:

1. Restore the indicated AFD to within the target band limits, or

2. Reduce THERMAL POWER to less than 90% of RATED THERMAL POWER.

b. With the indicated AFD outside of the above required target band for more t5an 1 hour of cumulative penalty deviation time during the previous 24 hours or outside the Acceptable Operation Limits of Figure 3.2-1 and with THERMAL POWER less than 90% but equal to or greater than 50% of RATED THERMAL POWER:

4

1. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes, and

2. Reduce the Power Range Neutron Flux - High Trip ** Setpoints to less than or equal to 55% of RATED THERMAL POWER within the rext 4 hours.

            *See Special Test Excentions Specification 3.10.2.

Surveillance testing of the Power Range Neutron Flux Channels may be performed pursuant to Specification 4.3.1.1 provided the indicated AFD is mairtained within the Acceptable Operation Limits of Figure 3.2-1. A total of 16 hours operation may be accumulated with the AFD outside of the above required target band d'jring D testing without penalty deviation. COMANCHE PEAK - UNIT 1 3/4 2-1

IIX-88512 AliACKINT 5 PAGE 6 0F 25 POWER DISTRIBUTION LIMITS O LIMITING CONDITION FOR CPERATION ACTION (Continued)

c. With the indicated AFD outside of the above required target band for more than 1 hour of cumulative penalty deviation time during the previous 24 hours and with THERMAL POWER less than 50% but greater than 15% of RATED THERMAL POWER, the THERMAL POWER shall not be increased equal to or greater than 50% of RATED THERMAL POWER until the indicated AFD is within the above required target band.

SURVEILLANCE REQUIREMENTS 4.2.1.1 The indicated AFD shall be determined to be within its limits during POWER OPERATION above 15% of RATED THERMAL POWER by:

a. Monitoring the indicated AFD for each OPERABLE excore channel:

1) At least once per 7 days when the AFD Monitor Alarm is OPERABLE, and

2) At least once per hour for the first 24 hours after restoring the AFD Monitor Alarm to OPERABLE status.

b. Monitoring and logging the indicated AFD for each OPERABLE excore

, Qt channel at least once per hour for the first 24 hours and at least l once per 30 minutes thereaf ter, when the AFD Monitor Alarm is

inoperable. The logged values of the indicated AFO shall be assumed to exist during the interval preceding each logging.

4.2.1.2 The indicated AFD shall be considered outside of its target band when two or more OPERABLE excore channels are indicating the AFD to be outside the target band. Pen'.lty deviation outsid9 of the above rJquired target band shall be accumulated on a time basis of:

a. One minute penalty deviation for each 1 minute of POWER OPERATION outside of the target 'oand at THERMAL POWER, levels equal to or above 50% of RATED THERMAL POWER, and

b. One-half minute penalty deviation for each 1 minute of POWER OPERATION outside of the target band ar. THERMAL POWER levels between 15% and 50% of RATED THERMAL POWER.

4.2.1.' The target flux difference of each OPERABLE excore channel shall be i determined by measurement at least once per 92 Effective Full Power Days. l The provisions of Specification 4.0.4 are not applicable. 4.2.1.4 The target flux difference shall be updated at least once per 31 Effective Full Power Days by either determining the target flux difference l pursuant to Specification 4.2.1.3 above or by linear interpolation between the most recently measured value and 0% at the end of the cycle life. The provi-l~ sions of Specification 4.0.4 are not applicable. COMANCHE PEAK - UNIT 1 3/4 2-2

p ' ;- . - j; IIX-88512. 4' AllACHMENT5 ,

;-                   PAGE 7 0F 25.
                                   ~ '

!: e , 4 s F e i v l. h t 4 . ( l I

'r j t ?

I; .

t i

j' i F f i l !

i. !

l t, i i. l l T

                                                                                       .                     i i

i L ( y

FIGURE 3.2-1 4 AXIAL FLUX DIFFERI.NCE LIMITS AS A FUNCTION OF i l RATED THERMAL POWER i f COMANCHE PEAK - UNIT 1 3/4 2-3 f t f

                                                                                         .__ ___________.a

IXX-88!i? ATTACHMENT 5 FAGE 8 0F 25 POWER OISTRIBUTION LIMITS { 3/4.2.2 HEATFLUXHOTCHANNELFACTOR-Fg LIMITING CONDITION FOR OPERATION 3.2.2 F (Z) q shall be limited by the following relationships: eO (Z) 5 [2.32) [K(Z)] for P > 0.5 P Fq (Z) $ [(4.64)] [K(Z)] for P $ 0.5

                                                              , and Where:     P _ THERMAL POWER RATED THERMAL POWER K(Z) = the function obtained from Figure 3.2-2 for a given core height location.

APPLICABILITY: MODE 1. ACTION: With F (2) exceeding its limit: 9 p a. Reduce THERMAL POWER at least M for each 3 F q (Z) exceeds the O' limit within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours; POWER OPERATION may proceed for up to a total of 72 hours; subsequent POWER OPERATION may proceed provided tne Overpower N-16 Trip Setpoints have been reduced at least a for each 3 Fg(Z) exceeds the limit; and

b. Identify and correct the cause of the out-of-limit condit'lon prior to increasing THERMAL POWER above the reduced limit re-quired by ACTION a., above; THERMAL POWER may then be increased provided F (Z) is demonstrated through incore mapping to be 9

within its limit. f O COMANCHE PEAK - UNIT 1 3/4 2-4

I TXX-88512 AliACHMENT $ FAGE 9 Y 25 l O , l O. FIGURE 1.2-2 K(Z) - NORMALIZED F g(Z) AS A FUNCTION OF CORE HEIGHT COMANCHE PEAK - UNIT 1 3/4 2-5

IXX 88512 ATTAC M WT 5 PAGE 10 0F 25

POWERDISTRIBUTiONLIMITS ynniI V I SURVEILLANCE REQUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable. 4.2.2.2 F xy shall be evaluated to determine if qF (Z) is within its limit by:

a. Using the movable incore detectors to obtain a power distribution map at any THERMAL POWER greater than 5% of RATED THERMAL POWER,

b. Increasing the measured F xy component of the power distribution map by 3% to account for manufacturing tolerances and further increasing the value by 5% to account for measurement uncertainties, C

c. Comparing the F xy computed (Fxj) obtained in Specification 4.2.2.2b.,

above to: ' The F

1) xy limits for RATED THERMAL POWER x(FRTP) for the appropriate measured core planes given in Specification 4.2.2.2e and f.,

below, and ,

2) The relationship:

e =FRTP [1+0.2(1-P)], x l Where F is the limit for fractional THERMAL POWER operation express as a function of F xRTP and P is the fraction of RATED ; THERMAL POWER at which F was measured. xy according to the following schedule:

d. Remeasuring F

1) When F x

C is greater than the F xRTP limit for th? appropriate measured core pla n but less than the F relationship, additional power distribution maps shall be taken dF x comparad to F xRTP and F xy either: a) Within 24 hours after exceeding by 20% of RATED THERMAL C POWER or greater, the THERMAL POWER at which F*Y ,,, ),,g determined, or b) At least once per 31 Effective Full Power Days (EFPD), whichever occurs first, l O l COMANCHE PEAK - UNIT 1 3/4 2 6 l l l

IXX-88512 AliACHMENT5 y PAGE 11 Of 25 POWER DISTRIBUTION LIMITS

     ,  SURVEILLANCE REQUIREMENTS (Continued)

2) When the F, i s less than or equal to the F,RTP limit for the appropriate measured core plane, additional power distribution maps shall be taken and F, compared to F,R and F at least once per 31 EFPD.

R

e. The F ) shall be provided for xy limits for RATED THERMAL POWER (F x all core planes containing Bank "0" control rods and all unrodded core planes in a Radial Peaking Factor Limit Report per Specifica-tion 6.9.1.6;

f. The F xy limits of Specification 4.2.2.2e. , above, are not applicable in the fn11owing core planes regions as measured in percent of core height from the bottom of the fuel:

1) Lower core region from 0 to 15%, inclusive,

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

3) Grid plane regicac at 17.8 1 2%, 32.1 1 2%, 46.4 1 2%, 60.6 2 2%,

and 74.9 1 2%, inclusive, and V' 4) Core plane regions within i 2% of core height [12.88 inches) about the bank demand position of the Bank "0" control rods.

g. With F x exceeding F, , the effects of F xy on F9 (2) shall be evaluated to determine if qF (Z) is within its limits.

4.2.2.3 When F9 (2) is measured for other than F determinations, an overall measured qF (Z) shall be obtained from a power distribution mao and increased by 3% to account for manufacturing tolerances and further increased by 5% to account for measurement uncer!.ainty. l O COMANCHE PEAK - UNIT 1 3/4 2-7 ? . i

            , -                ,my-    -   -       - - -   -,,nr-,.  - - - -,

IIX88512 ATTACHMENT 5 PAGE 12 0F 25 POWER DISTRIBUT' ION LIMITS -

3/4.2.3 #

a. SCO "LO'd RATC ANO NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR - F u LIMITING CONDITION FOR OPERATION t:" sw u we.1 iteJ by N wha. g M M e .p '

3.2.3 7j:g :; . tin t4ewf- 5*dicated-Reactoe-Coe16nt-System-(RCS)-total-flow-rate--and-R-shelbb+-maintained-withie-the-regi:r Of :11:e:54e-operation-shown-en44 9uee-3r2-3-fee-foue-100; operatioi>. c - Where:

                                                         ,. N
                                                                      '-~~ w p# 3u l, g .3- [/.0ec.A.(1.c4)]
                    .           a   _
                                    ~
                                                         'AH                 ,
                                          -h49-[IT 0-+-0 r 2-(-1. 0 - Ah}--

p , THERMAL POWER g -end. RATED THERMAL POWER

. F g - P,e;;rd relues of . g cMaM h u;bg Oc m'aMacee 4etectees-to-eMeir : p:u:r di+4Mht4en-capr-The-wasured.

velues-ef4Ng shall 0: ;. sed-to-colcula4+-A-64*ce449ure-3r2-1 includ:: pen:ltie: 'ee-unde tec ted-feedwa t e r-v en t u r t- fou l4 ng-o f-Gr-l% :nd for :::sterement-uncerta4nties-of-Ir84-for-f4ew--

nd 4% fe-4ecore-measurement-of-F-h. n h 000

( ! APPLICABILITY: . TOE 1. ACTION: (J.tk F ucnL) dsldd! W+th-the :;;ti[,atica n of "CS total flew rate :nd

Out:ide the regien of--

          -accept:ble Operation :h;wn n Figure 3.2-3t-

a. Within2hoursejther: -

1. Restore th$ : rtin:ti:r f RCS tet:! '!:r rat: :nd " to within theabovelimith,or

2. Reduce THERMAL POWEi' to less than 50% of RATtD THERMAL POWER and reduce the Power Range Neutron Flux - High Trip Setpoint to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours.

k 3 (a COMANCHE PEAK - UNIT 1 3/4 2-8

TXX88512 AlfACHMENT5 PAGE 13 0F 25 " l 10 l 00j9 FIGURE 3,2-3 r RCS TOTAL FLOW RATE VERSUS R - FOUR LOOPS IN OPERATION

O COMANCHE PEAK - UNIT 1 3/4 2-9

               .-  - . - ,-              -    --.,n.,        . - - - - - - -   - , - - - - , .
                                                                                               ,a. , -~ ~ ,

     .        IXX-88512 A11ACHIEWI 5

PAGE 14 0F 25 l g L POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION ACTION (Continued)

b. Within 24 hours uf initially being outside the above limits, verify through incere flux mapping :nd RCS t t:1 -flow rett-compaf4sen that 4

b ** ee-eem44*et4ee-ef-A-and-RGS-total-flow-rate-are restored to within the above limits, or reduce THERMAL POWER to lest than 5% of RA E0 THERMAL POWER within the next 2 hours p d

c. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced THERMAL POWER limit required by ACTION a.2 and/or b., above; subsequent POWER OPERATION may proceed provided that the-<0-binatisi Of 9 :nd indfeated-AGS-p"A

is -= tet:1 '1 e rate : demonstrated, through incore flux mappin 905 tet:1 'lew rate-comparison, to be within the regien e' :g:::pt:ble-and

           'h i - * :p:r:tien :heur en Figure 3r24 prior to exceeding the following THERMAL POWER levels:

Its 0047

1. A nominal 50% of RATED THERMAL POWER,

2. A nominal 75% of RATED THERMAL POWER, and l

3. Within 24 hours of attaining greater than or equal to S5% of RATED THERMAL POWER.

i l SURVEILLANCE REQUIREMENTS 4.2.3.1 The provision of Specification 4.0.4 are not applicable. 4.2.3.2 Th: ::: bin:ti:r Of 'ndicat+d-MS- tet:1 ' leu rate :nd R shall be deter-mined to be within th: r:gt:n Of ::ceptad1: Op:r:ti:n Of Figur: 3.2-3: its N3 hf wsin3 w ,m..w. inme atiecte., to .:6c in A ,w . e Ja rm ute W: -

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

b. At least once per 31 Effective Full Power Days.

c. ne. meuma F$ un u ncremd b an yfe.- m A 5 >< a+ t *a

u lut/.

         * ' 3. 3.

Th; iadice::d RCS-tetel-f4ew-rate-sha4+-be-vteif fed-to-be-within-the-regio cceptable operation of Figure 3.2-3 at least once per 12 hours'ivhen the most rec . obtained value of R, obtained per Specifica 4'2.3.2, is assumed to exist. 4.2.3.4 The RCS total flow rate at all be subjected to a CHANNEL yALIBRATION at least once per 18 shcIl be calibrated within measurement instrumentation mq s prior to the p ance of the calorimetric ' flow measurement.

6. 2. 3 e RCS total flow rate shall be determined by precision hea nesskremeet-et-leest-ene: per 18-months. balancj~

COMANCHE PEAK - UNIT 1 3/4 2-10

( . . IXX-88512 ATTACHMENT 5 PAGE 15 Of 25 POWER DISTRIBUTION LIMITS o V. 3/4.2.4 QUADRANT POWER TILT RAT 10 LIMITING CONDITION FOR OPERATION 3.2.4 The QUADRANT POWER TILT RATIO shall not exceed 1.02. APPLICABILITY: MODE 1, above 50% of RATLJ THERMAL POWER *. ACTION:

                   -e:-     With the QUADRANT POWER TILT RATIO determined to exceed 1.02

1. Calculate the QUADRANT-90WED TILT onTIO It !elst cnce per heur

                                  -tmM1--ett4m r:
                                    .4,          v, ,k, ,. s.

All AORAU.T. D .A.uf

                                                                           .          -. D. .      T. .? 4.T. D A T T. A. 4. ,.

m. m. +. a. o 4. +. A. 4 n

                                                                                                                                         ...A . . e. n. A.            .
                                                 -its limit, er
                                 .us             tur nui s nnuen                      2. __2..__;              ._ 3 __           au__        enw          , naven vornum.

Wj v r i k s s. 6m h I wwkn Ia i swwksw ww isaa wiius s Fvg vi nn I I.W u u l h as an h

                                                .nm,-

r vw on.

                           -2.a. Within 2 hours OithOr:                                                4                               ,

s, ,n,,.,2.... .w. n n i..n n,,a u.r. .n n.ue .. n, v, I-Ev. n i,v. e. n.

                                                                                                                                              ..      . . . u, .       4..,-                    IM 0051
      .gI                                               . .. .... 3                    .                                                      .. - . . .                 ..
 'V                                               limit, or
                                  -b-)               educe THERMAL POWER at least 3% from RATED THERMAL POWER I                                             b[foreach1%ofindicatedQUADRANTPOWERT excess of I and similarly reduce the Power Range Neutron l                                                 Flux-High Trip Setpoints within the next 4 hours.

i l J % G.T A -9 "edfy-that-the-QUADRANT-POWER TILT "ATIO is w'thir, its Timit-vtthirr-24 hours ef ter exceeding the 'imit Oc reduce T"En"iAL _nnues . 1... enW .# nivPm vufnuAf nN frn .JAuf. aL_ __ma_ t vnkn ww ruaa w . - vvm is vi nnukw s i s h nt sn h a wwkn n i wassis wins assaw-2 h00F0 Ind 70dWGe-th0 .0^'0I ROIiQC .00tTCI,ilun-High 0 ITipmm e n i .C a. + n n { n.e. ,.. . +

                                                               ..a 1.... a n , ',h..            a         -.1          ..
                                                                                                                       =v eew      .s     naven vor nu m ,

unikw s i n. nr in k r v w i. n 7- . .. .. y.. wem vi l r withi- th0 n0xt ' heur0; Ind l d

4. I dO I.%rNy 0 "d 04PfleC-t-th 0 00 U 0 0 O f t h,,,w,m O O u t ... ".O nnuen f ' i "i tA 0n0eFn .in-t tiO ne4ne +n (necanc4nn TLJf Du A t'= DAuco, a " u _

                                                                                                                     =s     **w"*M"*8'*                'V"b"        V              'V" r-    --- --
                                                                '-----a                  ' ' = ' -

ID0'.'O h3 0 f Day {Q Q,W A,{ ,Dg' gD { inn s pyggnad pygt,,jd a g' b "" 'th3( thg. m..nnou, mm ,r n ,,i, n l aws n y. T.a

                                                                                                              ..__2      ,,_2 usi i s isw
                                                                                                                                      .2o_
                                                                                                                                    .wi wii i n       2. wa

_ ,2_,. _. , ismiw aw isaaw_ _ _ . _ yvnwnnoi r w w u r,

                                 -0000 p"T h00F 707 12 h0GTO 07 UP.til '.'OTi#iid OCO2pt& Die Gt 0$$

n e.

                                   -        n, e. a m t.c e.
                                                 --              D A T. r n T W C D M A.f.
                                                                        --         ...               D A. u. . f. D l

l *See Special Test Exceptions Specific tion 3.10.2. O 1 l l COMANCHE PEAK - UNIT 1 3/4 2-11

                                                   ;{;                .    :-. ' . -:;..   .
                                                                                                  ;'           .G
                                                                                                                                                                                               ~a 4

TIX88512 ~.,,- ; AliAC M NT 5 '

          .               PAGC 16 0F 25
                                                                ~-            -
                                                                            - -> - c:    .            ,   5.;                                                                                     .

INSERT 4 s ,

                                                                                           * ,                          e

b. Within 24 hours a d every 7. days thereafter,' verify that Fo(Z) (by Fxy evaluation) and F g are within their limits by performing Surveillance Requirements 4.2. 2 and 4.2.3.2. THERMAL POWER.and setpoint reductions shall then be in accordance with the ACTION statements of Specifications 3.2.2 and 3.2.3. ,

4~ -

~~"

e 9 D* G 1 e 0 e _-m . m. < *%' =======-% . . - . . g, "Jf a g 4

                                                                                                                                .h                            .

O -

4 1

TrX-C512 l ATTACHMENT 5 PAGE 17 0F 25 , POWER DISTRIBUTION LIMITS i SURVEILLANCE REQUIREMENTS 4.2.4.1 The QUADRANT POWER TILT RATIO shall be determined to be within the limit above 50% of RATED THERMAL POWER by:

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

b. Calcula+'3g the ratio at least once per 12 hours during steady-state operatio,,. shen the alarm is inoperable.

4.2.4.2 The QUADRANT POWER TILT RATIO shall be determined to be within the limit when above 75% of RATED THERMAL POWER with one Power Range channel inoperable by using the movable incore detectors to confirm indicated QUADRANT POWER TILT RATIO at least once per 12 hours by either:

a. Using the four pairs of symmetric thimble locations or

b. Using the Movable Incore Detection System to monitor the QUADRANT POWER TILT RATIO. subject to the requir;;;r.t of 4;cificati;r. l

                               . . . . . .                                                        10 : 0052 O'

l l l lO I COMANCHE PEAK - UNIT 1 3/4 2-12

IIXC512 AliACMT : PAGE 18 0F 25 s POWER DISTRIBUTION LIMITS U 3/4.2.5 DNB PARAMETERS I

               , LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB-related parameters shall be maintained within the stated limits:

a. Indicated Reactor Coolant System T,yg <, 592'F 1 1 0050

b. '

c. IIndicatedPressurizerPressure>(2207)PSIG*/

nh.d ed. hd or APPLICABILITY: MODE 1. Co. 6 t s & pg, , g gq' ,Mm + 54rn]4 ) D ACTION: With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours. SURVEILLANCE REQUIREMENTS 4.2.5.l Each of the above parameters shall be verified to be within its limits at least once per 12 hours. L heT 4.2.5.2 The RCS total flow rate shall'be verified to be within its limits at least once per 31 days by plant computer indication or measurement of the RCS

elbow tap differential pressure transmitters' output voltage.

4.2.5.3 The RCS loop flow rate indicators shall be subjected to a CHANNEL CAllBRATION at least once per 18 months. The channels shall be normalized based on the RCS flow rate determination of, surveillance 4.2.5.4. Its 00s0 4.2.5.4 The RCS tstal flow rate shall be determined by precision heat balance measurement after each fuel loading and prior to operation above 75'4 of RATED

    .        THERMAL POWER.        The feedwater pressure and temperature, the main steam pressure, and feedwater flow differential pressure instruments shall be calibrated within 90 days of performing the calorimetric flow measurement.

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

O' kW Tnc\cdf.S o-I.6 9. hlow merso re. rne tsk u nce.r bo.6k . COMANCHE PEAK - UNIT 1 3/4 2-13

  .    !XX 88512 AITACHMENT5 PAGE 19 Of 25.     .

3/4.2 POWER DISTRIBUTION LIMITS BASES _ The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by: (.1) maintaining the minimum DNBR in the core greater than or equal to 1.30 during normal operation and in short-term transients, and (2) limiting tne fission gas release, fuel pellet temperature, and cladding mechanical properties to within assumed design criteria. In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200*F is not exceeded. The definitions of certain hot channel and peaking factors as used in i these specifications are as follows: Fq (Z) Heat Flux Hot Channel Factor, is defined as the maximum local neat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for manuf acturing tolerances on fuel pellets and rods; Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of Fh the integral of linear power along the rod with the highest integrated l O I p)wer to the average rod power; and Radial Peaking Factor, is defined as the ratio of peak power density F,y(Z) to average power density in the horizontal plane at core elevation Z. 1 3/4.2.1 AXIAL FLUX OIFFERENCE The limits on AXIAL FLUX DIFFERENCE (AFO) assure that the F (Z) upper 0 bound envelope of 2.32 times the normalized axial peaking factor 1s not exceeded during either normal operation or in the event of xenon redistribution following power changes. Target flux difference is determined at equilibrium xenon conditions. The iNil ur.;O :ods may be positioned within the core in accordance with their respective insertion limits and should be inserted near their normal position for steady-state operation at high power levels. The value of the target flux difference obtained under these conditions divided by the fraction of RATED THERMAL POWER is the target flux difference at RATED THERMAL POWER for the associated core burnup conditions. Target flux differences for other THERMAL POWER levels are obtained by multiplying the RATED THERMAL POWER value by the appropriate fractional THERMAL POWER level. The periodic updating of l the target flux difference value is necessary to reflect core burnup l considerations. Q - COMANCHE PEAK - UNIT 1 8 3/4 2-1 ,

t .

.      TIX-88512 AllACHMENT 5 PAGE 20 0F 25    .

POWER DISTRIBUTION LIMITS

  . BASES AXIAL FLUX DIFFERENCE (Continued)

Although it is intended that the plant will be operated with the AFD within the target band required by Specification 3.2.1 about the target flux difference, during rapid plant THERMAL POWER reductions, control rod motion will cause the AFD to deviate outside of the target band at reduced THERMAL POWER levels. This deviation will not affect the xenon redistribution suffi-ciently to change the envelope of peaking factors which may be reached on a subsequent return to RATED THERMAL POWER (with the AFD within the target band) provided the time duration of the deviation is limited. Accordingly, a 1-hour penalty deviation limit cumulative during the previous 24 hours is provided for operation outside of the target band but within the limits of Figure 3.2-1 while at THERMAL POWER levels between 50% and 90% of RATED THERMAL POWER. For THERMAL POWER levels between 15% and 50% of RATED THERMAL POWER, deviations of the AFD outside of the target band are less significant. The penalty uf 2 hours actual time reflects this reduced significance. Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitor Alarm. The computer deter-mines the 1-minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for two or more OPERABLE excore channels are outside the target band and the THERMAL POWER is greater e than 90% of RATED THERMAL POWER. During operation at THERMAL POWER levels between 50% and 90% and between 15% and 50% RATED THERMAL POWER, the domputer

  . outputs an alarm message when the penalty deviation accumulates beyond the limits cf I hour and 2 hours, respectively.

Figure B 3/4 2-1 shows a typical monthly target band. 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR / and AGS-fCd RATE ANO l NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR / UI 000 RCS ficw rete, and nuclear enthalpyThe rise limits onchannel hot heat flux hot channel factor factor /(1) the design limits on peak ensure that: local power density and minimum DNBR are not exceeded and (2) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200'F ECCS acceptance criteria limit. Each of these is measurable but will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3. This periodic surveillance is sufficient to ensure that the limits are maintained provide J:

a. Control rods in a single group move together with no individual rod insertion differing by more than 2 12 steps, indicated, from the group demand position;

b. Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.6; COMANCHE PEAK - UNIT 1 8 3/4 2-2 ,

____.__._,..._____....._.._........-.__...__..-__.___.____._.____.__.~~_m_

          .          . ..'                     TXX-86512:                                                                        ,

AfiAC M NT 5 PAGE 21 0F 25 , g, i' : t 4 ! i

i

! i a i- !

h i

t f l l ! ( l t i r

l i. 1 b ( I 1 i I FIGURE 8 3/4 2-1 : 1 TYPICAL INDICATED AXIAL FLUX DIFFERENCE VERSUS THERMAL POWER ! 9 ; l COMANCHE PEAK - UNIT 1 8 3/4 2-3 l 9 4 -- . mr<- ...-- . . . , - . . _ . . . . . . . . , . . P _ _ j

TIX-88512 AllACHMENT 5 - tkGE 22 0F 25 _ _ ,. POWEF DISTRIBUTION LIMITS w/ BASES HCAT FLUX HOT CHANNEL FACTORf and "CS "LCh' "ATE-Ahs NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued)

c. The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained; and The axial power distribution, expressed in terms of AXIAL FLUX

d. DIFFERENCE, is maintained within the limits. N F will be maintained within its limits provided Conditions a, through y

d. above are maintained. As .,sted-en figure 3.2-3, CCS fic., rate and f" :y be "traded off" agatast one another-4Le,,-a low-measwed RCS 'les rate is N

             -eeeept ble if the-meesured F g i: 210 10w) t: en gre th t the calculated DNES will not be belew the d::ign DMe" :lue.                                    The relar.ation of F H as a function   W OM7 of THERMAL POWER allows changes in the radial power shape for all permissible rod insertion limits.
      -                 R es calculated -in Specification 3.2.3 cfte used in Figure 3.2-3, ::ccunt:
   /                    P T g less 4 hen er equel to 1.40.                             TMt-value is used ift-the-verfeus-ace 4deftt-G)[        for an;1yses where F g M uences per d ers M 5er Ger. E , e.g., pesk d ad 4emperatwe, :nd the: 10 the m:*4mur " : :Ossured" v;lue allowed.

Fuel rod bowing reduces the value of DNB ratio. Credit is available to offset this reduction in the generic margin. The generic margins, totaling 9.1% DNBR completely offset any rod bow penalties. This margin includes the following:

a. Design limit DNBR of 1.3U vs 1.28,

b. Grid Spacing (K,) of 0.046 vs 0.059,

c. Thermal Diffusion Coefficient of 0.038 vs 0.051,

d. DNBR Multiplier of 0.86 vs 0.88, and

e. Pitch reduction.

The applicable values of rod bow penalties are referenced in the FSAR. O COMANCHE PEAK - UNIT 1 B 3/4 2-4 .

When FM is measured, an adjustnient for measurement uncertainty must be

                    ,    included for a full core flux map taken with tne incore Detector Flux Mapping System. -
                                                                       ~

POWER DISTRIBUTION LIMITS a : l

          /            BASES f

HEAT FLUX HOT CHANNEL FACTOR' / and RCS TLOW RATE-ANO- NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued) I When an qF measurement is taken, an allowance for both experimental error i and manufacturing tolerance must be made. An allowance of 5% is appropriate 1 for a full-core map taken with the Incore Detector Flux Mapping System, and a allowance is appropriate for manufacturing tolerance. The Radial Peaking Factor, Fxy(Z), is measured periodically to provide

      .               assurance that the Hot Channel Factor, F (Z), remains within its limit.               The 0

F limit for RATED THERMAL POWER (F RTP) as provided in the Radial Peaking xy xy Factor Limit Report per Specification 6.9.1.6 was determined from expected power control manuevers over the full range of burnup conditions in the core. H Whe&RGS-f4ow-rate-an44 pro-seasured,-no-addi4 tonal-a1!ewances are ni 0W neceseery prior to-compeefson-with-tb-limits-of-Hgures-a.F3-and 0. 2- 4. N Meesurement errers of-{-lh-1-3%-fee-RCS-1.otel-flow-rate-and-4Ffoe-F3g haeh a44 owed-fof40-detefetnatica of the-deTigfT-OfTBR value. (") g V The ::asurement-eeroe-fee-RGS-total-flow-cate-is-based-upon performing a preci icn h :t-balance-and-us4ng-the-result-to-ce44brate the RCS flow rate indicators. Pctentisi fouling Of the-feed-stwe venturi wh h.h miwhi n0L be

                   -deteel;d could bies-the-restrit-free-the precision heet belence in e noir conservative manner. Thereferera-penalty-of-f0-1-3Ffor undetected-foul 4ng-of-
                  -the feedwatee-venturi is inSuded-fn-Figure 3. 2-3.           Any fotriing hich si.ight-
                  -bfas-the RCS ficw rate-measurement-greater-then [0.13% can be detected by-
                   -menftering and trending varicus plant- performer.ce per; meters. If detected, a tica shell be teken befor: perf0 ming :ab; quent preci; ion he t balan:0-measur;;;nt;, fren-efther th; cffect of the fouling shall be quintified Ond compensated for in the PCS ficw rcte-measurement-ee-the venturi shell b; cleaned to elii;.inete the fouling.

The 12-50er periedic survei' lance of indicated D.C5 #1ew is su "icieat te detect caly ficw degradetion which-cculd ! cad to operation cutside th; : cept-ahle eqinn M naaratica sheva er eiger: 3, 2 4-- 3/4.2.4 QUADRANT POWER TILT RATIO

                -r o% 958           -i>

The QUADRANT P0t!ER TILT RAT 40-14mit-assures-that-the-cadial-powee-di+tribu-44+n-sa tri sties-the-de sign-value s- u sed-in- the- powe cap:bi'ity :n:!ysis; Redte4-power-dtstf4butien-measurements-ammade during STARTUP testing and p e r i o d i caMy4u ring-powe r-ope ra t4em. El 0051 he-14mtt-of-1 02 rat-whieh-correctiire- action i s r. qui red , prvv ide> dig s end-14neae-heat-generation-rate proteet4en with x y piene pc .r Lii15. A Ifmit of 1.02-was-s+1ected to provide aa a1!cwance fe"

nneartaintv

[v) m ci.t - h-the-indi ted#wertot. 111-88512 COMANCHE PEAK - UNIT 1 B 3/4 2-5

ATTACHMENT 5 PAGE 23 Of 25

                                           .x                                                                          -

g TXX 88512 . . - ' ATTACHMENT $ . ' fAE 24 0F 25

                                                                                                                                                          ~ ~ ~ ~ ~ ~ ~ -                                                       ' ~ ~

INSERT [, o V The Quadrant Power Tilt Ratio limit assures that the radial power distribution satisfies ' core design values by detecting gross changes in core power distribution between monthly incore flux maps. During normal operation, the QUADRANT POWER TILT RATIO is set equal.to zero once acceptability of core peaking factors has been established by review of incore maps. The limit of 1.02 is established as an indication that the power distribution has changed encugh to warrant further investigation. 4 O n 0 O . _ _ _ _ _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ - - _ - - - - _ - - - - - - - - - ' - - - - - - - - - - - - - - ~ ~ ~ ~

TXX-C'.512 AliACHMENT5 PAGE 25 0F 25 I POWER DISTRIBUTION LIMITS DRMT l l l BASES QUADRANT POWER TILT RATIO (Continued) The 2-hour time allowance for operation with a tilt condition greater than 1.02 is orovided to allow identificat:on and correction of a dropped or misaligned control rod. In the event such action action does not correct the tilt, the margin for uncertainty on F is reinstated by reducing the maximum 9 allowed power by 3% for each percent of tilt in excess of 1. For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperabla, the moveable incore oetectors are used to confirm that the normalized symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO. The incore detector monitoring is done with a full incore flux map or two sets of four symmetric thimbles. 3/4.2.5 DNB PARAMETERS The limits on the DNB-related parameters assure that each of the param-eters are maintained within the normal steady-state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the initial FSAR assLaptiens and have been analytically demonstrated adequate to maintain a minimum DNBR of 1.30 throughout each analyzed p transient. The indicated T,yg value of 592.7'F (conservatively rounded to OI 592*F) and the indicated pressurizer pressure value of 2207 psig correspond to analytical limits of 594.7'F and 2193 psig respectively, with allowance for measurement uncertainty. The indicated uncertainties assume that the reading from four channels will b. averaged before comparing with the required limit. The 12-hour periodic surveillance of these parameters througa instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and ot2)er expected transient operationf 3 a_%d,, I h Adec.t @ c.ut 4\ce ckegradL di o n elde

           ~~L. A Ca.g                 A A.        s;y%dem .
        *10The additional surveillance requirements associated with the RCS total flow                                          In con rate are sufficient to ensure that the measurement uncertainties are limited to 1.8% as assumed in the Improved Tneraal Design Procedure Report. for CPSES.

Performance of a precision secondary calorimetric is required ;.o precisely determine the RCS temperature. The Transit Time Flow mater, 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. O COMANCHE PEAK - UNIT 1 B 3/4 2-6 w.-- - . . . . - . . - - . - . - , . . . . . . . , . - . _ _ - -

               -A-   -                        *J _J           - --_-

T!X-885!2 AitkCHP.ENT 6 PAGE 1 0F 105 O' COMANCHE PEAK STEAM CLECTRIC STATION TECHNICAL SPECIFICATION 3/4.3.1 O O l

Txx-88512 ATTACMENT 6 CPSES Technical Specifications PAGE 2 of 105 NRC Draft 2 Markup Section 3/4.3.1 O Chanae 108 Justificati;n For Chance V 0057 This Table is being relocated to the CPSES Technical Specification Improvement Program. TV Electric believes the inclusion of this Table is unnecessary and the information would be more appropriately addressed in the CPSES Technical Specification Improvement Program. Relocation of this Table is consistent with the guidance provided in the NRC's Interim Policy Statement (52FR3788), February 6, 1987, and the recommendations of the Westinghouse Owners Group MERITS Program. Priority is given to the relocation of this Table since the detailed information is not used by the Licensed Operator, but is purely used as testing criteria provided by the vendor supplier. The information currently in this Table is more appropriately maintained in a document subject to TV Electric administrative control and 10CFR50.59 review under the CPSES Technical Specification Improvement Program. This change is similar to that Licensed at Shearon Harris, Seabrook and Vogtle. 0058 This Specification has been changed to clarify the p V 0079 0081 specific requirements of the Source Range Nuclear Instru-mentation (SRNI) and the Boron Dilution Flux Doubling 0085 Instrumentation. The Source Range Neutron Detectors supply input to the Boron Dilution Flux Doubling and Source Range Neutron Indication (which provides Reactor Trip Protection). The SRNI and the Flux Doubling are two separate drawers, feeding different logic circuits in the Solid State Protection System to provide protection from either a Continuous Rod Withdrawal Accident (CRWA) or a boron dilution accident, respectively. Since these functions provide protection from different accidents the required Limiting Condition For Operation and Action Statements should also be written for the specific accidents of concern. The first accident, CRWA, is mitigated by a source range trip or the operator taking action from visual indication therefore, while in Modes 2 through 5 with the reactor trip breakers shut and a channel is lost, the Action is to open the reactor trip breakers within 48 hours. If the reactoe trip breakers are open while in Modes 3, 4 and 5 the only requirement is that there be a single indication channel available since no trip function is required with the reactor trip breakers already open. This change is similar to that Licensed at Seabrook and Millstone. O

txx c 512 Attacm e t 6 CPSES Technical Sp;cifications PAGE 3 0F 105 ' NRC Oraft 2 Markup

                     . .              Section 3/4.3.1 y

V Change 10# Justification For Change A new Action Statemeat was added to address the situation of when no Source Range channels are Operable. This action is added to give specific direction as to what action to take in the event that no channels are Operable. Without this direction the generic 3.0.3 action would be followed which for this situation is in the nonconservative direction since there would be no neutron-monitors but positive reettivity would be added due to the required cooldown and if in Mode 5 there is no direction at all since the 3.0.3 action does not apply. This change is similar to that Licensed at Shearon Harris, Byron and Vogtle. The second accident, Boron Dilution, is mitigated by the Flux Doubling - Boron Dilution mitigation signal for Modes 3, 4 and 5. Mode 2 was not selected since the Flux Doubling signal would be blocked for the startup. The Flux Doubling signal would also have to be blocked in Mod 5 3 during a startup due to the fact that when the startup is commenced (i.e. rods pulled) the subtritical multiplication would trip the flux Doubling Bistable causing a spurious trip, resulting in Boron injection. The Flux Doubling Function has been separated from the (- SRN! trip and indication function since one of these s functions could be inoperable but have no affect on the other function. The way it is written now in the Standard Technical Specification the actions for both have to be performed when either function is inoperable. This is considered to be overly restrictive action for CPSES' specific design. 0079 See 10# 0058 0080 Change time requirement of Action 2a for Table 3.3-1 from 2 hours to 6 hours per WCAP-10271 and agreed on in the 1/11/88 onsite meeting. 0084 Delete the last sentence in Action 11 for Table 3.3-1 to prevent confusion of the interrelationship between Action 11 and Action 8. Another utility, which has the identical RTB Specification, was cited for a violation when they bypassed a RTB for 8 hours to repair a shunt trip attachment which was found inoperable during routine surveillance testing. Although this action is permitted in ACTION 12, the utility was cited for a tech spec violation on the basis that bypassing the RTB rendered it inoperable which made ACTION 8 applicable. Thus the utility violated the requirements of ACTION 8. , (3 (') 0081 0085 See 10# 0058 See 10# 0058

Txese512 AllAcimENT 6 '

          . PAGE 4 W 105                         CPSES Technical Specifications NRC Oraft 2 Markup
                                    ..    .              Section 3/4.3.1 m

Change 10# Justification For Change l- 0113 Westinghouse has calculated tne impact of the bypass

        ,         0125                      breaker failure probability on the reactor trip system L

0126 failure probability and concluded that the bypass breaker

. contribution is insignificant. These calculations are i based.on the trip breaker fault tree model presented in Supplement 1 to WCAP-10271.

In 0G-106, the Westinghouse response to NRC questions on WCAP-10271, a typical Westinghouse PWR repctor trip unavailability is estimated to be 1.5x10-3 No credit l- was taken for operation of the bypass breaker in the- ). evaluation from which these calculations were derived. Westinghouse calculated the impact on this unavailability of the reactor trip bypass breakers with the following results:

                                      .1. The bypass breakers are placed in service only when one train of the RPS is in test. The only circumstance in which the bypass breaker could affect RPS unavailability is the situation when one train is in test, a signal is generated in the operable redundant train and the main breaker fails to open.

O 2. The unavailability of the RPS attribute to failure of a maintrjpbreakerwiththeoppositetrainintestis 3.7x10 1.5x10- 5).orThis 2.5% of the total situation RPS unavailability constitutes the only (i.e. configuration in which the bypass breaker can affect RPS unavailability.

3. Taking credit for the bypass breaker would reduce the probability value of this situation to (3.7 x 10-7)(3.5 x 10-4) = 1.3 x 10-10 where 3.5x10-4 is the unavailability of the bypass breaker assuming bimonthly testing, or

                                  .         where 3.5x10-3 is the unavailability of the bypass breaker assuming testing on an 18 month interval.

Based on the above, testing of bypass breakers is not included in the technical specifications. As shown above, testing the bypass breakers on a 2 month or 18 month test interval will result in a 10-9 or 10-10 1 eye 1 congributiontotheRPSunavailabilityofapproximately 10 . Alternatively, the RPS unavailability increase that occurs by increasing the bypass breaker failure probability from O'4 to 100% is only 2.5% at the RPS level. 0125 See ID# 0113 0126 See 10# 0113 ____ _ _ _]

ixx-88512. NN3l5 CPSES Technical Specifications NRC Draft 2 Markup

                                      , .               Sectirn 3/4.3.1 A      Change 10#       Justification For Change Q1 0118       Table 4.3-1, changed notations 3 and 6 from a calendar-based frequency to an Effective Full Power Day (EFPD) frequency since the nature of this specification is specifically related to fuel burnup. That-is, flux profiles change with fuel burnup, not calendar time, ibis change was previously agreed upon in the 1/11/88 meeting with the NRC.

0121 Change power restriction from 75% to 50% for notation 6 of Table 4.3-1. The goal of the-Incore-Excore calibration is to achieve the means of knowing, within acceptable bounds, what the incore axial power distribution is, based upon the excore detector currents. A calibration performed at 50% RTP achieves this goal as well as one. pe formed at 75'4 RTP; therefore this proposed change can be affected with no impact on safety. A large number of Westinghouse plants have performed their Incore-Excore calibration at 75% power; one has recently performed it at 50% power. In either case, a linear relationship between the incore axial offset and the excore axial offset is obtained. One can do a check on any calibration with the following infonnation available: A a flux map at power and the corresponding excore carrents

V recorded during the map. Using the above mentioned linear relationships and the excore currents, one obtains a,1 inferred incore axial offset which can then be directly compared to the measured incore axial offset from the flux map. If these offsets agree within several percent, it is

, reasonable to say that the calibration, regardless of the power level at which it was obtained, does a sufficiently good job of representing what the incore power distribution is. Table 1 presents the calibration constants associated with axial offset for 18 Westinghouse plants; 17 of the calibrations were performed at 75%. It is evident that the values of these constants show a bit of variation and that the values obtained at 50% fit in quite well with the rest of the group. Aftcr these calibration results were obtained, one to four additional flux maps were obtained during the initial startup program. The results of comparing each inferred axial offset with the incore measured values are presented in Table 2 as the measured value minus the inferred value. Again, the results for the calibration performed at 50'4 fit in well with those performed at 75%. Figure 1 shows the distribution of differences; nearly all of them are l'4 or less. The standard deviation of the differences with

                    \                     the 50% calibration is .605'4 and it is .803% for the 75%

calibration, l

TXX-88512 ATTACHrENT 6 PAGE 6 0F 105 1 i TABLE 1 O AXIAL OFFSET CONSTANTS

CAL. MAN M41 CHAN N42 CHAN N43 CHAN N44 POWER N b M b M b M b l 504 1.742 20.60 1.654 -0.33 1.653 9.24 1.626 -0.25 754 1.832 2.55 1.714 2.38 1.755 -0.33 1.706 5.35 75% 1.851 4.56 1.929 2.08 1.894 3.77 1.867 3.70 75% 1.910 18.50 1.810 13.00 1.847 13.90 1.773 14.10 75% 1.440 4.71 1.470 5.56 1.461 5.30 1.488 6.07 754 1.682 0.47 1.621 -3.67 1.680 -3.11 1.581 2.48 754 1.329 6.46 1.344 8.84 1.327 4.00 1.333 19.90 754 1.771 -0.85 1.794 -0.78 1.844 -2.61 1.722 -0.59 75% 1.824 21.70 1.583 5.85 1.802 20.40 1.595 7.78 75% 1.628 5.69 1.654 5.55 1.645 6.19 1.665 1.51 75% 1.508 6.82 1.513 9.10 1.542 10.10 1.468 15.20 7E% 1.641 1.79 1.681 -4.09 1.726 -3.97 1.626 -6.71 75% 1.663 0.75 1.603 0.59 1.652 1.57 1.605 -1.59 75% 1.695 7.38 1.655 4.79 1.648 3.28 1.759 3.35 754 1.710 5.95 1.674 5.42 1.701 3.12 1,658 -7.31 754 1.447 -0.13 1.509 3.72 1.466 0.76 1.473 3.20 75% 1.720 -3.90 1.603 -0.03 1.643 3.16 1.614 -2.75 75% 1.508 1.53 1.472 2.06 1.499 4.42 1.495 0.71 Where: (AO-IN) = (AO-EX) *M+b I M = Slope b = Intercept i i O . T-T F

TIX48512 AfiACHFINf 6 l PATE 7 0F 105 l TABLE 2 O AXIAL OFFSET DIFFERENCES CAL. POWER POWER LEVEL M41 N42 N43 N44 50% 72.9% -0.15 0.49 -0.29 0.20 50% 90.1% -0.92 -0.44 -1.10 -0.36 50% 98.6% -0.99 -0.82 -1.58 -0.31 50% 99.3% -1.37 0.46 -1.43 -0.17 754 88.04 -0.60 -0.77 -0.26 -0.51 75% 100.0% -0.04 -0.32 0.44 0.14 75% 90.0% -0.48 0.07 -0.96 -0.21 75% 99.0% -1.39 -3.05 -1420 -1.77 75% 89.7% 0.12 -0.64 -0.03 -1.08 75% 98.2% -0.56 -1.59 -0.26 -1.89 75% 90.9% -1.65 -0.33 -0.91 -0.94 75% 99.7% -2.07 -0.40 3.56 -1.09 75% 90.0% 1.67 -0.99 -0.45 0.98 75% ; 100.0% -1.03 -0.58 -1.00 0.88 75% 91.34 -0.82 NA 0.74 NA 75% 99.9% -2.07 0.71 -1.77 -1.90 75% 90.0% -1.65 -1.60 -1.72 -1.83 75% 100.0% -1.59 -0.97 -1.10 -1.50 75% 87.5% 0.19 0.00 0.97 0.28 75% 87.2% 0.48 0.24 0.76 0.47 75% 99.9% 1.07 0.36 1.14 0.74 O 75% 75% 754 97.2% 89.8% 100.5% 0.59

                                          -0.18
                                          -0.43 0.14
                                                        -0.15
                                                        -0.50 0.78
                                                                    -0.18
                                                                    -0.04 0.40
                                                                                          -0.35
                                                                                          -0.52 75%          90.0%       -1.49         -1.26       -1.17                 -1.74 75%          98.0%       -0.57         -0.00       -0.33                 -0.56 75%          90.4%       -0.70         -0.68       -0.66                     NA 75%          99.2%       -0.66         -0.71       -0.24                 -0.82 75%          99.8%       -0.69         -1.25       -1.62                 -0.83 75%          90.04       -0.21         -0.00       -0.08                 -0.14 75%          99.04       -0.34         -0.03       -0.35                 -0.36 75%          90.1%         0.47         0.61        0.57                  0.26 75%          99.6%       -0.08         -0.12       -0.02                  0.01 75%          61.7%       -0.38          0.47        0.15                  0.07 75%          95.8%       -0.54          0.87       -0.09                 -0.37 75%          90.0%         0.78        -0.34        0.32                 -1.42 75%          99.2%         0.34        -1.10       -0.93                 -2.14 754          90.4%       -0.61         -0.06      -0.49                  -0.75 75%          99.7%       -0.45         -0.21       -0.32                 -1.13 O .

IIX-83512 AllACHMEHi 6 PAGE 8 Of 105 FIGURE 1

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TIX 88512 AHAM 6 CPSES Technical Specifications PAGE 9 0F 105 NRC Oraft 2 Markup

                                     , .                     Section 3/4.3.1 Change 10#                    dustification For Chailge 0121 (cont.)             Additional support for this position can be provided if required. This method of performing incore-Excore Calibrations is documented in WCAP-9648 and is routinely practiced at both the McGuire and Catawba Nuclear Stations.

0123 Added a 4.0.4 exception to notation 9 of Table 4.3-1. The operability of the Source Range Nuclear Instrumentation is required in MODES 2 (below P-6), 3, 4 and 5 by Table 3.3-1, item #6. Table 4.3-1 requires a quarterly Analog Channel Operational Test (in addition to 18 month surveillances) which includes checks of the Boron Dilution Alarm. During normal MODE 1 Power Operation there are no operability requirements; therefore, pursuant to Specification 4.0.3 surveillance is not required and the instrument can be considered technically inoperable when the surveillance intervals of Specification 4.0.2 are exceeded. Subsequently, for a normal plant shutdown, the lapsed surveillances would be performed prior to power decrease below P-6 pursuant to Specification 4.0.4 and the Source Range instruments declared OPERABLE. In the event of a {- Reactor Trip, however, the MODE changes from MODE 1 to 3 x are not under the control- of the operator but occur automatically. Thus, an OPERATIONAL MODE could be entered without the Limiting Condition for Operation being met - a possible violation of Specification 3.0.4. Note that the bases of 3.0.4 im)1ies that its purpose applies only going to higher MODES; lowever, this is not explicitly stated in 3.0.4 itself and could be the subject of debate. Further more, 3.0.4 wouldn't apply if this was a forced shutdown to comply with an ACTION requirement. Adding an exception to 4.0.4 will clarify this situation. O V

III-88512 ATTACHMENT 6 PAGE100F105 3/4.3 INSTRUMEN1 ATION O 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION O LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the Reactor Trip System instrumentation channels and interlocks of Table 3.3-1 shall be OPERABLE with Reactor Trip System RESPONSE i TIMES _. ....-. . ... _.. .. WitW n u e. W \ , W A s . l ID I: 0057 APPLICABILITY: As shown in Table 3.3-1. l ACTION: As shown in Table 3.3-1. SURVEILLANCE REQUIREMENTS _. 4.3.1.1 Each Reactor Trip System instrumentation channel and interlock and the automatic trip logic shall be demonstrated OPERABLE by the performance of the Reactor Trip System Instrumentation Surveillance Requirements specified in Table 4.3-1. f 4.3.1.2 The REACTOR TRIP SYSTEM RESPONSE TIME of each Reactor trip function shall be demonstrated to be within its limit at least once per 18 mo' ths. Each test shall include at least one train such that both trains art. tested at least once per 36 months and one channel per function such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific Reactor trip function as shown in the "Total No. of Channels" column of Table 3.3-1.

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

O - O O - TABLE 3.3-1 ; =.; = - a M: 7 REACTOR TRIP SYSTEM INSTRUMENTATION u~ 9 MINIMUM g* TOTAL NO. CHANNELS CHANNELS APPLICABLE A FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE NOGES ACTION R

   . 1. Manual Reactor Trip                       2                1           2    1, 2               1                      ,

8 8 E 2 1 2 3 , 4*, 5 9 Z g 2. Power Range, Neutron Flux b

a. High Setpoint 4 2 3 1, 2 2 d D

b. Low Setpoint 4 2 3 I,2 2 _

b

3. Power Range,teeutron Flux 4 2 3 1, 2 2 High Positive Rate w 4. Power Range, Neutron Flux, 4 2 3 1, 2 2b i High Negative Rate d

  ,',  5. Intermediate Range, Neutron Flux          2                1           2    I,2                3

6. Source Range, Neutron Flux C 4 o Startup 2 1 2 2 4 , t 2.

. -A 2) Shutdown 2

?-

1 o 2 3'4I5" -S- 9 ,17-3)5 6 tA w a 1 3, 4, 5 32. 7 7 0vertemperature N-16 4 2 3 1, 2 6b D

8. Overpower N-16 4 2 3 1, 2 6

9. Pressurine Pressure--Low 4 2 3 l' 6b ,f 10 8: 0058 b

10. Pressurizer Pressure--High 4 2 3 1, 2 6 a,hclor 'Tr.7 --A TaA;c k;*" O

                                                                                         .                           llllE3 b.h a bition Flot boM.$

E I g i,4,5 S b i

O O . O 2%a n TABLE 3.3-1 (Continued) Ns% y UME REACTOR TRIP SYSTEM INSTRUNENTATION %5" 2 g-E MINIMUM m TOTAL NO. CHANNELS CHANNELS APPLICA8LE { FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE NODES ACTION D

11. Pressurizer Water Level--High 3 2 2 l' 6 .

Z 12. Reactor Coolant Flow--Low Single Loop 3/!oop b

   -                                         a.                                                        2/ loop in 2/ loop          19       6 any loop h        b

b. Two Loops 3/ loop 2/ loop in 2/ loop 1 6 any two loops w 13. Steam Generator Water 4/sta. gen. 2/sta. gen. 3/sta. gen. 1, 2 6b ,f 1 Level--Low-Low in any sim.

y gen. w

14. Undervoltage--Reactor Coolant e b Pumps 4-1/ bus 2 3 1 6 -

15. Underfrequency--Reactor Coolant a b

 -                                           Pumps                                  4-1/ bus                2              3     1          6

16. Turbine Trip

a. Low Fluid Oil Pressure 3 2 ? 1* 6'D

b. Turbine Stop Valve Closure 4 4 1 l' 10 i

17. Safety Injection Input

from ESFAS 2 2 1. 2 ' 8 l i

i 3 21

O . O O - TABLE 3.3-1 (Continue.1) ,_ 3; = = a REACTOR TRIP SYSTEM INSTRUMENTATION

                                                                                                                            ]h g                                                                        MINIMUM 2c
                                                                                                                            ,_ [

TOTAL K. CHANNELS CHANNELS APPLICA8LE 3 FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

18. Reactor Trip System Interlocks ,

e a. Intermediate Range ( Neutron Flux, P-6 2 1 2 2' 7

b. Low Power Reactor TripfBlock,P-7

1) P-10 Input 4 2 3 1,2 7

2) P-13 Input 2 1 2 1 7 -

c. Power Range Neutron Flux, P-8 4 2 3 1 7 u

2 d. Power Range Neutron w Flux, P-10 4 2 3 1,2 7

19. Reactor Trip Breakers 2 1 2 1, 2 8, 11 8 8 8 2 1 2 3,4,5 9

20. Automatic Trip and Interlock 2 1 2 1, 2 8 8

Logic 2 1 2 3*, 4 S* 9 e ll22

>=

Id

III C 512 AllACHMENT6 FAGE140F105 TA8tE 3.3-1 (Continued) ,A TABLE NOTATIONS b

  'Whenthe[eactorfripbreakersareintheclosedpositionandtheControlRod Drive System is capable of rod withdrawal, b

The provisions of Specification 3.0.4 are not applicable. C Below the P-6 (Intermediate Range Neutron Flux Interlock) Setpoint, d Below the P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.

  'Above the P-7 (At Power) Setpoint Theapplicable[oksandactionstatementsforthesechannelsnotedin Table 3.3-3 are more restrictive' and therefore, applicable.

9 Above the P-8 (3-loop flow permissive) setpoint, h Above the P-7 and below3 P-8 setpoints. ACTION STATEMENTS ACTION 1 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours or be in HOT STANDBY within q the next 6 hours. ACTION 2 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/cr POWER OPERATION may proceed provided the following conditions are satisfied a. The inoperable within 4- hours , channel is placed in the tripped condition l G 10 8: 0080

b. The Minimum Channels OPERABLE requirement is met; however,I the inoperable channel may be bypassed for up to 4 hours for surveillance testing of other channels per Specification 4.3.1.1, and

c. Either, THERMAL POWER is restricted to less than or equal to 75% of RATED THERMAL POWER and the Power Range Neutron Flux Trip Setpoint is reduced to less than or equal to 85% of RATED THERMAL POWER within 4 hours; or, the QUADRANT POWER TILT RATIO is monitored at least once per 12 hours per Specification 4.2.4.2.

T6 bocon d. ,\ oi t on 4 \o v. Ac>ob\',n3 sic [s ma.g be Meckcd., d o <'i g r e a ch o r do.A q g , , ,,n O COMANCHE PEAK - UNIT 1 3/4 3-5

. ATIACHM(Ni6 FAGE 15 0F 105 TABLE 3.3-1 (Continued)

/7 AC1 ION STATEMENTS (Continued) U ACTION 3 - With the number of channels OPERABLE one less than the Minimum Channels OPERABLE requirement and with the THERMAL POWER level:

a. Below the P-6 (Intermediate Range Neutron Flux Interlock)

Setpoir.t, restore the inocerable channel to GPERABLE status prior to increasing THERMAL POWER above the P-6 Setpoint,

b. Above the P-6 (Intermediate Range Neutron Flux Interlock)

Setpoint but below 10% of RATED THERMAL POWER, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above 10% nf RATED THERMAL POWER. ACTION 4 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, suspend all operations involving positive reactivity changes. ACTION 5 - With the number of OPERABLE channels one less than the Mini.: rum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours or within the next hour open- -

                    -the-Reactor-Te4p-Systee4reakers                                                                                                                               n -suspend-ak!       Op:r: tier.: i velv- - $

ing-p::itiv: reactivity ch: g : :nd verify either valvelltS-8455 .:: or valves ,lICS-8439,l/CS-8441, andffCS-8453 e ere closed and1JtS-8560, secured in pfes FCV-11h:ition, and verify this positio p hast once per 14 days thereafter. With no channels OPERABLE cotiplete all the above actions within 4 hours and serify the positions of the above valves at least once per 14 days thereafter. ACT:0N 6 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a. The inoperable channel is placed in t'.-e tripped condition within 6 hours, and

b. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours for surveillance testing of other ch:innels per Specification 4.3.1.1, 3 ACTION 7 - With less than the Minimum Number of Channels opt.' TABLE, within 1 hour determine by observation of the associated permissive annunciator window (s) that the interlock is in its required state for the existing plant condition, or apply Specification 3.0.3.

O o COMANCHE PEAK - UNIT 1 3/4 3-6

in 88512 f.ifACKENT 6 PAGE160F105 TABLE 3.3-1 (Continued) g ynne i ACTION STATEMENTS (Continued) ACTION 8 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours; however, one channel may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.1.1, provided the other channel is OPERABLE. ACTION 9 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requi-ement, restore the inoperable chanr.el to OPERABLE status within 48 hours or open the Reactor Trip System breakers within the next hour. ACTION 10 - With the number of OPERABLE channels ins than the Total Number of Chant.els, operation may continue provided the inoperable channels are placed in the tripped condition withir 6 hours. ACTION 11 - With one of the diverse trip features (undervoltage or shunt trip attcchment) inoperable, restore it to OPERABLE status within 48 hours or declare the breaker inoperable and appl - ACTION 8. Th;-br;;h;r :h:11 not b; by;;;;;d whil; ;,; ; M ' jiver:0t*4p-feawec: i; in;;;r:b1; ;x;;pt f;r the tim; re wired maint;n n;; to r;;t;r; th; tr;;h;7 ;; opga,gg., yy i v i - ..,

                                         , i. n . . .

10 8: 0084 ACTio4 iz - tu ak. no c.ho.n ne.is OPER W >I m edided - b r e a. kers , ses,pe.nl opn ne r e.a. c.4e v- t cy reacW H

                                  ,'          asi oyetalions WyoW q/m 4                      3 ebs ors  W v e.

Cg E3 g es o.n a. e ,cw b mwh 9 a.\ve 105-8455 of " \Ve* g s.65La ,FCV- i d D > iCS'84'M >105' O Nk ud. iCS- B4 53 a 'c e c\ ose.1 ad sew r e'A" g T033 en a.n A. 40.T k 3* ghs s ecca er- a n oces . - e.9 6.V $ g i O COMANCHE PEAK - UNIT 1 3/4 3-7

O - O - O _. TA8LE 't.3-2 % #57 l 8 g REAC10R TRIP SYSTEM INSTRUMENTATION RESPONSE TINES x 9 [ FUNCTIONAL UNIT RESPONSE TIME w - x 1. Manual Reactor Trip N.A. g 2. Power Range, Neutron Flux $ 0.5 conda

Z g 3. Power Range, Neutron Flux, High Positive Rate N.A.

4. Power Range, Neutron Flux, High Negative Rate 5 0.5 second*

5. Intermediate Range, deutron Flux N.A.

R 6. Source Rarxje, Neutron Flux $ 0.5 seconds

7. Overtemperature N-16 /

m

                                                                                                $ 7 seconds *#

8. Overpower N-16 5 7 seconds * .

9. Pressurizer Pressure--Low $ 2 seconds

10. Pressurizer Pressure-- igh ~< 2 seconds 553 Pressurizer. Water level--High

    " ~ .4,          11.                                                                        N.A.
    ~AE                                        /
    %3"
    -e                                /

l 3 /

Neutron /g m 1 detectors are exempt from response time testing. Response time of the neutron flux /gasuna signal portion of the channel ..iall be measured from detect 0r o:.:tput or input of first electronic component in channel.

                     # Response tir includes the thermal well response time.                                                             g m
                                                                                                                  .__                      .        a

i O O_ . O_ i i j TABLE 3.3-2 (Continued) REACTOR TRIP SYSTEM INSTRUMENTA1 ION RESPONSE TIMES ID8,0057 I E FUNCTIONA4. UNIT

,                         m                                                                                           RESPONSE TIME h             12. Reactor Coolant Flow--Low j                          c                 a. Single Loop (Above P-8)

3 b. Two Loops (Above P-7 and below P-8) $ [1] s ond l H $ [1 econd w 13.

!                                           Steam Generator Water Level--Low-Low                                     _ [2] seconds 1

14. Steam Generator Water 4

I Level-Low Coincident with Steam /Feedwater Flow Micmatch N.A.

15. Undervoltage - Reactor Coolant Pumps $ [1.5] seconds

{ 16. UMerfrequency - Reactor Coolant Pumps l 5 [0.6] second { 17. Turbine Trip j r.. Low Fluid Oil Pressure N.A. ! b. Turbine Stop Valve Closure N.A. Safety Injection Input frop4SF

18. N.A.

!l ;== 19. Reactor Trip System I erlocks M.A. i N51

                      *; y E
                      .                20 Reactor Trip Brpers                                                      M.A.

2s~ j g-u.

21. Automatic T p and Interlock Logic M.A.

a

                                       /                                                                                                        ll:0

p=

l M

O .

                                         .                         O                                              O              -

O TABLE 4.3-1 ~k n=~ REACTOR TRIP SYSTEN INSTRUNENTATION SURVEILLANCE REQUIRLhiNTS 5* g TRIP y ANALOG ACTUATING N00ES FOR

  ,                                                                   CHANNEL         DEVICE                 WilCN c                                       CHANNEL    CHANNEL           OPERATIONAL     OPERATIONAL ACTUATION  SURVEILLANCE-2        FUNCTIONAL UNIT                CHECK      CALIBRATION       TEST            TEST        LOGIC TEST IS REQUIRED
 ~          . Manual Reactor Trip          N.A. N.A.                N.A.           R(14)        N.A. 1, 2, 3*, 4*, 5*

2. Power Range, Neutron Flux 15

a. High Setpoint S D(2, 4), Q(17) N.A. N.A. 1, 2 N(3, 4),

Q(4, 6),

 ,                                                   R(4,  5) c 1             b. Low Setpoint            S        R(4)                S/U(1)         N.A.         N.A. 1,2 w                                                                          nS
 ,,       3. Power Range, Neutron Flux, M.A.       R(4)                Q(H)           M.A.         N.A. 1, 2 o             High Positive Rate 15

4. Power Range, Neutron Flux, N.A. R(4) Q(H) N.A. N.A. 1, 2 High Negative Rate C

5. Intermediate Range, S R(4, 5) S/U(1) N.A. N.A. I,2 Neutron Faux 6 b

6. Source Range, Neutron Flux 5 R(4, 13) S/U(1),Q(9,17) R(12) N.A. 2 , 3, 4, 5 15

7. Overtemperature N-16 5 D(2, 4) Q(H) N.A. N.A. 1, 2 N(3, 4)

Q(4, 6) R(4, 5) 15

8. Overpower N-16 S D(2, 4? Q(H) N.A. M.A. 1, 2 g P.(4, 5) y

9. Pressurizer Pressure--Low 5 R f5 Q(8, W ) N.A. N.A. I d W 15

10. Pressurizer Pressure--High 5 R Q(17) N.A. N.A. 1, 2

                                                           ~                                                   ~

G J L) C R* TABLE 4.3-1 (Continued) mg4 o 7, C O REACTOR TRIP SYSTEM INSTRUMENTATION SURVE!LLANCE REQUIREMENTS f 5* 5 TRIP $ ANALOG ACTUATING MODES FOR v CHANNEL DEVICE WHICH - S CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SUE!EILLANCE 7 FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED E 11. Pressurizer Water Level-- d S R Q( ) N.A. N.A. l

-4      High
-                                                                      15 d

12. Reactor Coolant flow--Low 5 R Q(17) N.A. N.A. 1 iS

13. Steam Generator Water Level-- S R Q(8, R) N.A. N.A. 1, 2 Low-Low 15 d

14. Undervoltage - Reactor Coolant N.A. R N.A. Q(10, It) N.A. I w Pumps D 15 d w 15. Underfrequency - Reactor N.A. R N.A. Q(it) N.A. I Coolant Pumps h

16. Turbine Trip d

a. Low fluid Oil Pressure N.A. R N.A. S/U(1, 10) N.A. I

' d

b. Turbine Stop Valve N.A. R N.A. S/U(1, 10) N.A. I Closure

17. Safety Injection Input from N.A. N.A. N.A. R N.A. 1, 2 ESFAS

18. Reactor Trip System Interlocks -

a. Intermediate Range Neutron Flux, P-6 N.A. R(4) R N.A. N.A. 2 O

W M

O O O [. 53

                                                                                                                                                  '" M J, TABLE 4.3-1 'I ntino-d)                                                                                    :s g g g 5..

REACTOR TRIP SYSTEM INSTRUMENTAsi2A SURVEILLANCE REQUIREMENTS g* E TRIP v ANALOG ACTUATING MODES FOR

9 CHANNFL DEVICE WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIREU E

G 18. Reactor Trip System Interlocks (Continued) w

b. Low Power Reactor Trips Block, P-7

1) Power Range Neutron Flux P-10 N.A. R(4) R N.A. N.A. 1, 2 R

2) Turbine First Stage Pressure P-13 N.A. R R N.A. M.A. 1 Y

U c. Power Range Neutron Flux, P-8 N.A. R(4) R N.A. N.A. I cl,# Power Range Neutron Flux, P-10 N.A. R(4) R N.A. N.A. 1, 2

19. Reactor Trip Breaker N.A. N.A. N.A. N(7, 11) N.A. 1, 2, 3", 4*, 5*

20. Automatic Trip and Interlock N.A. N.A. N.A. N.A. M(7) 1, 2, 3*, 4*, 5*

Logic ,

  -HeeeteMr4p-Bypass-Breaker-- N. A.--        - - N. A.  --- - - M.A.             M(15),R(16)                       N.A.       1, 2, 38 , 48 , 5'     _

10 : 0113 O N I 3=' . M [

TIIC512 AliACHMENT 6 FAGE 22 0F 10$ , TABLE 4.3-1 (Continued) y d TABLE NOTATIONS "When the Reactor Trip breakers are closed and the Control Rod Drive System is capable of rod withdrawal. D 0elow P-6 (Intermediate Range Neutron Flux Interlock) Setpoint. C Below P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint. d 9 ^ f AbovetheP-7(,atpower),setpoint. (1) If not performed in previous 31 days. (2) Comparison ef 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 cdlorimetric 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 2 or 1. (3) Single point comparison of incore to excore AXIAL FLUX DIFFERENCE - above 15% of RATED THERMAL POWER. Recalibrate if the absolute 103:011g difference is areater than or equal to 3%. For the purpose of these surveillancel"M"f requirementsfis defined as at least once per 31 doye. EF Pb. The provisions of Specification 4.0.4 are not applicsble for entry M., O MODE F or S. l f i 1 (4) Neutron and N-16 detectors may be excluded fros.i CHANNEL CALIBRATu (5) Detectorplateaucurvesshallbeohtainedandevaluated. For the Intermediate Range,a # Power Range Neutron Flux and N-16 channels the provisionsofSpecjification4.0.4arenotapplicableforentryinto MODE 1 or 2. gym yg* .$ %s AML FLM bif fdNCE Iti 0121 (6) Incore-ExcoreCalibration,fabovef6%ofRATEDTHERMALPOWER. For the purpose of these surveillance requi;rements "Q" is defined as at least once per 92g dey+. The provisions of Specification 4.0.4 are not applic-able f eniry t into MODE 1 or 2. l EFPO. 50 (7) Each train shall be tested at least every 62 days on a STAGGERED TES1 BASIS. (8) The surveillance frequency and/or modes specified for these cha1nels in Table 4.3-2 are more restrictive and therefore applicable. a (9) Quarterly surveillance in MODES 3 8 4*, and S shall also include verifica-tion that permissives P-6 and P-10 ar* in Geir required state for exist-ing plant conditions by observation of the permissive annunciator window. Quarterly surveillance shall include verification of the Boron Dilution AlarmSetpointoflessthanorequaltofanincreaseoftwicethecount rate within a 10-minute period /. T he 7cuaio% J Spcific h,en 4.o.4 a.cc mi aqyucab\t lov wy to 2A bor5 Eo uo s a. yco.tkor ty;p. 108:0123 COMANCHE PEAK - UNIT 1 3/4 3-13 ,

IIX-68512 AliACHn(Ni 6 PAGE 23 0F 105

TABLE 4.3-1 (Continued) . o Q TA8LE NOTATIONS (Continued) (10) Setpoint verification is not applicable. (11) The TRIP ACTUATING DEVICE OPERATIONAL TEST shall independently verify the OPERASILITY of the undervoltage and shunt trip attachfesnts of the Reactor Trip Breakers. (12) At least once per 18 months during shutdown, verify that on a simulated Boron Dilution Doubling test signal the normtl CVCS discharge valves close and the centrifugal charging pumps suction' valves from the RWST open. (13) With the high voltage setting varied b w ,mmended by the manufacturer, an initial discriminator bias curve s! . 4 de measured for each detector. Subsequent discriminator bias curves shu 1 be obtained, evaluated and compared to the initial curves. (14) The TRIP ACTUATING DEVICE OPERATIONAL TEST shall independently verify the OPERASILITY of the undervoltage'and shunt trip circuits for the "" "al Reactor Trip Function. The test shall also verify the OPERABILI " of the Bypass Breaker trip circuit (s). (15) Lecel-menuel-thunt-trip-pelee-to-pl4cing4eeakeMa-serv 4ce. (Or for p'ent that de net-actuate-the-shur,t-trip-attachment of the bype:5= breekers en a =anual reactor _ trip)L_ Remota manual undervaltag trin when breaker placed 4n-senice- g

         ' ,-ifr)-Atttomet4e-ttedervettege--teipr-(if) Each channel shall be tested at least every 92 days on a(staggered test?   ' ~~

15 dasii; - /'~~

                                                             '                         ,10 l: 0126
       ,                            At c. m g ter.x ,r.c.c l

f O COMANCHE PEAK - UNIT 1 3/4 3-14 ,

TXX-86512 ,

ATTACHMENT 6 PAGE 24 0F 105 + I I O l l l COMANCHE PEAK STEAM ELECTRIC STATION TECHNICAL SPECIFICATION 3/4.3.2 O O

TXX-08512 CPSES Technical Sp:cifications AllACHMENT6-NRC Draft 2 Markup PAGE 25 0F 105 Section 3/4.3.2 Change 10# Justification For Change N,) 0128 This Table is being relocated to the CPSES Technical 0904 Specification Improvement Progra.n. TV Electric believes the inclusion of this Table is unnecessary and the information would be more appropriately addressed in the CPSES Technical Specification Improvement Program. Relocation of this Table is consistent with the guidance provided in the NRC's Interim Policy Statement (52FR3788), February 6,1987, and the recommendations of the Westinghouse Owners Group MERITS Program. Priority is given to the relocation of this Table since the detailed information is not used by the Licensed Operator, but is purely used as a testing criteria. The information currently in this Table is more appropriately maintained in a document subject to TV Electric administrative control and 10CFR50.59 review under the CPSES Technical Specification Improvement Program. This change is similar to that Licensed at Shearon Harris, Seabrook and Vogtle. 0145 The deletion of the trip of all main feed pumps and loss (3 of offsite power that start the AFW pumps is based on the N / fc11owing. The requirement to maintain OPERABILITY of the I function to start the motor driven AFW pumps on a trip of all main feedwater pumps or a loss of offsite power for CPSES's design are anticipatory actions that are not assumed in the accident analysis. For a loss of feedwater accident in conjunction with a loss of offsite power condition, the steam generator low-low level is the assumed protective function that is used in the accident analysis to start the AFW pumps to supply water to the l steam generators. This assumed function is presently in the Technical Specification to ensure the accident analysis assumptions are available in the MODE that the accident is credible. 0146 Description of 6.9kv Safeguards Undervoltage System. 0156 0173 The purposes of the entire system is to ensure: 0176

1. The Class 1E bus is disconnected from non-1E buses; Q

v

       - IXX 885l2 .

ATTACHMENT 6 PAGE 26 0F 105 CPSES Te:hnical Specifications NRC Draft 2 Markup Section 3/4.3.2 r^'3 Change 10# Justification For Change V

2. Sufficient loads on the 1E bus are shed.to avoid subsequent Diesel Generator overloads when it connects to the bus and sequentially loads; and

3. The Diesel Generator starts and is ready for loading within the required time.

The CPSES system uses several different channels of Undervoltage protection to accomplish these tasks:

1. Ta divorce the 6.9kv IE bus from offsite power, the incoming feeder breakers receive trip signals from Undervoltage Relays sensing the associated incoming feeder - either the preferred offsite source (directly from the Startup Transformer, independent of Preferreri Transformer, and via non-Safeguards 6.9kv bus). Jecause Technical Saecifications provide for the use of either of t1ese offsite sources (3.8.1.1), both relay networks are needed to ensure bus disconnection.

2. Load shedding of the 6.9kv loads is accomplished by Bus Undervoltage Relays which sense the bus itself. Essentially all loads except 480v

\- transfonners are shed.

3. The signal to start the Diesel Generator also comes from the Bus Undervoltage Relay. Once started, the diesel comes up to speed and voltage and will connect to the bus provided both incoming offsite feeder breakers are open. In addition a backup start signal is provided from the Preferred Offsite Undervoltage Relays.

The Degraded, or 2nd level, Undervoltage protection for the 6.9kv bus uses separate Undervoltage Relays that sense the bus directly. Upon sensing a degraded condition, these relays trip both offsite feeder breakers, if shut, then the Bus Undervoltage Relays described above will function. Each of the Undervoltage channels consists of 2 sensing relays in a 2 out of 2 logic.

m.gg AllACHMENT6

                   ' PAGE D OF 105           CPSES Technical Specifications NRC Oraft 2 Markup
                                   . .               Section 3/4.3.2 n                   Change ID#          Justifica; ion For Change k.)

The proposed change to Table 3.3-3 adds the alternate offsite Undervoltage based on the need to ensure the feeder breaker is tripped to disconnect the IE system described above. It also adds the Degraded Voltage which is consistent with the original Staff Position (SER 8.2.4) to include these in Technical Specifications. This is also consistent with the Standard Technical Specifications The Preferred offsite and Bus Undervoltage have been re-written to be a functional unit (channels) rather than listing discrete components (Undervoltage Relay, timer) that had been done previcusly. The "channel" includes these previously listed components, as well as additional relays that were not listed in the previous version. Listing in terms of the function is consistent with the other items on the ESF table which don't specify each channel component. The channels to trip and Action Statement have been made consistent with this channel approach. The previous listing for Bus Undervoltage function to initiate the Solid State Safeguards System Sequencer (SSSS) is now included in Table 4.3-2 item lib, the Black Out Sequence portion of the 5555. The sequencer has (l (/ separate Undervoltage Relays dedicated to it, which reset and initiate the timing sequence following blackout. Tables 3.3-4, 3.3-5 and 4.3-2 formats require changing for consistency with the change in Table 3.3-3 justified above. In addition, Table 4.3-2 requires these changes:

1. Add Note (2) (TXX-6905) to the Trip Actuating device operational test, which takes exception to the setpoint verification normally required by the TA00T definition. The design of the Undervoltage Relays does not include provision to vary and measure the input voltages to obtain a trip setpoint. In order to do this type of test, test blocks have to be removed, modified to a test configuration and reinstalled with test equipment connected to the switchgear. Setpoint information can then be measured and the system restored to normal. During this time, the Undervoltage function is inoperable (bypassed), these test

IxX-88512 CPSES Technical Specifications AllACHNENT6 NRC Draft 2 Markup PAGE 28 0F 105 Section 3/4.3.2 Chanae ID# Justification For Chanae U blocks were designed for calibration activities normally done during extended outages. The Undervoltage channels are designed with test switches that allow exercising of the various relays (actuating devices) and the final components (breakers, loads, etc.). The design is consistent with the requirements of IEEE-338 and was sufficient to meet the CHANNEL FUNCTIONAL TEST requirement. To measure setpoints would be hazardous to the plant (by removing Undervoltage protection), to personnel and contrary to IEEE-338 wnich prohibits make-shift test set ups.

2. Add Note (5) (TXX-6905) which changes the TAD 0T test requirement from a monthly to a Cold Shutdown Test requirement. As described above, the system design allows exercising of the relays, but causes actuation of the final components. These components include the major 9.6kv Switchgear from offsite power (see description of the system) and the major 6.9kV loads, some of which are normally operating. For example, the Station Service Water and Component Cooling Water are normally in service as supporting equioment for both safety and non-safety related equipment. Testing bus f,)'

v Undervoltage Relays would shutdown tr,ase pumps; therefore significant interruptic:. c,i normally running equipment would occar or have to be shifted to the opposite trair. (c,g. HVAC, RCP Seal Cooling, Letdown Heat Exchanger, Boron Recycle, etc.), this major plant testing is undesirable to be run with the plant at power. The risk of major plant upset or trip due to equipment losses or voltage functions is nct warranted. Testing at Cold Shutdown adequately demonstrates the functioning of this equipment without undue risk of plant upset. 0156 See 10# 0146

IXX-8851;- l AllACHalNI 6 ' PAGE 29 0F 105 CPSES Technical Specifications NRC Draft 2 Markup Section 3/4.3.2 l Chanae 10# Justification For Chanae b(N 0173 See ID# 0146 0176 See 10# 0146 0571 Removal of Je requirement for manual initiation of AFW is 0572 based on CPSES design. There is no single initiation control function for the AFW System. This equirement l would have to be satisfied by manually starting each pump individually and verifying valve positions by the operators. This testing is very redundant since on a monthly basis the pumps are started from the control room to perform Inservice Testing of the pumps. Also these controls are used during plant startup, operations and shutdown which gives additional confidence of the systems operability. 0605 This change revises the applicability for the Preferred Offsite Source Undervoltage to require OPERABILITY only when the feeder breaker from Preferred Offsite Power is shut. The safety function of these undervoltage relays that is relied on is to trip the power feeder breaker in the event of a loss of offsite power. This trip isolates the IE 6.9 kv safety bus from the non-lE supply, to assure (] proper isolation load shedding (from separate Bus V undervoltage relays), and subsequent diesel generator loading. These relays additionally provide an anticipatory diesel start signal which is redundant to the start signal from the lE bus undervoltage relay. In order to allF placing the associated startup transformer out of servic. for maintenance or testing, the undervoltage relays and/or the diesel generar start signal must be defeated to avoid a start of the aiesel generator and running the engine unloaded. The present LC0 Action Statement will not allow the inoperability of all Preferred Offsite Undervoltage Relays, which can be accomplished by removal of the test block, disconnecting the relays from the bus. The alternatives would be to disable the diesel start signal (via jumpers or lifted leads) leaving the undervoltage relays OPERABLE, but tripped, when the bus is de-energized. This is less desirable due to use of jumpers and lifted leads. Note that the out-of-service of the offsite power supply is governed by the allowed outage time of Specification 3.8.1.1 which allows up to 72 hours in this condition provided both diesel generators and the other offsite circuit (Alternate Offsite Source) are CPERABLE. The diesel generator remains OPERABLE getting the start signal from the 6.9kv Bus Undervoltage Relays in the event of loss of off:ite power or loss of bus voltage.

IIX-88512 CPSES Technical Specifications ATTACHMENT 6 NRC Draft 2 Markup PAGE 30 0F 105

                .   ,              Section3/4.3.2 Chanae ID#           Justification For Chance 0904           See ID# 0128 0930           This change was never requested, and we suspect that this is a typographical error.

0947 The Containment Spray (CS) Pump has been added to the time response table 3.3-5 under the mitigating signal that produce Safety Injection. Upon receipt of any of thesa signals the CS pumps start and remain on recirculation upon receipt of High-3 signal the pump receives a redundant start signal and the discharge valves start to open while the recirculation valves close. The valves are tested under the ASME Section XI program to ensure their opening times are consistent with that assumed in the accident analysis. The pump is tested under the ASME Section XI program as well as time response tested for the four (4) signals that cause a pump start. O

TXX-08512 AfiACMENT 6 PAGE 31 Of 105

INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2 The Engineered Safety Features Actuation System (ESFAS) instrumentation channels and interlocks shown in Table 3.3-/ shall 2 be OPERABLE with their Trip Setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-f and with ESF RESPONSE TIMES as shown in Tab h 3.3-5. 3 MW hw

                                                                              \ wA va.We. .

APPLICABILITY: AsshowninTable3.3-/. 2 ID 1: 0128 ACTION: l

a. With an ESFAS Instrumentation or Interlock Trip Setpoint trip less conservative than the value shown in the Trip Setpoint column but more conservative than the value shown in the Allowable Value column of Table 3.3-f, adjust the Setpoint consistent with the Trip Setpoint value. .3

b. With an ESFAS Instrumentation or Interlock Trip Setpoint less conserva-tive than the value shown in the Allowable Value column of Table 3.3 , either:

1. Adjust the Setpoint consistent with the Trip Setpoint value of Table 3.3- N and determine within 12 hours that Equation 2.2-1 g'

V was satisfied for the affected channel, or

2. Declare the channel inoperable and apply the applicable ACTION statement requirements of Table 3.3-/2until the channel is restored to OPERABLE status with its 5etpoint adjusted consistent with the Trip Setpoint value.

Ecuation 2.2-1 Z + R + 5 1 TA Where: Z = The value from Column Z of Table 3.3- for the affected channel, R = The "as measured" value (in percent span) of rack error for the affected channel, S = Either the "as measured" value (in percent span) of the sensor error, or the value from Column S (Sensor Error) of Table 3.3-(3 for the affected channel, and TA = The value from Column TA (Total Allowance) of Table 3.3-4 for l the affected channel,

c. With an ESFAS instrumentation channel or interlock inoperable, take the ACTION shown in Table 3.3-4, 1

2 b COMANCHE PEAK - UNIT 1 3/4 3-15 l

IIX 88512

    .                     Ai!ACHMENT 6 PAGE 32 0F 105 INSTRUMENTATION-o (y   ,                 gRVEILLANCE REQUIREMENTS 4.3.2.1      Each ESFAS instrumentation channel and interlock and the automatic actuation logic and relays shall be demonstrated OPERABLE by performance of the ESFAS Instrumentation Surveillance Requirements specified in Table 4.3-2.

4.3.2.2 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function shall be demonstrated to be within the limit at least once per 18 months. Each test shall include at least one train such that both trains are tested at least once per 36 months and one channel per function such that all channels are tested at least cnce per N times 18 months where N is the total number of redundant channels in a specific ESFAS function as shown in the "Total No. of Channels" column of Table 3.3-J. 2 O' . O COMANCHE PEAK - UNIT 1 3/4 3-16

O O O - 2. TABLE 3.3-+ , _ . _ , n $55 g z ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION ggg - g5~ 9

                                                                                                             - v.

MINIMUM S A TOTAL NO. CHANNELS CHANNELS APPLICABLE <

y. FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

1. Safety Injection (ECCS, *

25. Reactor Trip, Feedwater

Isolation, Control Room

~ Emergency Recirculation, Emergency Diesel Generator Operation, Containment Vent Isolation, Station Service Water, Phase A Isolation, Auxiliary Feed-water-Motor Driven Pump, R Turbine Trip, Component

Cooling Water, Essential y Ventilation Systems, and y Containment Spray Pump.

17

a. Manual Initiation 2 1 2 1,2,3,4 4G-13

, b. Automatic Actuation 2 1 2 1,2,3,4 42 Logic and Actuation Relays Id

c. Containment 3 2 2 1,2,3 4ba Pressure--High-1 18 D

Ma

d. Pressurizer 4 2 3 1,2,3 Pressure--Low D *

e. Steam Line Pressure--Low 3/ Steam Line 2/ Steam Line 2/ Steam Line 1, 2, 3 In any Steam Line C "23 h

"4

2. TABLE 3.3-J (Continued)

                                                                                                                                    , =: =T ~

n g ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION g@$ z *EG j 9 gm MINIMUM

A TOTAL NO. CHANNELS CHANNELS APPLICABLE

   $   FUNCTIONAL UNIT                                  OF CHANNELS      TO TRIP        OPERABLE        MODES            ACTION

2. Containment Spray c 17 Z

a. Manual Initiation 2 pair 1 pair 2 pair 1, 2, 3, 4 -16

operated

   "                                                                     simultaneously                                     g

b. Automatic Actuation 2 1 2 1,2,3,4 12 Logic and Actuation Relays 15

            -              ' * * '     . cressure--          4                  2           3        1,2,3                 1+

High-3

3. Containment Isolation w

l ,', a. Phase "A" Isolation g7

1) Manual Initiation 2 1 2 1,2,3,4 Ifr .

s3

2) Automatic Actuation 2 1 2 1, 2, 3, 4 42-Logic and Actuation Relays

3) Safety Injection See Item 1. above for all Safety Injection initiating functions and requirements.

b. Phase "B" Isolation r7

1) Manual Initiation See Item 2a above. Phase "B" isola- 1,2,3.4 -1fr ~

tion is manually initiated when containment spray function is manually initiated. e Automatic Actuation 2 1 2 1,2,3,4 -12 2) Logic and Actuation Relays D Containment 4 2 3 1,2,3 N

3) M Pressure--liigh-3

O O O 2 TABLE 3.3-2 (Continued) h ENGINEEREO SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION N 9 MINIMUM A TOTAL NO. LGNNELS CHANNELS APPLICABLE R FUNCTIONAL UNIT OF CHANNELj TO TRIP OPERABLE MODES ACTION c c. Containment Vent - 3 3 3,, .,5;aj Isolation w - g 1) Manual Initiation See Item 2a. and 3.a.1 above. Contsinnent vent isolation is manually U~ F " t"j initiated when Phase "A" isolation function or containment spray s ' " i '*" ^ function is manually initiated.' 4 is ' N d " ' d"" fl, ?, 3, 4 15f "' '" " " P W

2) Automatic Actuation 2 1 2 1,2,3,4 46-Logic and Actuation m Relays D

w 3) Safety Injection See Item 1. above for all Safety Injection initiating functions and g requirements.

4. Steam Line Isolation

a. Manual Initiation U

                                             ; c ;;                            1) Individual Steam                         1/ steam line         1/ steam line 1/ operating    1,2 d,3 d  -24 NsZ                                   Line                                                                        steam line UAE                                                                                                                                   d    d  M S; ';

2) System 2 1 2 1,2 ,3 -20 M

1,2 d,3 d

                                             -e S           b. Automatic Actuation                                               2                       1            2                    19-                                  -

Logic and Actuation Relays d d Na

c. Containment Pressure-- 3 2 2 1, 2 , 3 -13 O High-2 y W

M

m 3 %

                                              ,                                                                                    \        .

8 2 TABLE 3.3-7 (Continued) c> - y ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION Mr MINIMUM A TOTAL NO. CHANNELS CHANNELS APPLICABLE R FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

4. Steam Line Isolation (Continued) k

-4

d. Steam Line Pressure--Low 3/ steam line 2/ steam line 2/ steam line 1, 2d , 3b,d a in any steam

1 line

, e. Steam Line Pressure - 3/ steam line 2/ steam line 2/ steam line 3c ,d 18] a y Negative Rate--High any steam line

5. Turbine Trip and Feedwater Isolation g

1 ? a. Automatic Actuation 2 1 2 1, 2 42-y Logic and Actuation ! g Relays ga

b. Steam Generator 3/sta. gen.I 2/sta. gen. 2/sta. gen. 1, 2 -la i Water Level-- in any stm.

I Hinh-High -(."- 14 ) gen.

c. Safety Irjection See Item 1 above for all safety injection initiating functions and

   , , _ ,                                              requirements.
,  --~                                                                                                                                          i l
   ] y { 6.        Auxiiiary FeedWater 4

2,,6_.6 a. .":nual-Initiatica -2 1 2 1, 2, 3 29 <--ID 3: 0572., zo i ae

a. .-b. Automatic Actuation Logic 2 1 2 1,2 3 49- -

and Actuation Relays

b. A Stm. Gen. Wat.:r Level--

Low-Low C

1) Start Motor- i8 Driven Pumps 4/sta. gen. 2/sta. gen. 3/sta. gen. 1, 2, 3 4?a in any oper- in each ating stm. gen. operating stm. gen.

O O O - 2 i TABLE 3.3-J (Continued) ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION i Y z Q MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE

      ;R  FU$fCTIONAL UNIT                      OF CHANNELS     TO TRIP         OPERABLE         MODES       ACTION
       . 6. Anxiliary Feedwater (Continued)                                                                                     ,

E 2) Start Turbine- 18 y Driven Pump 4/sta. gen. 2/sta. gen. 3/sta. gen. 1, 2, 3, 47,

      ~                                                          in any         in each 2 operating     operating stm. gen.       stm. gen.

d c -tb Safety Injection Start Motor-Driven Pumps See Item 1. above for all Safety Injection initiating functions and requirements. i m d. +. Loss-of-Of fsite Power

      }             Start Motor-Driven m            Pumps and Turbine-                                                                          i7

] g, Driven Pump 1/ train 1/ train 1/ train 1, 2, 3 %

)              -f. Trip-o f- All-Hein-                                                                                     l
                   -Feedwater-Pumps-ID t: 0145
-                  -Start-Mo tor--                                                                                          g I

Briven-Pumps 2/ pump-- - ---1/ pump 1/p 7 1, 2 E I

    $ $ $ 7. Automatic Initiation of ECCS

[$s Switchover to Containment k$5

    - o.        a. Automatic Actuation            2                 1             2         1,2,3,4 13
  ,                  Logic and Actuation Relays J

l2ll3 3:= M

3 r\ (3 J . V U 2 n TABLE 3.3-3'(Continued) c> I ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION M E MINIMUM y TOTAL NO. CHANNELS CHANNELS APPLICABLE y FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION c 7. Automatic Initiation of ECCS . 2 Switchover to Containment U Sump (Continued) 18

b. RWST Level--Low-Low 4 2 3 1,2,3,4 -l+

Coincident With[ Safety See Item 1. above for all Safety Injection initiating functions Injection and requirements.

8. Loss of Power (6.9 kV Safeguards System Undervoltage)

{ a. Preferred Offsite 2/bos 2/bos 1/bos if 2.0 , f 9 24- ids 0605

   ,                                                                                           Source Undervoltage e                                                                                                                                                                                               a g                                                                                       )--Undervel tage-Relay-2/ bus                      2/ bus        1/ bus         1, 2, 3, 4         24, 2-)----Diesel Start-T4mer 1/bu,                   1/ bus        1/ bus         1, 2, 3, 4         27 3
                                                                                         -3)         Scurce-Bke-Trip           1/ bus            1/ bus        1/ bus         1, 2, 3, 4         27 b.

Nffc7D y,745e 2-/ bos 2/bos I/ bos t,2,3,4 2.4[ C . +. BusUnde[ 2,/bos 2/bos 1/bos l3 1,3,4 M o.,

                                                                                                     ^t volta

d. D.F*Diese, y

Y6 h'ge m, A$ 2/ bos 2./ bo s 1/ bas 1 3 1, L 4 14

= = -e)---Undervol tage----2/ bus 2/bu, 1/ bus 1, 2, 3 ? 20*

Nk M *Y 24 a ID 1: 0146 ta g g b) Tircr 1/ bus 1/ bus 1/ bus 1, 2, 3, 4 % =; " g* 2) Initiat4on-of-- Solid-State-Safe-guards Syst-- c _ m_._ _ _y_..__. d

2) u.,eer= nage e/ bus 2/ bus 3/ bus 1, 2, 3, 4 17

                                                                                                              -Relay-                                                                               a                                     ,L:.
                                                                                                    -b)-----T imer             e/ bus            2/ bus        3/ bus         1, 2, 3 -i         17                                        .-

O O * (A} v i z I TABLE 3.3-1 (Continued) ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION 9 MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE ' A FUNCTIONAL UNIT OF CHANNELS TO 1 RIP OPERABLE MODES ACTION

                        . 9. Control Room Emergency                                                                                              .

c- Recirculation

                       ?.                                                                                                                 25

a. Manual Initiation 2 1 2 All "

25

b. Automatic Actuation 2 1 2 1,2,3 24 Logic and Actuation Relays

c. Safety Injection See Item 1. above for all Safety Injection initiating functions and requirements M

10. Engineered Safety Features Y Actuation System Interlocks 19 m

                       "         a.           Pressurizer Pressure,              3             2                 2        1,2,3            It            .

P-11 21

b. Reactor Trip, P-4 2 2 2 1,2,3 M

11. Solid State Safeguards

                     ;; = =      Sequencer (5555)
                     $; y Z                                                                                                                13 t: 3 y      a.           Safety Injection             1/ train       1/ train        1/ train        1, 2, 3, 4      -It g5~                      Sequence g *~                                                                                                                 2.*l             _

b. Black Out Sequence 1/ train 1/ train 1/ train 1, 2, 3, 4 -2fr C

M

1

TXX-88512 i ATTACHMENT 6 PAGE 40 0F 105 d " f TABLE 3.3-2 (Continued)

Y' m TABLE NOTATIONS a The provisions of Specification 3.0.4 are not applicable, b Trip function may be blocked in this MODE below the P-11 (Pressurizer Pressure Interlock) Setpoint. c Trip function automatically blocked above P-11 and may be unblocked below P-11 by blocking the Safety Injection on low steam line pressure. d Not applicable if each affected main steam isolation valve and its associatsJ upstream drain pot isolation valve per steam line is closed.

          '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 hour and maintained in the tripped condition with the exception that the channel may be taken out of the tripped condition for up to 2 hours to allow-testing of redundant channels. IJ)S 6RT A -> ACTION STATEMENTS 13 ACTION LE - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within pd 6 hours and in COLD SHUTDOWN within the following 30 hours; however, one channel may be bypassed for up to 2 hours for surveillance testing per Specifiestion 4.3.2.1, provided the other channel is OPERABLE. 4 ACTION 49 - 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 1 hour. 15 ACTION F& - 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 Channels OPERABLE requirement is met. One additional channel may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.2.1. 10> ACTION 16 - With less than the Minimum Channels OPERABLE requirement, opera- } tion may continue provided the containment pressure relief valves are closed within 4 hours and maintained closed. 17 ACTION M - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours or be in at least HOT STANDBY within the next 6 hours and in COLJ SHUTDOWN within the following O 30 hours. COMANCHE PEAK - UNIT 1 3/4 3-24 .

TXX-885!2 AliACHMENT 6 PAGE 41 OF 105 INSERT A

g. Not applicable if Preferred offsite Source Breaker is open.

O O O

IXX-88512 AliACMEWI 6 PAGE 42 0F 105 3 TABLE 3.3-7 (Continued) .i O} ( \ ACTION STATEMENTS (Continued) tb l ACTION 7 - With the number of OPERABLE channels one less than the Total I Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following ccnditions are satisfied:

a. The inoperable channel is placed in the tripped condition within 1 hour, and

b. The Minimum Channels OPERABLE requirement is met; however, one additional channel may be bypassed for up to 2 hours for surveillance testing of other channels per Specifica-tion 4.3.2.1.

6 ACTION -le - With less than the Minimum Number of Channels OPERABLE, within 1 hour determine by observation of the associated permissive annunciator window (s) that the interlock is in its required state for the existing plant condition, or apply Specification 3.O.3. lo ACTION 19 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within l 6 hours and in at least HOT SHUTDOWN within the following 6 hours; however, one channel may be bypassed for up to 2 hours O for surveillance testing par Specification 4.3.2.1 provided the other channel is OPERABLE. (/ 11 ACTION 20 - With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours or be in at least HOT STANDBY within 6 hours and in at least HOT SHUTDOWN witnin the following 6 hours. 22. ACTION 21 - With the number of OPERABLE channels one less than the Total l Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours or declare the associated valve inoper-able and take the ACTION required by Specification 3.7.1.5. 23 ACTION B& - With the number of OPERABLE channels one less than the Minimum l Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours; however, one channel may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.2.1 provided the other channel is OPERABLE. 24 ACTION 23 - With the number of OPERABLE channels one less than the Total l Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a. The inoperable channel is placed in the tripped condition within 1 hour, and O

COMANCHE PEAK - UNIT 1 3/4 3-25 ,

rAhc 3.1. ?_ ( tenb;n ce Al m-W50 ATTACMENT 6 PAGE 43 Of 105 A CTtod STAT (mCATS kC6^hlMeb

b. The Minimum Channs OPERABLE requirement is met, q g5 h ACTION M - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to GPERABLE status within 48 hours or initiate and maintain operation of the Control Room Emergency Recirculation System.

14 ACTION & - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable Channel to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in at least HOT SHUTDOWN within the following 6 hours. 27 ACTION f6 - With the number of OPERABLE Channels on one or more trains less than the Minimum Channels OPERABLE requirement, declare the diesel generator (s) associated with the affected train (s) inoperable and apply the appropriate ACTION for Specification

3. 8.1.1.

AGHCN 27 - With-less-than-the-Minimum-Channels-OPERABLEStaetup-and/or-- Powe r-Oper a t i on- may- p roc eed - p rovi ded- t he- t i meHfr-t he-aff ected channel-i s- bypa s s ed- a nd - ac t i o ns - a re- ta ke n- i mmed i a tely-to-- restu re-the-time r- te-OPERAB L E- s t atus . 10 l[ Old O O O O COMANCHE PEAK - UNIT 1 3/4 3-26 ,

O - . O - 3 TABLE 3.3-( 8 g ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS M E

                          ,                                                                          SENSOR g                                                      10TAL               ERROR                                  '

x FUNCTIONAL UNIT ALLOWANCE (TA) Z_ (5) TRIP SETPOINT ALLOWA8LE VALUE

1. Safety Injection (ECC Reactor Trip, '

q Feedwater Isolation, Control Room g Emergency Recirculation, Emergency Diesel Generator Operntion, Contain-ment Vent Isolation, Station Service Water, Phase A Isolation, Auxiliary feedwater-Motor Driven Pump, Turbine Trip, Component Cooling Water, Essential Ventilation Systems, and Containment Spray Pu . w D a. Manual Initiation i!. A. N.A. N.A. N.A. N.A. w h b. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A. and Actuation Relays .

c. Containment Pressure--High 1 2.5 G ?1 1. 5 5 3.35 psig 5 3.9 psig

d. Pressurizer Pressure--Low 16.1 14.41 1.5 > 1829 psig t 1823 psig ER% e. Steam Line Pressure--Low 17.3 14.81 1.5 1 605 psig* 1 586 psig" ENE

                       *5~  2. Containment Spray 3*

a. Manual Initiation N.A. N.A. N.A. N.A. N.A.

b. Automatic Actuation Logic M.A. N.A. N.A. N.A. N.A.

and Actuation Relays

c. Containment Pressure--High-3 2.5 0.71 1. 5 < 18.35 psig -< 18.9 psig 4::=3 ll2:3 Oc=

N /~'N L j i Q )j TABLE 3.3-A'(Continued) I E g ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS M g SENSOR TOTAL EER0R

y FUNCTIONAL UNIT ALLOWAf;CE (TA) Z (S) TRIP SETPUINT ALLOWABLE VALUE .

x { e 3. Containment Isolation c *

        %           a,   Phase "A" Isolation

1) Manual Initiation N.A. N.A. N.A. N.A. N.A.

?

2) Automatic Actuation Logic N.A. N.A. N.A. al. A. M.A.

and Actuation Relays 1

3) Safety Injection See Item 1. above for all Safe'.y Injection Trip Selpoints and Allowable Values, w

. A b. Phase "B" Isolauon w i E

1) Manual Initiation See Item 2.a above. Phase "B" isolation is manually initiated when containment spray function is manually initiated.

2) Automatic Actuation Logic N.A. N.A. N.A. N.A. -

N.A. and Actuation Relays

3) Containment Pressure-- 2.5 0.71 1.5 $ 18.35 psig 5 18.9 psig

   ,,_                        High-3 539 g=           c. Containment Vent Isolation i   " !E r

c. x =

_' ; 1) Manual Initiation See Items 3.a.1 and 2.a above. Containment Vent Isolation is - 3 manually initiated when Phase "A" isolation f.inction or [ containment spray function is manually initiarden ed..

2) Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A.

and Actuation Relays O

                                                                                                                                ,: 0

3) Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and 'e z Allowable Values. -

                                                                                                                                ..=-

O O-O 3 TABLE 3.3-X (Continued) E ENGINEEREO SAFETY FEATURES ACTUATION SYS1EM INSTRUMENTATION TRIP SETPOINTS E z n SENSOR 5 TOTAL ERROR FUNCTIONAL UNIT ALLOWANCE (TA) Z_ (S) TRIP SETPOINT ALLOWABLE VALUE . 7 4. Steam Line Isolation E a. Manual Initiation N.A. N.A. N.A. N.A. N.A. e b. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A. and Actuation Relays

c. Containment Pressure--High-2 2.5 0.71 1. 5 $6.35 psig $6.9 psig

d. Steam Line Pressure--Low 17.3 14.81 1.5 1605 psig* 1586 psig a

e. Steam Line Pressure - 8.0 0.5 0 $100 psi3** $111.6psif*

s

Negative Rate--High a O 8 Y S. Turbine Trip and Feedwater O! Isolation

a. Automatic Actuation logic M.A. N.A. N.A. N.A. N.A.

and Actuation Relays , , _ _ . , b. Steam Generator Water 7. 6 4.3 1. 5 $82.4% of $84.2% of narrow j $gM Level--High-High M narrow range range instrument j ggg instrument span. j ggn span. , 5 c. Safety Injection See Item 1 above for all Safety Injection setpoints and allowable - values. D

t:3 I:::=

O O - O TABLE 3.3-4 (Continued) 8 g ENGINEERED SAFETY FEATURES ACTUATION SYSTEN INSTRUMENTATION TRIP SETPOINTS M g SENSOR

     ,                                                 TOTAL                   ERROR g   FUNCTIONAL UNIT                               ALLOWANCE (TA).Z         (S)        TRIP SETPOINT    ALLOWABLE VALUE           '

n a

6. Auxiliary Feedwater E g'

     ~
     "         +.      Manual-Initiation               N:A.             N:A.      N A. N:A.             N:A:--

ID 3: 0571 e.-b: Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A. I and Actuation Relays b -c. Steam Generator Water 8.8 7.08 1.5 > 43.4% of > 42.1% of narrow Level--Low-Low narrow range range instrument instrument span. span. c.-d: Safety Injection - Start See Item 1. above for all Safety Injection Trip Setpoints and y Motor Driven Pumps Allowable Values. d -e. Loss-of-Of fsite Power N.A. N.A. N.A. N.A. N.A. I

                -f. Trip-of-All-Main-Feedwater-- N. A.--   ---- N: Ar          N:A. N A.             N Ar                     IDI 0145
                     -Pumps-                                                                                                             '

7. Automatic Initiation of ECCS R5$ Switchover to Containment Sump cN3 a. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A.

g; E G and Actuation Relays 5

b. RWST Level--Low-to.< [2.1] [0.71] [1.2] >

                                                                                           -[40.6%of        >    34.94%Lof span               .

Coincident With span ~ Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and Allowable Values. D m

O O O TABLE 3.3-4 (Continued) 8 g ENGINEERED SAFETY FEATURES ACTUATION SYSTEN INSTRUNENTATION TRIP SET >0lNTS M g SENSOR

  ,                                                   TOTAL                   ERROR g   FUNCTIONAL UNIT                                 ALLOWANCE (TA) Z         (S)         TRIP SEIPOINT  ALLOWABLE VALUE                                     '

n a g 8. Loss of Power (6.9 kV Safeguards q System Undervoltage) s

a. Preferred Offsite Source Undervoltage v4.A. 4,g, g , g e7] y

                                                                                                          ,, g y y 4}-Undervol tage- Relay- - - - - N . A.     --
                                                                 - --- N: A----N: A        >4800-V        : E92-V-4)-Diese3-5 tart-Timer               N,Ac-            N,A.      MrA. 70r75-6         20,825                    )--Source-Bkr.-Trip -Timer---N. A.                N:A.      N A. 70.5 3 70rSS-s-offs fe br(e Onderv. ticn.q,             g g,

b. Alf<ra=+c 3e g.g, [Wr3 Y E E. la fE ' 3 Y w c. 4h Bus Undervoltage o.4, 2 C\ater3 V A. A. A. A. a c tuter 3 V

d. begeded. Volta C A. h. n A. tL A . >(idu3 V 3 [ lafer 3 Y w 4)--Diesel-Sta in gig w

a)-Undervoltage-Relay N:A. N.A. N.A. 22100 Y 11992 Y

                      -b)--Timer                      N;A.             N A.      N:A.      10r75:         10.825s--

2) Initiation-of-Solid-State-Sefeguards-Systen-Sege:.ce 2 "; O Mg2 -a)-Undervoltage-Relay MrA H.A. N.A. 24800-V 24592-V-agg b) Timer N;A. N: A:- N. A.- 10.5 s 10.55 5

% "; ~ ge 9. Control Room Emergency Recirculation y

                                                                                                                                   -til

a. Manual Initiation N.A. N.A. N.A. N.A. N.A. j

b. Automatic Actuation Logic and N.A. M.A. N.A. N.A. N.A. ..

Actuation Relays

c. Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and Allowable Values.

2S TABLE 3.3-4 (Continued) 8 g ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS M z SENSOR TOTAL ERROR FUNCTIONAL UNIT ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE

     .            10. Engineered Safety Features

c Actuation System Interlocks
1971 s9 Go

a. Pressurizer Pressure, P-11 N.A. N.A. N.A. < [1^,05] psig < [1^^,0] psig

b. Reactor Trip, P-4 N.A. N.A. N.A. N.A. N.A.

11. Solid State Safeguards Sequencer N.A. N.A. N.A. N.A. N.A.

(5555) R. M 6 M7 ONY ae~ 5* C W 3:=

TXX-88512 ATTACHMENT 6 FACE 50 0F 105

                ~    '

TABLE 3.3-4 (Continued) TABLE NOTATIONS

Time constants utilized in the lead-leg controller for Steam Line Pressure-Low are i > 50 seconds and T2 5 5 seconds. CHANNEL CALIBRATION shall ensure that these time constants are adjusted to these values.

   **The time constant utilized in the rate-lag controller for Steam Line Pressure-Negative Rate-High is greater than or equal to 50 seconds. CHANNEL CALIBRATION shall ensura that this time constant is adjusted to this value.

4 O O COMANCHE PEAK - UNIT 1 3/4 3-33

TXX-88512 AllACHMENT6 IDI 0904 7 s, TABLE 3.3-5 3D1{'L (U) , ENGINEERED SAFETY FEATURES RESPONSE TIMES WITIATION SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS

1. Manual Initiation

a. Safety Injection (ECCS) N.A.

b. Containment Spray (Phase "B" Isolation N.A.

nd Containment Ventilation Isolation)

c. Ph se "A" Isolation (Containment N.A.

Ven lation Isolation)

d. Steam ine Isolation N.A.

e. Feedwate Isolation (SI) N.A.

f. Auxiliary F'edwater (SI) N.A.

g. Station Serv e Water (SI) N.A.

h. Component Cooli g Water (SI) N.A.

i. Control Room Emer ency Recirculation (SI) N.A.

j. Reactor Trip N.A.

k. Emergency Diesel Gene tor Operation N.A.

1. Essential Ventilation S tems (SI) N.A.

m. Turbine Trip N.A.

2. Containment Pressure--High-1

a. Safety Injection (LCCS) 1 27(1,5(a))/ 12 (4'0D}

b. Reactor Trip i2

c. Feedwater Isolation < 6.5

d. Phase "A" Isolation i 17(2)/27(1)

e. Containment Ventilation Isolation N.A.

f. Auxiliary Feedwater < 60

g. Station Service Water N. .

h. Component Cooling Water N.A.

i. Essential Ventilation Systems N.A.

j. Emergency Diesel Generator Operation i 12

k. Turbine Trip N.A.

1. Control Room Emergency Recirculation N.A.

m. Cukimsd 5pecq Pup (7) j7/Q \ \

N i s COMANCHE PEAK - UNIT 1 3/4 3-34 -

TIX-88512 AITACHl 2 6 PAGE 52 0F 105

                       .        .                                                IDI 0904 TABLE 3.3-5 (Continued) n)
 \
  '                                  ENGINEERED SAFETY FEATURES RESPONSE TIMES hkTIATINGSIGNALANDFUNCTION                                  RESPONSE TIME IN SECON05

3. ressurizer Pressure--Low

a. Safety Injection (ECCS) 1 27(1,Sa)/12(4,5b)

b. ' actor Trip 12

c. Fee ater Isolation <7

d. Phase 'A" Isolation 17(2)/27(1)

e. Containm t Ventilation Isolation 1 5(6)

f. Auxiliary edwater 1 60

g. Station Servi Water N.A.

h. Compenent Coolin Water N.A.

i. Essential Ventilat n Systems N.A.

j. ' Emergency Diesel Gene tor / Operation i 12

k. Turbine Trip N.A.

1. Control Room Emergency Rec culation N.A.

l m, Coat Lahnsd Sp,y pr. p('/) 17/;7 ng y

4. Steam Line Pressure--Low l

a. Safety Injection (ECCS) 1 22(3,5b)/12(4,5b)

b. Reactor Trip 12

c. Feedwater Isolation < 6.5

d. Phase "A" Isolation 17(2)/27(1)

e. Containment Ventilation Isolation .A.

f. Auxiliary Feedwater 10

g. Station Service Water N.A.

h. Component Cooling Water N.A.

O COMANCHE PEAK - UNIT 1 3/4 3-35

III 88512 i A!!ACHMfMi 6 PAGE S3 0F 105 Om TABLE 3.3-5(Continued}

y) ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING 3IGNAL AND FUNCTION RESPONSE TIME IN SECONOS

4. Steam Line Pressure--Low (Continued)

                 .      Essential Ventilation Systems                            N.A.

J. Emergency Diesel Generator Operation 5 12

k. Turbine Trip N.A.

1. C trol Room Emergency Recirculation N.A.

7 s. Ste Line Isolation 6.5 8

              -,n.      Cc4i my sp,w, iLp (7)                                   //p7                   10t 090

5. Containment essure--High-3 I

a. Containee Spray Fump 1 22(2)j;;(1) 7)

G

b. Phase "B" Is lation N.A.

6. Containment Pressure -High-2 Steam Line Isolat n 1 6.5

7. Steam Line Pressure - Neg ive Rate-High Steam Line Isolation 17 O 8. Steam Generator Water Level-Hig -High

a. Turbine Trip N.A.

b. Feedwater Isolation i 11

9. Steam Generator Water Level-Low-Low

a. Motor-Oriven Auxiliary l Feedwater Pumps s 60

b. Turbine-Oriven Auxiliary Feedwater Pump

                                                                           \     5 60 10,    Loss-of-Of fsite Power Auxiliary Feedwater
                                                                                \N.A.

l 11. Trip of All Main Feedwater Pumps i

A l l -Met.or-Drivc Auxiliary Feedwater Pumps N.A.: 12. RWST Level--Low-Low Coincident with Safety Injection l; a. Automatic Initiation of ECCS Switchover to Containment Sump 130 1O COMANCHE PEAK - UNIT 1 3/4 3-36 .

l

11X 88512

ATTACMENT 6 k p PAGE'540F105 k'IDIJ704

                   .     .                                                                         2 TABLE 3.3-5 (Continued)

ENGIt!EERE0 SAF2TY FEATURES RESPONSE TIMES NG SIGNAL AND FUNCTION RESPONSE TIME IN SECON05 9

13. Loss of er (6./ kV Safeguards System Undervolta

a. Preferred Offsit Source Undervoltage < Ql

b. Bus Undervoltage 4 < 0]

01 < l I O COMANCHE PEAK - UNIT 1 3/4 3-37 .

    .      IXX-88512 ATTACHMENT 6                                                                                           I I

PAGE 55 0F 105

_ . . IDI0904 f- TABLE 3.3-5 (Continued)

TABLE NOTATIONS 4 Diesel generator starting and sequence loading delays included. (2) esel generator starting delay g included. Offsite power ava'lable. l 1 (3) Diesel nerator starting delay included. RHR pumps ng included. ] (4) Diesel gene tor starting and sequence loading delays not included. l RHR pumps not neluded. l (5) Response Time Lia1( includes opening of injection path valves. Following additional \ time is allowed for completion,of the transfer ) of the pump suction fr' the VCT to the RWST. a) 10 seconds b) 15 seconds (6) Includes containment pressure reMef line isolation only. (7) fespos e. L,wt / mie is mp dkevy k u ,w p b u d e c lo s < c>s if. ik 2k l '. ,w v. shoas e ,s c3 d e. s epucs de Icy cmd

                                                                                    / cad 'A3 Os                     ek Ar9er           a.mb r es /s o        e n c. L d e s       a s e l c) e,ie, a fe,-

6 A Fd A ssg c e ,9

t O COMANCHE PEAK - UNIT 1 3/4 3-38

O O
O -

TABLE 4.3-2 S ENGINEERED SAFETY FE/.TURES ACTUATION SYSTEM INSTRUMENTATION j SURVE]LLANCE REQUIREMENTS

                           %o                                                                                                                          TRIP o                                                                             ANALOG                                                                                              '

ACTUATING MODES 9

CHANNEL DEVICE MASTER SLAVE FOR WHICH CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY
RELAY SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST E E@s IS REQUIRED Z 1.SafetyInjectiondreactor l

                          ..                       Trip, Feedwater Isolation, Control Roon Emergency Recirculation, Emergency Diesel Generator Opera-tion, Containment Vent Isolation, Station Service Water, Phase A Isolation, Auxiliary Feedwater-Motor R*

Driven Pump, Turbine Trip, Component Cooling Water, Y Essential Ventilation M Systems, gnd Containment SprayPumpf

a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. .N.A. 1, 2, 3, 4

b. Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q 1, 2, 3, 4 Logic and Actuation Re1ays

c. Containment Pressure- S R M N.A. N.A. N.A. N.A. 1, 2, 3 High-1 kh d. Pressurizer Pressure S R M N.A. N.A. N.A. N.A. 1, 2, 3 D-g3$ Low N,.

    }f~                                           e. Steam Lic.e              S            R                  M                                   N.A.                        N.A.       N.A. N.A. 1, 2, 3     Q S:                                               Pressuie-Low

2. Containment Spray

a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1, 2, 3, 4

l.

                                                  .                           O   .                                               -

O - l TABLE 4.3-2 (Continued) 8 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION f SURVEILLANCE REQUIREMENIS 5 x m TRIP j m ANALOG ACTUATING MODES ' i 9

CHANNEL DEVICE MASTER SLAVE FOR WHICH

! CHANNEL CHANNEL CHANNEL

         '                                                                    OPERATIONAL OPERATIONAL ACTUATION     RELAY   RELAY SURVEILLANCE

! FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST t C - IS REQUlWED

5

b. Automatic Actuation M.A. N.A. N.A. N.A. M(1) M(1) 1, 2, 3, 4 Q

Logic and Actuation Relays , c. Containment Pressure- S R M N.A. N.A. N.A. N.A. 1, 2, 3 { High-3

3. Containment Isolation ids 0930

a. Phase "A" Isolation T

n.g. n . A. R R.R. W . R. ~ N 4. -

1) Manual Initiation dn. item-2ra-abover-Phn "B" isolet-ion-sanually-initieted t.cr. contebr - 1,2,3,4 Y ment- spray- func ti on-is-menue Hy-i ni t i e tM .

i $ 1 2) Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q 1,2,3.4 Logic and Actuation Relays

 .                       3) Safety Injectinn       See Item 1. above for all Safety Injection Surveillance Requirements.
           ;:' O O EE2       b. Phase "B"   Isolation UAE S 'i "        1) Manual Initiation      See Item 2.a. above. Phase "B" isolation is manually initiated                   1, 2, 3, 4 g*                                      when containment spray function is manually initiated.

2) Automatic Actuation N.A. N.A. N.A. N.A. N(1) M(1) Q 1, 2, 3, 4 l Logic and Actuation Relays

3) Containment 5 R M N.A. N.A. N.A. N.A. 1, 2. 3 O Pressure-High-3

=a

O n - O TABLE 4.3-2 (Continued) n ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUNENTATION SURVEILLANCE REQUIREMENTS ! 9 TRIP i ANALOG ACTUATING MODES i A CHANNEL DEVICE MASTER SLAVE FOR WHICH E CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE

                         . FUNCTIONAL UNIT                          CHECK       CALIBRATION TEST             TEST         LOGIC TEST TEST    TEST   IS REQUIRED E         c. Containment Vent Isolation U                         .

r 1) Manual Initiation See Item 3.a.1 and 2.a above. Containment vent isolation is manually 1,2,3,4 initiated when Phase "A" isolation function or containment spray function is manually initiated.

2) Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q 1, 2, 3, 4 Logic and Actuation Relays

3) Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.

4. Steam Line Isolation w

a. Manuai Initiation N.A. N.A. N.A. R N.A. N.4. N.A. 1, 2, 3

b. Automatic Actuation M.A. N.A N.A N.A. M(1) M(1) Q 1, 2, 3 Logic and Actuation Relays

c. Containment Pressure- S R M N.A. N.A. N.A. N.A. 1, 2, 3 High-2

d. Steam Line 5 R M N.A. N.A. N.A. N.A. 1, 2, 3 Pressure-Low

e. Steam Line Pressure- S R M N.A. N.A. N.A. N.A. 3 ,

~~

Negative Rate-High 353 5. Turbine Trip and Feedwater

                     '." 2y- $
                     ,             Isolation c.; E U       a. Automatic Actuation               N.A.       'N.A.           N.A.          N.A.         N(1)        N(1)   Q      1, @

ge Logic and Actuation -y Relays p ,

p ,Q (7 _

TABLE 4.3-2 (Continued) n ENGINEERED SAFETY FEl.TURES ACTUA110N SYSTEM INSTRUMENTATION h SURVEli!ANCE r.t0UIREMENTS 9 TRIP A ANALOG ACTUATING MODES ' s CilANNEL DEVICE MASTER SLAVE FOR MlICH

      ,            CilANNEL                      CilANNEL CilANNEL          OPERATIONAL OPERATIONAL ACTUATION              RELAY  RELAY SURVEIL. LANCE FUNCTIONAL UNIT                        CilECK    CALIBRATION TEST                   TEST             LOGIC TEST TEST    TEST    IS REQUIRED Z    S. Turbine Trip and Feedwater Isolation (Continued)

b. Steam Generator Water S R H N.A. N.A. N.A. N.A. 1, 2 Level-liigh-liigh

c. Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.

6. Auxiliary Feedwater ID : 0571

     $         .ar-Manual-Initiation             N.A.       N . A .- - - - N. A. -         -- R -- - - - H , A .       - N, A. EA.      IdA l y    a...b- Automatic Actuation             N.A.      N.A              N.A.             N.A.             M(1)        M(1)   Q        1, 2, 3 g            Logic and Actuation Relays b .- c . Steam Generator Water         S         R                H                N.A.             N.A.        N.A    N.A      1, 2, 3 Level-Low-Low C..d,    Safety Injection              See Item 1. above for all Safety Injection Surveillance Requirements.

d.+ Loss-of-0f f site Power N.A. R N.A. M(3, 4) N.A. N.A. N.A 1,2,3

              - Trip- nf-A11- Mai n Feed -- - N. A.-    - N . A .- -- -- N. A. - - - ---- R -- - - - - N . A .          N A. N.A     17-  2    O
                                                                                                                                                             ~

> ID 1: 0904 E59 7. Automatic Initiation of g@S ECCS Switchover to Con-gEG tainment Sump 5 a. Automatic Actuation N.A. N.A. N.A. H.A. M(1) P'l) Q 1,2,3,4 Logic and Actuation Relays

O O . O - 1 1 TABLE 4.3-2 (Continued)

!   8                                      ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUNENTATION f                                                               SURVEILLANCE REQUIREMENTS M

E TRIP 1 o ANALOG ACTUATING MODES l C

CHANNEL DEVICE MASTER SLAVE FOR WHICH CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED fi

b. RWST Level-Low-Low 5 A M N.A. N.A. N.A. N.A 1, 2, 3, 4 w

Coincident With l Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements. i

8. Loss of Power (6.9 kV i

Safeguards System Undervoltage)

l
w a. Preferred Offsite ID s: 0173 i

i Source Undervoltage t4 . R. K Q.A.

    ,                                                                                           (3,5)                N.A.          n . g. ro.q      g , 7 , i, , q 1
                   -1)-Undervoltage w
                        -Relay                    M.A.       R                 N:A.              M                    N A.          N.A. N.A. 1,-2,-3,                     -3)-Diesel-Startr-
                         -T-imer                  N A.       R -- -            N:A.              M      ------ --- N . A .- - - N . A . M.A. 1, 2, 3, 4
                   -3)-Source-Bke-Teip-g p geroajcimeo g3.-E E rc Udv 4//.N.A.tj 4 R

g. N: A:-- r& A. M (_3,5) N.- A. 9.A. MrA. g.A-M-A. 172;---3 ;-- Bus Undervoltage M.A. t ' 2. , 3,

il

            '.-b.                                    e.g . R                  g.g.

g '$ ) g ,9 , g,q, g ,q ' L b-1-)r"A*d Y* R"3e Diesel-Start: N4. R. u.0 t 3 .5) a.4- a p. . u.6 'td,2', Ag >,i4

    ,__                 -a)-Undervol tage-E ': 0                   -Reiay               N A .-- R - - - -- -- -- N . A .               N - - --- --- - - - N . A . -     M: A.-  N.A. 174, 3, 4 -
    "' M a SiE ga~
                        -5)-T imer              - N : A .- -- R - - -- -- N . A :-- -- - - - - - H - - -- - - - N .-A .            N.-A. N-A. 1.-2, 3, 4 -

O

                                                                                                                                                    ~.x5 n==

M

O O TABLE 4.3-2 (Continued) n y ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION z SURVEILLANCE REQUIREMENTS 9 TRIP ! 7, ANALOG ACTUATING MODES ' 3E CHANNEL DEVICE MASTER SLAVE FOR WHICH

. CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE i c- FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED 4

2 Z Mnitiation-of-

     ~                         SoHd- State-Safe -                                                                                                  In : 01M i                               gu;rds-System Seq u r.ce-l

a) L'n&rvoltage-Relay N:A.- R- N:A. M MrA. N:A. MrA. 1, 2, 3, 1

                              -b)-Timee            N.A.      R                          N,A.        M            N.A.        N.A. M.A. 1, 2, 3, A -

J y

j 9. Control Room Emergency j ';" Recirculation J j-

a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. All

b. Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q 1, 2, 3 i'

      .                    Logic and Actuation Relays

c. Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.

10. Engineered Safety features Actuation

System Interlocks khh a. Pressurizer N.A. R M N.A. N.A. N.A. N.A. 1. 2, 3 eS3 Pressure, P-Il 43U 5 C

=a

TABLE 4.3-2 (Continued) n o ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION h SURVEILLANCE REQUIREMENTS 9 m l TRIP l g ANALOG ACTUATING MODES g CHANNEL DEVICE MASTER SLAVE F05t WHICH

          ,                                                       CHANNEL                    CHANNEL CHANNEL                                                             OPERATIONAL OPERATIONAL ACTUATION                                             RELAY  RELAY SURVEILL94CE c                                                  FUNCTIONAL UNIT                   CHECK                                                 CALIBRATION TEST                   TEST                                      LOGIC TEST TEST               TEST   IS REQUIRED h                                                      b. Reactor Trip, P-4        N.A.                                                  N.A.                  N.A.         R                                        N.A.                    N.A. N.A. 1, 2, 3

11. Solid State Safeguards Sequencer (5555)

a. Safety Injection N.A. R N.A. N.A. M(1,2) N.A. N.A. 1,2,3,4 Sequence

        ,                                                      b. Blackout Sequence        N.A.                                                  R                     N.A.         N.A.                                     M(1,2)                  N.A. N.A. 1, 2, 3, 4 w

l l

 ;:' O O ID 8: 0176 EE2
 }{"g                                                                                         TABLE NOTATION
    *                                                     (1) Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS.                                                                                                                             O (2) Performed by Solid State Safeguards Sequencer Automatic Test.                                                                                                                                               .

(3) Setpoint verification is not applicable. (4) Actuation of final devices not included. (S) whenever %e pant h in Coi.b mluT N rJ fo r 72 hooc3 or en o r e a.n A. d kka so<oe Nue tes4' as hs nd bee.n pr$ o < med, 6 7 <e o '.so s Al a9

IXX-08512 AllACHMEHi6 PACE 63 0F 105 O COMANCHE PEAV, STEAM ELECTRIC STATION TECHNICAL. SPECIFICATION 3/4.3.3 AND 3/4.3.4 O O

     #885U                          CPSES Technical Specifications A11At m 6                             NRC Draft 2 Markup PAGE 64 0F 105 Section 3/4.3.3 Change 10# Justification For Change hV 0178 0182 Delete criticality monitors and associated notes, Action, etc. from Table 3.3-4, Radiation Monitoring. The 0569          requirement for criticality accident monitors are contained in 10CFR70.24 for special nuclear material storage. This does not require underwater monitoring when special nuclear material is handled or stored beneath water shielding, which is the case for the CPSES spent fuel pools. Furthermore, Regulatory Guide 8.12 Revision 1, January 1981 indicates that where geometric spacing of special nuclear material (fuel) is such to preclude the possibility of criticality, alarms are not required and that it is appropriate to request exemption from the provisions of 10CFR70.24.

The design of the CPSES fuel storage is described in FSAR 9.1.1 and 9.1.2. Assessment of criticality potential is addressed in FSAR 4.3 2.6. Technical Specification 5.6 also reiterates the fuel storage design that maintains subcritical configurations under worst case postulated conditions. For spent fuel storage Keff is maintained less then .95 even with unborated water and

                    'for new fuel storage Keff is maintained less than .98 even assuming aqueous foam moderation. The fuel storage is seismically qualified and protected from heavy loads. For these reasciis, criticality in the fuel storage areas is not credible. Since the fuel storage design precludes a criticality event, the criticality
   ^

monitors are not relied upon to perform any safety function and do i not meet the criteria of 10CFR50.36 for items to be included in the plant Technical Specifications. The criticality monitors are described in FSAR 12.3 and Table 12.3-8. TV Electric considers this sufficient controls for the monitors and thus incorporation of these monitors into the CPSES Technical Specifications is inappropriate. This change is similar to that Licensed at South Texas, Vogtle, and Seabrook 0179 Revises Applicability to exclude Mode 5 and 6 except 0568 during Core Alternations or movement of irradiated fuel within the containment. (Add Note #1) During Mode 5 or 6 Operations, the event of concern with potential for Radioactive release is the Fuel Handling Accident. In Mode 5 or 6 when Core Alternations or movement of irradiated fuel is not in progress, this is not a credible accident. This is clearly evidenced by the requirements for the containment as a fission product barrier. In Modes 1 to

4. Containment Integrity is required ensuring a low leakage, pressure tight barrier to the environment. During Core Alternations or movement of irradiated fuel, Specification 3/4.9.4 controls the allowable status of containment penetrations, including requirements associated with Containment Ventilation Isolation from the Gaseous Radioactivity Monitor. At other times during Modes 5 and 6, there is no restriction on containment penetration status, it makes no sense to require isolation functions of the gaseous monitor to be OPERABLE if the equipment hatch and airlocks are open as is allowed. In addition, CPSES

-O takes no credit for the isolation functions of this monitor at any time, but feels it prudent to include during Core Alternations and movement of irradiated fuel.

In48512 CPSES Technical Specificatune ATTAC MENT 6 NRC Draft 2 Markup PAGE 65 0F 105 Section3/4.3.3 Chanae ID# Justification For Chance O This change will allow an appropriate period of downtime for maintenance, repair, or modification of the monitors without impacting the ability to purge containment during outages for habitability considerations. The purge path is monitored by the primary plant ventilation monitors in the plant stack, required by Specification 3.3.3.10 at all times. This change is similar to that Licensed at Farley, Catawba and Summer. 0189 The Limiting Conditions for Operations and Surveillance 0190 Requirements were rewritten to qualitatively specify the 0466 components which require transfer and/or control circuits to maintain independence from the Control Room so that safe shutdown can be achieved in the event of a Control Room fire. The transfer switches and Control Room circuits have been removed from Table 3.3-7, Remote Shutdown. This is based on such a listing would be extensive and would not provide any useful information to the operators. The inclusion of this information is unnecessary and would be more appropriately maintained in a document subject to administrative control and 10CFR50.59 review process. Administrative control of this information is consistent with the guidance provided in the NRC's Interim Policy Statement (52FR3788), February 6, 1987, and the recommendations of the hq Westinghouse Owners Group MERITS Program. This change is similar to that Licensed at Shearon Harris and Vogtle. 0197 The Instrument portion of Table 3.3-8, Accident Monitor, is revised to be consistent with specific CPSES plant design and to incorporate specific action statements for each instrument. These action statements have time allowances appropriate for the various degrees of redundancy and significance of the instrument channel to post-accident assessment. The instruments listed in the STS are, as labeled, for "Illustration only." The instruments as now listed specifically correspond to the instrumentation for CPSES which are classified as Type A variables. The CPSES Type A variables are listed in Table 7.5-2 in the FSAR. The CPSES Type A variables provide the primary information required to permit the Control Room Operating Staff to:

1. Perform the diagnosis specified in the applicable CPSES Emergency Response Guidelines;

2. Take specified pre-planned manually controlled actions, for which no automatic control is provided, that are required for safety systems to accomplish their safety function to recover from the Design Basis Accident Event (verification of actuation of safety systems are not Type A variables at CPSES); and

3. Reach and maintain a safe shutdown condition.

Tu 88512 CPSES Technical Specifications AliACHMENT6 l

NRC Oraft 2 Markup ! PAGE 66 Of 105 Section 3/4.3.3 Change 10# "Justification for Change , All Type A variables at CPSES are "Key" variables (provide the most direct measure of the information required) and all are category 1. 0213 Delete reference to Table 4.11-2. This reference is deleted since Table 4.11-2 has been relocated to the Radioactive Effluent and Environmental Monitoring Manual. 0216 This change specifies that Hydrogen Recombiner H2 and 02 monitors shall be calibrated with test gases reccannended by the vendor, rather than l' 4and 4'< of the appropriate gas. Hydrogen and oxygen monitoring of the Waste Gas Holdup System is provided as part of the Hydrogen Recombiner, monitoring both the inlet and outlet. Operation of the system, described in FSAR Section 11.3, consists of the constant recirculation of the on-service holdup tank by the waste-gas compressor, processed by the recombiner between the compressor and holdup tank. Waste gases are admitted to the compressor suction and treated by the hydrogen recombiner, prior to reaching the holdup tank The recombiner is-desioned to remove up to 6's hydrogen in a sinhle pass, requiring approximately 3% oxygen to be added to the inlet. Up to 9% hydrogen can be accommodated on the inlet and still ensure that the outlet remains less than 3% (maximum 3% oxygen added). The major source of hydrogen to the system is from the continuous O bleed established from the Volume Control Tank. Under no 'nal operation of the system, this hydrogen is removed by the recombiner prior to recirculation to the waste gas holdup ink. The most likely source of oxygen to the system is the oxy n that is intentionally added at the recombiner inlet which shou' ' be totally used within the recombiner. Based on the operation described above, the hydrogen and c gen instrument ranges are selected based on the anticipated concentration of gas in order to achieve the best accuracy 1d resolution. This results in the following ranges and recou.;nded the gases used to calibrate the primary range of interest which is underlined in the Table below. Vendor Recommended Inst Range (s) Span Gas Inlet Hydrogen 0-10% 8% inlet Oxygen 0 - 5'4 , 0-10'4 4's 0-25% Outlet Hydrogen 0-0.4%, 0-2% 1500 ppm 0-10'4 Outlet Oxygen 0-100 ppm, X_1, 1 75 ppm X10, x100 O

IXX88512 CPSES Technical Specifications ATTACHMENT 6 NRC Rev. 2 Markup PAst 67 0F 105 Section 3/4.3.3 Chanae 10# Justification For Chance '. VO In addition to the recomended span gases, a zero gas of essentially pure Nitrogen is used. These zero and span gases are consistent with standard industry practice of using "zero" gases in the range of 0-25% (usually 0-10%) and span gases in the range of 75-100% of the range of interest. Because of these varying requirements, the present specification mandating 1% and 4% gases is not warrarted since this does not achieve optimum resolution for the instruments. Requiring these adds an extra burden and cost to the calibration and isn't justified. The vendor's (Westinghouse /Teledyne) recomendations are the optimum for these instruments. 0466 See Change ID# 0189 0566 Remove Particulate Radioactivity as a functional unit for Containment Ventilation Isolation. Moniuring and Sampling of the Containment / Containment Penetration Pathway and the automatic control features for Containment Ventilation Isolation is prescribed by the Standard Review Plan (SRP) Section 11.5 Table 1. Pursuant to the SRP, only a single gaseous radioactivity channel is required to have a control feature i.e. Ventilation Isolstion. This proposed change is consistent with the SRP and is as submitted and accepted in the CPSES Final Draft Technical Specifications (1984). The Particulate Channel that causes Containment Ventilation Isolation is required to be Operable pursuant to item la of Table 3.3 6 (RCS Leakage Detection) subject to Limiting Conditions for Operations and Action Statements appropriate for that monitor. In addition this monitor also contains an Iodine Channel in addition to the gaseous and particulate, all of which will be normally available to cause Containment Ventilation Isolation during the extremely limited periods of time during which ventilation valves will be open. All three of these monitors -Particulate, Iodine and Gas (PIG) - share a comon electronics package, including microprocessor and trip function outputs. Additionally, CPSES takes no credit in any analysis for Isolation from these radioactivity monitors. This change is similar to that previously licensed at Farley, Catawba and Sumer. 0567 This changes the minimum channels Operable from 2 to 1. The CPSES design only includes one channel of this function. CPSES does not rely on or take credit for the isolation functions of this monitor. For the purposes of effluent release monitoring, all releases from the containment are directed through the Primary Plant Ventilation and monitored by the Gaseous Effluent Monitoring Instrumentation specified in Specification 3.3.3.8 (Table 3.3-10). These monitors are microprocessor based with several internal s self-checking features. For most failure conditions, the monitor goes into a circuit failure condition which outputs an alarm and trip signals thus failing safe.

l Txx 88512 CPSES Technical Specifications AIIA N I ' NRC Rev. 2 Markup PAGE 68 M 105 Section 3/4.3.3 Char,ce ID# Justi'fication For Chanae 0568 See Change ID# 0179 0569 See Change ID# 0178 0C01 The Action Statement for radiation monitors has been deleted since there are no longer any radiation monitors designated as type A variables. 0602 Deleted from Action Statement b. the reference to vent-high range noble gas monitor and steam relief high range radiation monitor since these monitors are no longer used in the accident monitoring Table 3.3-8. 0934 This Technical Specification is being relocated to the CPSES Technical Specification Improvement Program. TU Electric believes the inclusion of this Specification is unnecessary and the information would be more appropriately addressed in the CPSES Technical Specification Improvement Program. Relocation of this Specification is consistent with the guidance provided in the NRC's Interim Policy Statement (52FR3788), February 6, 1987, and the recommendations of the Westinghouse Owners Group MERITS Program. Priority is given to the relocation of this Specification since the detailed information is not used by the Licensed Operator, and requires no immediate action from the Licensed Operator if the (G) Action Statement is applied. The information currently in this Specification is more appropriately maintained in a document subject to TV Electric administrative control and 10CFR50.59 review under the CPSES Technical Specification Improvement Program. This change is similar to that Licensed at South Texas Seabrook and Vogtle. 0935 Added radioactivity limit for the lower limit of detection to make it consistent with Action Statement 34 (marked-up). 0936 This Technical Specification is being relocated to the CPSES Technical Specification Improvement Program. TV Electric believes the inclusion of this Specification is unnecessary and the information would be more appropriately addressed in the CPSES Technical Specification Improvement Program. Relocation of this Specification is consistent with the guidance provided in the NRC's Interim Policy Statement (52FR3788), February 6, 1987, and the recommendations of the Westinghouse Owners Group MERITS Program. Priority is given to the relocation of this Specification since the detailed information is not used by the Licensed Operator, and requires no immediate action from the Licensed Operator if the Action Statement is applied. The information currently in this T Specification is more appropriately maintained in a document u subject to TV Electric administrative control and 10CFR50.59 review under the CPSES Technical Specification Improvement Program

111-88512 AffACHM(Ni6 PAGE 69 0F 105 INSTRUMENTATION , V 3/4.3.3 MONITORING INSTRUMENTATION RADIATION MONITORING FOR PLANT GPERATIONS LIMITING CONDITION FOR OPERATION 3.3.3.1 The radiation monitoring instrumentation channels for plant operations shown in Table 3.3-59shall be OPERABLE with their Alarm / Trip Setpoints within the specified limits. APPLICABILITY: As shown in Table 3.3-f.# ACTION:

a. With a radiation monitoring channel Alarm / Trip Setpo,jnt for plant operations exceeding the value shown in Table 3.3-f,Tadjust the Setpoint to within the limit within 4 hours or declare the channel ,

inoperable. <

b. With one or more radiation monitoring channels fpr plant operations inoperable,taketheACTIONshowninTable3.3-f.

c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicaole, b(

SURVEILLANCE REQUIREMENTS 4.3.3.1 Each radiation monitoring instrumentation channel for plant operations shall be demonstrated OPERABLE:

a. At least once per 12 hours by performance of a CHANNEL CHECK,

b. At least once per 18 months by performance of a CHANNEL CALIBRATION,

c. At least once per 31 days by performance of a DIGITAL CHANNEL OPERATIONAL TEST.

O COMANCHE PEAK - UNIT 1 3/4 3-46 ,

s . k TABLE 3.3-g i , RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS 5 x MINIMUM A CHANNELS CHANNELS APPLICABLE ALARM / TRIP ' y FUNCTIONAL UNIT TO TRIP / ALARM OPERABLE MODES SETPOINT ACTION h 1. RCS Leakage Detection z U a. Particulate Radioactivity N.A. I 1, 2, 3, 4 N.A. -}1- 30 ~ b. Gaseous Radioactivity N.A. I 1,2,3,4 N.A. -al- 30

2. Containa.2nt Ventilation Isolation i

2. P::-ticulate-Radieaet4*ity 1 2 ^11 2S ID I: 0566 e

                -b. Gaseous Radioactivity                1                 14-           -AH

28
1,2,3,4 /, #
3. F=1 S* ers;;e Pee! Are?' ID I: 0567 w ID I: 0568 w

                                                                                           **                                       I y               M,41-icality-Radiation-Level             1                   2                          1 15 r.4/h    30     ID 1: 0569 I            -

3.4r Control Room E=r;;ency Recirse!atice

a. Air Intake-Radiation Level 1/ intake 2/ intake All d[Later] 29 l pCi/m1 1 N

'" n eL 2: Mi 3 45C Es

D I

IIX 08512 ATTACHM(Ni6 PAGE 71 0' 105 Y - n TABLE 3.3-B'(Continued) (V) TABLE NOTATIONS Must satisfy Specification 3.11.2.1 requirements, i

       **                                                                                               10 8: 0178 W it tt-f uel-i ft-t he-f uel-s t o rage- pool- a rea s-o e f uel- buih i ng--.

no us

     ***     "c:t ::ti:fy Sp::ific:ti;r 3.11 2.1 require.ents.# ~                      'n M l

gy bw,rig;;s C. ORE GtTC2. Ar so45 o < mooeme.nt o$ iredid el Eve \ ID I: 0179 M c o n ta ' n m e n +- ACTION STATEMENTS ACTION 28 - With the number of OPERABLE channels less than the Minimum Channels OFERABLE requirement, operation may continue provided the containment ventilation valves are maintained closed. The containment pressure relief valves may only be opened in com-pliance with Specification 3.6.1.7 and 3.3.3..d. 8 ACTION 29 - With the number of OPERA 8LE channels one less than the Minimum Channels OPERABLE requirements, within 1 hour secure the Coatrol Room makeup air supply fan from the affected intake or initiate and maintain operation of the Control Room Emergency Air Cleanup System in emergency recirculation. ACTION 30 - L'ith is:: than th: "i 4 r Ch:ene!: OPEP^.BLE requi-: :nt. epere-p* tica ::y ::ntinu -fer up te 30 d:y: previded ea esprepaiete pertebi; continuou: =0nitor with th: :::: ^1: = 50t?01^t it gd previded in the fuel :terage pcci er;a.

ter; th: in per:b' monitor: t: OPEPABLE etstu: "ithia 30 day: er zu pend :!'

eper;ti;n: inv;1ving fusi acv;;;nt in the fuel :torege p001 ID1:0182

re::.

30 ACTION 41 - With the number of OPERABLE channels less than the Minimum l Channels OPERABLS requirement, comply with the ACTION re. quire-ments of Specification 3.4.6.1. t COMANCHE PEAK - UNIT 1 3/4 3-48 ,

111-88512 I - ATTACMENT 6 FAGE 12 OF 105 i i ( STRUMENTATION Un , MOV LE INCORE OETECTORS D C U to' o's' LIMITI CONDITION FOR OPERATION 3.3.3.2 The vable Incore Detection System shall be OPERABLE with:

a. At le st 75% of the detector thimbles,

b. A minim of two detector thimbles per core quadrant, and

c. Sufficient ovable detectors, drive, and readout equipment to map these thimb s.

APPLICABILITY: When the vable Incore Detection System is used for:

a. Recalibration of he Excore Neutron Flux Detection System, or

b. Monitoring the QUA0 NT POWER TILT RATIO, or N

c. Measurement of Fg,Fq ) and F xy, ACTION:

With the Movable Incore Detection Syste inoperable, do not use the system for the above applicable monitoring or calib tion functions. The provisions of Specifications 3.0.3 and 3.0.4 are not ap icable. SURVEILLANCE REQUIREMENTS 4.3.3.2 The Movable Incore Detection System shall be .monstrated OPERABLE within 24 hours prior to use by normalizing each detect output when required for:

a. Recalibration of the Excore Neutron Flux Detectio System, or

b. Monitoring the QUADRANT POWER TILT RATIO, or N

c. Measurement of F ,iH, Fq(Z) and F,y,

                                                                                      \
                                                                                        \

O ' COMANCHE PEAK - UNIT 1 3/4 3-49 . t

111 88512 ATTACHMENI 6 PAGE 73 0F 105

INSTRUMENTATION Q(% SEISMIC INSTRUMENTATION LI ITING CONDITION FOR OPERATION 1.3.3.3 The seismic monitoring instrumentation shown in Table 3.3-7 shall be OPERABLE. APPLICABILIT

At all times.

ACTION:

a. With one r more of the above required seismic monitoring instruments inoperabl for more than 30 days, prepare and submit a Special Report to t Commission pursuant to Specification 6.9.2 within the next 10 days utlining the cause of the malfunction and the plans for restoring e instrument (s) to OPERABLE status.

b. The provisions o Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.3.1 Each of the above required s.ssmic monitoring instruments shall be demonstrated OPERABLE by the performance f the CHANNEL CHECK, CHANNEL CALI-BRATION, and ANALOG CHANNEL OPERATIONAL TE at the frequencies shown in Table 4.3-4. 4.3.3.3.2 Each of the above required seismic m itoring instruments which are accessible during power operations and which is tuated during a seismic event greater than or equal to 0.01g shall be restored t OPERABLE status within 24 hours and a CHANNEL CALIBRATION performed within 0 days following the seismic event. Data shall be retrieved from actuate instruments and analyzed to determine the magnitude of the vibratory ground mot . A Special Report

  -     shall be prepared and submitted to the Commission pursu                              to Specifica-tion 6.9.2 within 14 days describing the magnitude, frequ cy spectrum, and resultant effect upon facility features important to safety Each of the above seismic monito?ing instruments which it actu ted during a seismic event greater than or equal to 0.01g but is not accessible during power operation shall be restored to OPERABLE status and a CHANNEL CAlfRRATION per-formed the next time the plant enters MODE 5 or below. A supp1tme'ntal report shall then be prepared and submitted to the Commission within 14 days pursuant to Specification 6.9.2 describing the additional data from these instruments.
                                                                                                           '\

COMANCHE PEAK - UNIT 1 3/4 3-50

111 99512 ATTACHMENT 6 PAGE 74 of 105 DRAFT TABLE 3.3-7 SEISMIC MONITORING INSTRUMENTATION gg gg4 MINIMUM INSTRUMENTS INSTR ENTS AND SENSOR LOCATIONS. OPERABLE i

1. Tri ial Time-History Accelerographs

a. As elerometer-Fuel Building i

b. Ar.ce rometer-Containment I

c. Acceler eter-Electrical Manhole 1

d. Seismic Tr ger-Fuel Building i

e. Recorder Unit SHA-3 1

f. Playback Unit, P-1 1

2. Triaxial Peak Accelerog phs

a. Pressurizer Lifting T nion 1

b. Reactor Coolant Piping 1 10 V( c. Component Cooling Water Hea Exchanger 1

3. Triaxial Seismic Switch Fuel Building la 4 Triaxial Response-Spectrum Recorders

a. Fuel Building i

b. Reactor Bldg. Internal Structure 1

c. Safeguards Building -

1

5. Response Spectrum Annunciator la 4

COMANCHE PEAK - UNIT 1 3/4 3-51 .

F 1I1 88512 . AllACHMEN!6 - ' FAGE 75 0F 105

y TABLE 4.3-3 (, 7 to: 09%

SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS ANALOG

CHANNEL CHANNEL CHANNCL CPERATIONAL INSTR ENTS AND SENSOR LOCATIONS CHECK CALIBRATIOP TEST

1. Tria ial Time-History Accelerographs

a. A eierometer-Fuel Building M R SA

b. Acre romete r-Containment M R SA

c. Accelero eter-Electrical Manhole M R SA

d. Seismic Tr ger-Fuel Building M R SA

e. Recorder Unit, SMA-3 M R SA

f. Playback Unit, S -1 M R SA

2. Triaxial Peak Accelerograph

a. Pressurizer Lifting Tr ion N.A. R N.A.

b. Reactor Coolant Piping N.A. R N.A.

c. Component Cooling Water Heat N.A. R N.A.

Exchanger

3. Triaxial Seismic Switch Fuel Building ** R SA

4. Triaxial Response-Spectrum Recorders

a. Fuel Building N.A. R N.A.

b. Reactor Bldg. Internal Structure N.A. R N.A.

c. Safeguards Building N.A. R N.A.

5. Response Spectrum Annunciator ** M R SA

N N

                                                                                       \
       *Setpoint verification is not applicable.
      **With control room indication.                                                      \s O

COMANCHE PEAK - UNIT 1 3/4 3-52 -

IXI-0512 1 2

         .                                   AliACHMENT 6                        l PAGE 76 0F 105                      l
                                       ,                                                                                      i.

INSTRUMENTATION METEOR E 0 LOGICAL INSTRUMENTATION

                                           ,L    ITING CONDITION FOR OPERATION 3.3.3.        The meteorological monitoring instrumentation channels shown in Table 3.3-8 sh 11 be OPERABLE.

APPLICABILI Y: At all times. ACTION:

a. With on or more required meteorological monitoring channels inoperable for more han 7 days, prepare and submit a Special Report to the Comission ursuant to Specification 6.9.2 within the next 10 days outlining t cause of the malfunction and the plans for restoring the channel (s to OPERABLE status.

b. The provisions o Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.4 Each of the above meteorological onitoring instrumentation channels shall be demonstrated OPERABLE:

a. At least once per 24 hours by perfo ance of a CHANNEL CHECK, and i b. At least once per 184 days by performa e of a CHANNEL CALIBRATION.

l l l l 1 l l COMANCHE PEAK - UNIT 1 3/4 3-53 - l l

TIX-88512 ATTACHM(Ni6 PAGE 77 0F 105

                          .       .                            RE.0CATEt"o=

p TABLE 3.3-8 d \. METEOROLOGICAL MONITORING INSTRUMENTATION MINIMUM INST MENT LOCATION OPERABLE

1. W SPEED 1 of 3

a. -SY-4117 Nominal Elev. 60 m.

b. X-S -4118 Nominal Elev. 10 m.

c. X-SY-4 8* Nominal Elev. 10 m.

2. WINO DIRECTION 1 of 3

a. X-ZY-4115 Nominal Elev. 60 m.

b. X-ZY-4116 Nominal Elev. 10 m.

c. X-ZY-4106* Nominal Elev. 10 m.

3. AIR TEMPERATURE - AT 1 of 2

a. X-TY-4119 (ominal Elev. 60 m and

       ,                                              Nominal Elev. 10 m.

b. X-TY-4120 Nom al Elev. 60 m. and Nomin Elev. 10 m.

                                                                           \
                                                                            \
                                                                              \\
                                                                                \ -
             "Mounted on backup tower.

i

O COMANCHE PEAK - UNIT 1 3/4 3-54 i

IXXC512

     .          AliACHMENT 6 PAGE 78 0F 105 Q

INSTRUMENTATION y IL REMOTE SHUTDOWN {YSTE" INSTRUMENTATION LIMITING CONDITION FOR OPERATION 4.I g I 3.3.3 5 Thefemote monitoring instrumen' shutdown Oy;ts tr=:f:r : wit:M;, pw;r, contreh-ee h 0 89 tation channels shown in Table 3.3-J. shall be OPERABLE. l

           .Z'45eTT~ /} ->                                                                            M h 0190 APPLICABILITY: MODES 1, 2, and 3.                                                          I ACTIO_N,:

a. With the number of OPERABLE remote shutdown monitoring channels less than the Minimum Channels OPERABLE as required by Table 3.3-g,7 restore the inoperable channel (s) to OPERABLE status within 7 days, or be in HOT SHUTOOWN within the next 12 hours,

b. With the number of OPERABLE remote shutdown monitoring channels less than the Total Number of Channels as required by Table 3.3-y,7 within 60 days restore the inoperable channel (s) to OPERABLE status or, pursuant to Specification 6.9.2, submit a Special Report that defines the corrective action to be taken.

c. With one or more Remote Shutdown Sy;t; transfer switches, power, or control circuits inoperable, restore the inoperable switch (s)/

circuit (s) to OPERABLE status within 7 days, or be in HOT STANDBY within the next 12 hours,

d. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3. 1 Each remote shutdown monitoring instrumentation channel shall be l demonstrated OPERABLE

a. At least once per 31 oays by performance of a channel check, and

b. At least once per 18 months by performance of a channel calibration."

4 1 4.3.3.5.2 Each Remote Shutdown-Sy:t:: transfer switch, power and control circuit shall 3 be demonstrated OPERABLE at least once per 18 months by verifying its capability to perform its intended function (s), l t e. p red by 5 u'.h d.m 3. 4.3 '/ . A , f "Neutron detectors may be excluded from channel calibration, l COMANCHE PEAK - UNIT 1 3/4 3-55 ,

INSERT A TXX 88512 ATTACHMINI 6 FAGE 79 Of 105 _ 3.3.3.4.2 The remote shutdown transft switches and controls of system I] components required for 1) reactivity control, 2) RCS pressure control 3) decay heat removal 4) RCS inventory control and 5) support systems required for the above functions shall be OPERABLE. O l i i 1 e

TXX-8 dst 2 ATTACHMENT 6 PAGE 80 0F 105 U TABLE 3.3-( 7 " h I REMOTE SHUTOOWN -GMTEM MONITORING INSTRUMENTATION TOTAL NO. MINIMUM READOUT OF CHANNELS INSTRUMENT LOCATION C,tANNELS OPERABLE

Neutron Flux Monitors HSP 2 1

2. Wide Range RCS Temp. - T HSP 1/ Loop 1/ Loop c

3. Wide Range RCS Temp. - T HSP 1/ Loop 1/ Loop h ,

4. Pressurizer Pressure HSP 1 1

5. Pressurizer Level HSP 2 1

6. Steam Generator Pressure HSP 1/SG 1/SG

7. Steam Generator Level HSP 1/SG 1/SG

8. Auxiliary Feedwater Flow HSP 2/SG 1/SG Rate to Steam Generator

9. Candensate Storage Tank Level HSP 2 1

10. Charging Pump to CVCS HSP 1 1 Charging and RCP Seals -

Flow Indication

                                                                          -SWI-Tfn--
           --TRANSFER-SVITCHES [I14ustrat4enal-only-}                      LOCATION b(             1. Auxiliary feedseter Centrci
             --2 . Safe Shutdown Equi pent Power

a. AuwH4ery-Feedwater----

b. Chargfot-

                     -c. PressMzer-Heetees-J. s'inii

3. CVCS Makeup Flow Cent +ol-l . Ofe:e1 Generator Centrol

    .           5. Elesteical-Olstributien Sy: tem-Cont-col--

l SWITC" Its 0190

           -CONTROL CIRCUITS [Illustr&tienel only]                         LOCATICH
      .        1. 6xili:rj Feefaeter ricw

2. oressuriis He:ters--

3. CVCS "ekaup Flow l

            -- 4 . Diesel Geneeatoe-
           ' 5.       E!ectricaLaistr-ibutter Syste>-

l fq - HSP = Hot Shutdown Panel l Q SG = Steam Generator l COMANCHE PEAK - UNIT 1 3/4 3-56 . i I

  '. TXI-88512 ATTACHMENT 6 PAGE 81 0F l05

,. INSTRUMENTATION ! /T U ACCIDENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.f'5'The accident monitoring instrumentation channels shown in Table 3.3-yf 8 shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3. ACTION: bf

a. With the number of OPERABL ascident monitoring 4astrumentation channels less than the -Teh4- Number of Channels shown in Table 3.3-J6,3 restore the inoperable channel (s) to OP_ERABLE gg status within 7 days, or be in at least HOT STAN00Y 4ithin the next % hours. nd 4a at least w3T SHL'TOOWu "ithia the f0110=ing S hoursb l2.

b. With the number of OPERABLE accident -enitoringn i :tr=entatien channels except the unit vent =h4 9h-r-ange noble gas-aonitor, :nd -the ID) C602 ste : relief-high r ng; r: diction monit e less than the Minimum l Channels OPERABLE requirements cf Table 3.3-MT,8 restore the inoper- i able channel (s) to OPERABLE status within 48 hours or be in -' '-+

[,} b HOT STANOSY within the next S heur: :nd *^ st 10 st HOT SHUTDOWN within the foll: wing $ hours. Oe 4 IL With the number of OPERABLE channels for the unit vent-high ra n as monitor, or the steam relief-high range radia monitor or the c ment atmosphere-hich range radiatio Itor, or the reactor coolant tion level initor les an required by the Minimum Channels OPERA uiremen nitiate an alternate method 101 0601 of monitoring the appropriate er(s), within 72 hours, and either restore the inop e channel s OPERABLE status within 7 days or prepare ' submit a Special Report . Commission, pur-suant to S ication 6.9.2, within 14 days that prov actions take use of the inoperability, and the plans and schedule toring the channels to OPERABLE status. C. A The provisions of Specification 3.0.4 are not applicable. l (+ COMANCHE PEAK - UNIT 1 3/4 3-57 ,

IXX-88512 AllACHMENT6 PAGE820F105 INSTRUMENTATION l d ACCIDENT MONITORING INSTRUMENTATION J( SURVEILLANCE REQUIREMENTS

              .6 4.3.3.pl Each accident monitoring instrumentation channel shall be demonstrated OPERABLE:

a. At least once per 31 days by performance of a CHANNEL CHECK, and

b. At least once per 18 months by performance of a CHANNEL CALIBRATION.\ 10l'060L 1

    ^etm**in= erit Area Radiation (High Range) CHANNEL CALIBRATION may conh' an-electronic cali                             nel pt incloding Uie detector, for range decades abovo 10 R/h and

c..; evint CrMtHutinn check of the detector below 10 2/A-wittT Tn installed or portable gamma source. -

COMANCHE PEAK - UNIT 1 3/4 3-58 ,

                                              --w    ..-e a p----,

IT v' %) ' TABLE 3.3-10'8 O ACCIDENT MONITORING INSTRUMENTATION E z g -TOTAt.R Eco sRob MINIMUM m NO. OF CHANNELS y INSTRUMENT CHANNELS OPERABLE 2-

"K

1. Conta inment-Peessure-(Wide-Ha nge ) 2 1

[ c ID 5: 0197 z t -2. Containment Pressure (Narrow Range) 2 1 , a e z. .-3. 2 Reactor Coolant Outlet Temperature - TH0T (Wide Range) 1 3.4. Reactor Coolant Inlet Temperature - TCOLD (Wide Range) 2 1

4. -9. Reactor Coolant Pressure - Wide Range 2 1 S. -fr. *ressurizer Water Level 2 1 t"

4 6. ,7. Sten: Cer.erater Water Leve! - Mide R ng^ 2nd Auxiliary 2 t/ steam generator 1/ steam generator feedwater Flow (Secondary Coolant Availability) 7.-8. Steam Generator Water Level - Narrow Range 2.1/ steam generator 1/ steam generator 8.-9. Containment Water Level (Wide Range) 2 1 9.14. Core Exit Temperature (Thermocouples) 4/ core quadrant 2/ core quadrant

,2   -

si3 9 11. Steam-Relief-Vent-Nobic Cas Nr. iter 5 N . ^. . ' BEB S 4em1;ng LArea :4ces g; E C n n Cente:.A ..t Rad!2tien ("!gh R2nge)- 2 1

-+

8 t3--Reactor-vessel-Water Level 2 1 M W 7:=

.]

7 _-__- ______. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - - _ _ -

                                                               -                           f%

G ~/

8 TABLE 3.3-10'(Continued) n ACCIDENT MONITORING INSTRUMENTATION Y z 9 TOTAL MINIMUM NO. OF CHANNELS A INSTRUMENT CHANNELS OPERABLE 2-x

           .       - 14 . "entrorrflux (Seuice Range)                                                2                        1                        .

C j'i -15. Neutron-F-lux-(-Intermediate Range) 2 1 a

81 4 &. Condensate Storage Tank Level 2 (2 Sen: crc /Channe!) 1 (1 Senter/Channe!)_

i?. Ma;n s ka m l ;we ?resore ( Shm Generc~hr Teessore) Stean-Line Pi essu. c 1/5te wWe 2/:tec= genereter I/ Shm Une

                  --17 .                                                                                                      1/;tec= genereter ids 0197 13.4 8-      Refueling Water Storage Tank Watee Level                                  2                        1                           !I

43. Reactor--Coolant-Syste= Sebeool-ing-Maegin Meniter 2 -I u

N

20. Plant-Vent-Stack-HobHe Ces ".caitar; w

g a. Intermediate-Haage '" ' h )(inh omnna _ u n, _ M 7 I sus ~ - G.

v m

e ; TXX 08512 AITACHMENT6 PAGE 85 0F 105 . INSTRUMENTATION CHLORINE DETECTION SYSTEMS , LIMITING CONDITION FOR OPERATION lo 3.3.3.T Two independent Chlorine Detection Systems for each fresh air intake, with their Alarm / Trip Setpoints adjusted to actuate at a chlorine concentra-tion of less than or equal to 5 ppm, shall be OPERABLE. APPLICABILITY: All MODES, if chlorine gas is stored on site in quantities greater than 20 lbs. ACTION:

a. With one Chlorine Detection System at a fresh air intake inoperable, restore the inoperable system to OPERABLE status within 7 days or within the next 6 hours isolate the affected fresh air intake and

                         ' comply with the provisions of Specification 3.7.7 and either (1) operate the Control Room HVAC System from the unaffected fresh air intake or (2) initiate and maintain operation of the Control Room HVAC System in the isolation mode of operation.

b. With both Chlorine Detection Systems at a freah air intake

    N                   inoperable, within 1 hour isolate the affected fresh air intake and comply with the provisions of Specification 3.7.7 and either (1) operate the Control Room HVAC System from the unaffected fresh air intake or (2) initiate and maintain operation of the Control Room HVAC System in the isolation mode of operation.

c. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.I Each Chlorine Detection System shall be demonstrated OPERABLE:

a. At least once per 12 hours by performance of a CHANNEL CHECK, i

b. At least or.ce per 31 days by verifying alana and trip relay actuation when each channel is tested using installed test circuitry, and

! c. At least once per 18 months by performance of a CHANNEL CALIBRATION, l l COMANCHE PEAK - UNIT 1 3/4 3-61 . l

i

  .      TXX-88512 AllACHMENT6 PAGE 86 0F 105                                                                 m INSTRUMENTATION o/'    yL.SE-PART DETECTION SYSTEM
                                                                             '}

LIMITING CONDITION FOR OPERATION

3.3.3.8 T\Looso-PartDetectionSystemshallbeOPERABLE. 18 099 APPLICABILITY: MODES 1 and 2. ACTION:

a. With one o more Loose-Part Detection System channels inoperable for more than 3 days, prepare and submit a Special Report to the Commission pu uant to Specification 6.9.2 within the next 10 days outlining the c use of the malfunction and the plans for restoring the channel (s) t OPERABLE status.

b. The provisions of S cifications 3.0.3 and 3.0.4 are not applicable.

O SURVEILLANCE REQUIREMENTS 4.3.3.8 Each channcl of the Loose-Part Datect n Systems shall be demonstrated OPERABLE by performance of:

a. A CHANNEL CHECK at least once per 24 hour

b. An ANALOG CHANNEL OPERATIONAL TEST

at least nce per 31 days, and

c. A CHANNEL CALIBRATION at least once per 18 mont .

    =
                                                                                \
        *Setpoint verification is not applicable.

COMANCHE PEAK - UNIT 1 3/4 3-62 ,

TXX-88512 l AllACHMEHi 6 PA?E 87 0F 105 (' INSTRUMENTATION RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 7 3.3.3./Theradioactqveliquideffluentmonitoringinstrumentationchannels shown in Table 3.3- R shall be OPERABLE with their Alarm / Trip Setpoints set to ensure that the limits of Specification 3.11.1.1 are not exceeded. The Alarm / Trip Setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in the OFF3ITE DOSE CALCULATION MANUAL (ODCM). APPLICABILITY: At all times. ACTION:

a. With a radioactive liquid effluent monitoring instrumentation channel Alarm / Trip Setpoint less conservative than required by the above specification, immediately suspend the release of radioactive liquid effluents monitored by the affected channel, or declare the channel inoperable.

b. Witt less than the minime'n number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, take the ACTION shown fm in Table 3.3-H9 Restore the inoperable instruinentation to OPERABLE i

( status within 30 days and, if unsuccessful, explain in the next i Semiannual Radioactive Effluent Release Report pursuant to Specifi-cation 6.9.1.4 why this inoperability was not corrected in a timely manner.

c. The provisions of Specifications 3.0.3 and 3.0.4, are not applicable.

SURVEILLANCE REQUIREMENTS 7 4.3.3./ Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and DIGITAL CHANNEL OPERATIONAL TEST or ANALOG CHANNEL.0PERATIONALTESTatthefrequenciesshowninTable4.3-f. 4 i r l COMANCHE PEAK - UNIT 1 3/4 3-63 , 4 t {

o o n v TABLE 3.3-H 9 n j RADIO?CTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION l M x MINIMUM A CHANNELS 3E INSTRUMENT OPERABLE ACTION c 1. Radioactivity Monitors Providing Alarm and 5

            -4 Automatic Termination of Release

a. Liquid Radwaste Effluent Line 1 _3333

b. Turbine Building (Floor Drains) Sumps Effluent Line 1 -n 34

2. Radioactivity Monitors Providing Alarm But Not Providing Automatic Termination of Release Service Water System Effluent Line 1/ train -3+ 36

{ Y 3. Flow Rate Measurement Devices Liquid Radwaste Effluent Line 1 % N$5 EAE o, z e+ 5" O N 7:n

                                                                                                                       .; , g
                                                                                                                       . -,_g

TXX-88512 AliACHMEHi6 PAGE 89 0F 105 TABLE 3.3-+1-(Continued) ACTION STATEMENTS 33 ACTION 32,- With the number of channels OPERABLE less than regt. ired by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue pri ided that prior to initiating a release:

a. At least two independent samples are analyzed in accordance with Specification 4.11.1.1.1, and

b. At least two technically qualified members of the facility staff independently verify the release rate calculations and discharge valve lineup.

Otherwise, suspend release of radioactive effluents via this pathway. ACTION 'M - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided grab samples are analyzed for radioactivity at a lower limit of detection of no more than 10 7 microcurie /ml:

a. At least once per 12 hours when the specific activity of the secondary coolant is greater than 0.01 microcurie / gram CJD DOSE EQUIVALENT I-131, or

b. At least once per 24 hours when the specific activity of the secondary coolant is less than or equal to 0.01 microcurie / gram DOSE EQUIVALENT I-131.

ACTION % - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, operations may continue provided that:

a. With the component cooling water monitor OPERABLE r.nd indicating an activity of less than (1X10 4] microcurie /ml, a grab sample is col'ected and analyzed for radioactivity at a lower limit of detection,at least every 31 days; or (C 6 nc wt Sr., /C I

                                                                                   />r acc C -c . < /4

b. At least once per 12 hours, grab samples are collected and analyzed for radioactivity at a lower limit of detectiony o f ,3o -c & h a ' m .s c ..- 4 /. f . 0935 3 With the number of channels OPERABLE less than required by the 101 I

ACTION 35 - Minirau.a Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours during actual releases. Pump perfor-mance curves generated in place may be used to estimate flow. O COMANCHE PEAK - UNIT 1 3/4 3-65 ,

(v >

                                                                                                                      ')    .

TABLE 4.3-E RADI0 ACTIVE LIQUID EFFLUENT MDNITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS 5r DIGITAL ANALOG A CHANNEL CHANNEL' 3E CHANNEL SOURCE CHANNEL OPERATIONAL OPERATIONAL

      . INSTRUMENT                                              CHECK       CHECK    CALIBRATION        TEST        TEST

1. Radioactivity Monitors Providing Alarm and Automatic Termination of Release

a. Liquid Radwaste Effluent Line D P R(4) Q(1) N.A.

b. Turbine Building (Floor Drains) Sumps Effluent Line D M R(4) Q(2) N.A.

R 2, Radioactivity Monitors Providing Alarm But

Not Providing Automatic Termination T of Release Service Water System Effluent Line D M R(4) Q(3) N.A. -

3. Flow Rate Measurement Devices Liquid Radwaste Effluent Line D(5) N.A. R N.A. Q

 %29                                                                                                                              '

805

 *EC                                                                                                                            -

V O

n s

 '                                                                                                   I r ;
     . IXX-88512 AllACHMENT 6 PAGE 91 0F 105 TABLE 4.3-3 (Continued')

TABLE NOTATIONS (1) The O!GITAL CHANNEL OPERATIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occur if any of the following conditions exists:

a. Instrument indicates measured levels above the Alarm / Trip Setpoint, or

b. Circuit failure (Channel Out of Service - Loss of Power, Loss of Counts, Loss ofpFlow, or Check Source Failure).

b.w ple (2) The DIGITAL CHANNEL OPERATIONAL TEST shall also demonstrate that automatic

       .           flowdiversionofthispathhay(fromtheLowVolumeWasteTreatment System to the Co-Current Waste Treatment System) and Control Room alarm annunciation occur if any of the following conditions exist:

a. Instrument indicates measured levels above the Alarm / Trip Setpoint,

.or

b. Circuit failure (Channel Out of Service - Loss of Power, Loss of Counts, Loss of Sample Flow, or Check Source F iid).

                                                                         %%ee (3) The DIGITAL CHANNEL OPERATIONAL TEST shall also demonstrate that control
   ,               room alarm annunciation occurs if any of the following conditions exists:

a. Instrument indicates measured levels above the Alarm Setpoint, or

b. Circuit failure (Channel Out of Service Loss of Power, Loss of Counts, Loss of Flow or Check Source Failure).

5 pts (4) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards (NBS) or using standards that have been obtained from suppliers that participate - in measurement assurance activities with NBS. These standards shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration, reference standards certified by NBS, or standards that have been obtained from suppliers that participate in l measurement assurance activities with NBS shall be used. l 1 i (

<~

l 1 COMANCHE PEAK - UNIT 1 3/4 3-67 .

1 + IXX 885l2 , AllACHMENT6-PAGE 92 0F 105 i 4 TABLE 4.3-4 (continued)' DRAFT ' TABLE NOTATIONS-(Continued) l (5) CHANNEL CHECK shall consist of verifying indication of flow during periods of release. CHANNEL CHECK shall be made at least once per 24 hours on days on which continuous, periodic, or batch releases are made.

                                               ~,A l
                                             /
                         ,m'-                      --

mcce Eo ?rui*~5 ?"Y O 4 O COMANCHE PEAK - UNIT 1 3/4 3-68 .

TXX-88512 AliACEENT6 PAGE 93 0F 105 , INSTRUMENTATION RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 9

3. 3. 3. M The radioactive gaseous effluent monitoring instrumentation channels shown in Table 3.3-1M3 hall be OPERABLE with their Alarm / Trip Setpoints set to ensure that the !tnits of Specifications 3.11.2.1 and 3.11.2.5 are not exceeded.

The Alarm / Trip Setpoints of these channels meeting Specification 3.11.2.1 shall be determined and adjusted in accordance with the methodology and parameters in the ODCM. APPLICABILITY: As shown in Table 3.3- W lo ACTION:

a. With a radioactive gaseous effluent monitorir, instrumentation channel Alarm / Trip Setpoint less conservative than required by the above specification, immediately suspend the release of radioactive gaseous effluents monitored by the affected channel, or declare the channel inoperable.

b. With less than the minimum number of radioactive gaseous effluent n.onitoring instrumentation channels OPERABLE, take the ACTION shown m in Table 3.3-12!0 Restore the inoperable instrumentation to OPERABLE hr status within 30 days and, if unsuccessful explain in the next Semi-annual Radioactive Effluent Release Report pursuant to Specifica-tion 6.9.1.4 why this inoperability was not corrected in a timely manner.

c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.10 Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and DIGITAL CHANNEL OPERATIONAL TEST or ANALOG CHANNELOPERATIONALTESTatthefrequenciesshowninTable4.3-f. O COMANCHE PEAK - UNIT 1 3/4 3-69 .

O O O - n TABLE 3.3-1210 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION 9 m v MINIMUM CHANNELS , 9 n INSTRUMENT OPERABLE APPLICABILITY ACTION

1. WASTE GAS HOLDUP SYSTEM +

E M a. Noble Gas Activity Monitor -

 ~             Providing Alarm and Automatic Termination of Release                     1/ stack                   ***         -3&3'7

2. WASTE GAS HOLDUP SYSTEM Explosive Gas Monitoring System

a. Hydrogen Monitors 1/recombiner **

                                                                                                   -41 'f 0  'a c tic   _

K

                                                                                    **            4 2, a r racn e s to        i

b. Oxygen Monitors 2/recombiner ;uc :asm.ca r y . . _ _ , -

M 3. Primary Plant Ventilation

a. Noble Gas Activity Monitor 1/ stack

1*j

b. Iodine Sampler 1/ stack

4 I

c. Particulate Sampler 1/ stack

40- 41 h d. Flow Rate Measuring Device 1/ stack * -37 36 h e. Sampler Flow Rate Monitor 1/ stat.k
0 e

4/5

~%

IXX-88512 ' AliACHMENT 6 PAGE 95 OF 105

                     ~   ~
                                                    \D

(^ TABLE 3.3- M (Continued) TABLE NOTATIONS

At all times.

          ** During WASTE GAS HOLOUP SYSTEM operation.
       *** Ouring Batch Radioactive Releases via this pathway.

ACTION STATEMENTS

        .         II                                                                                 l ACTION g -         With the number of channels OPERABLE less than required by the             i Minimum Channels OPERABLE requirement, the contents of the tank (s) may be released to the environment provided that prior to initiating the release:

a. The auxiliary building vent duct monitor is confirmed OPERABLE, or

b. At least two independent samples of the tank's contents are analyzed, and

c. At least two technically qualified members of the facility staff independently verify the release rate calculations and discharge valve lineup.

I Otherwise, suspend release of radioactive effluents via this pathway. l ACTION 37 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requir nent, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours. El ACTION g - With the number of channels OPERABLE less than required by the { Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided grab samples are taken at least once per 12 hours and these samples are analyzed for radioactivity within 24 hours. l 40 ACTION)df- With the number of channels OPERABLE one less than required by l the Minimum Channels OPERABLE requirement, suspend oxygen supply l to the recombiner. Addition of waste gas to the system may con-l tinue provided grab samples are taken and analyzed at least once l per 4 hours during degassing operations or at least once per l 24 hours during other operations and the oxygen concentration remains less than 1 percent. ACTION g - With the number of channels OPERABLE less than required by the l Minimum Channels OPERABLE requirement, effluent releases via I the affected pathway may continue provided samples are contin-l uously collected with auxiliary sampling equipment as required in tam ; t.11-2.- l 10 8: 0213 ! COMANCHE PEAK - UNIT 1 3/4 3-71 , t

i e l

    ,      IXX 88512 ATTACHMENT 6 PAGE 96 0F 105
                          ~  ~

l s 30 TABLE 3.3-4F (Continued) ]'. , TABLE NOTATIONS (Continued) l l 0 l ACTION 4 a. With the outlet oxygen monitor channel inoperable, opera- i tion of the system may continue provided grab samples are taken and analyzed at least once per 24 hours and the oxy-gen concentration remains less than 1 percent.

b. With the inlet oxygen monitor inoperable, operation may continue if inlet hydrogen monitor is OPERABLE.: c. With both oxygen channels or both of t e inlet oxygen and

. inlet hydrogen monitors inoperable, suspend oxygen supply J tc, the recombiner. Addition of waste gas to the system , may continue provided grab samples are taken and analy.':ed at least once per 4 hours during degassing operations or at least once per 24 hours during other coerations and the oxygen concentration remains less than 1 percent. S O COMANCHE PEAK - UNIT 1 3/4 2-72 -

v (o n TABLE 4.3-6 f RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS 9 m m DIGITAL ANALOG S CHANNEL CHANNEL 7 CHANNEL S0llRCE CHANNEL OPERATIOM L OPERATIONAL

 .              INSTRLMENT                                 CHECK        CHECK   CALIBRATION      TEST                  TEST c                                                                                                                                          ,

I 3 w 1. WASTE GAS HOLDUP SYSTEM e

a. Noble Gas Activity Monitor -

Providing Alarm and Automatic

 ;                       Termination of Release               P           P        R(3)            Q(1)              N.A.

1

2. WASTE GAS HOLDUP SYSTEM Explosive j ,

Gas Monitoring System s [ a. Hydrogen Monitors D N.A. Q(4) N.A. M

;           U        b. Oxygen Monitors                      D           N.A.      Q(4)            N.A.              M i                3. Primary Plant Ventilation 1

, a. Noble Gas Activity Monitor D Mf R(3) Q(2) N.A. l i

b. Iodine Sampler W(S) H.A. N.A. N.A. N.A.

l l ,,_

c. Particulate Sampler W(5) N.A. N.A. N.A. N.A.

! k39

         .g=         d. Flow Rate Measuring Ocsite           D           N.A.       R              N.A.              Q "4r nxn

_" ; e. Sampler Flow Rate Monitor N.A. i D N.A. R Q l 3 i Cl5

n l

i

~ in-88512 ATTACMENT 6 PAGE 98 0F 105 6 TABLE 4.3-R (Continued)- y TABLE NOTATIONS l

       #Also prior to any release from the waste gas holdup system or containment purging or venti.ng.

(1) The DIGITAL CHANNEL OPERATIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if any of the following conditions exists:

a. Instrument indicates measured levels above the Alarm / Trip Setpoint, or

b. Circuit failure (Channel Out of Service - Loss of Power, Loss of Counts, Loss of Sample / Flow, or Check Source Failure).

(2) The OIGITAL CHANNEL OPERATIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the follu sing conditions exists:

a. Instrument indicates measured levels above the Alarm Setpoint, or

b. Circuit failure (Channel Out of Service - Loss of Power, Loss of Counts,LossofSample/ Flow,orCheckSourceFailure).

(3) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards (NBS) or using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS. These standards shall permit V('N calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration, reference standards certified by NBS, or standards that have been obtained from suppliers that participate in measurement assurance activities with NBS shall be used. (4) The CHANNEL CALIBRATION shall include the use of standard gas samples r[N k.k'N .

a. One vcle percent hydrogen, ba'aa" nuracan- act i 10 h 0216 Woue-volwpeccent-hy4cogen, balance aitrogea l (5) The Channel Check shall consist of visually verifying that the collection element (i.e. , filter or cartridge, etc.) is in place for sampling.

O COMANCHE PEAK - UNIT 1 3/4 3-74 , f

TXX-88512 AliACHMENT6 PAGE 99 OF 105 ) g INSTRUMENTATION yn ' < Ul k 3/4.3.4 TURBINE OVERSPEED PROTECTION LIMITING CONDITION FOR OPERATION , j 3.3.4 At least one Turbine Overspeed Protection System shall be OPERABLE. APPICABILITY: MODES 1, 2*, and 3*. ACTION: i Ccnktc\

a. With one stop valve or one gev:ract valve per high pressure turbine steam line inoperable and/or with one reheat stop valve or one G7drol +=heet intercept valve per low pressure turbine steam line inoperable, restore the inoperable valve (s) to OPERABLE status within 72 hours, or close at least one valve in the affected steam line(s) or isolate the turbine from the steam supply within the next 6 hours,

b. With the above required Turbine Overspeed Protection System otherwiti inoperable, within 6 hours isolate the turbine from the steam supply.

SURVEILLANCE REQUIREMENTS 4.3.4.1 The provisions of Specification 4.0.4 are not applicable.

4.3.4.2 The above required overspeed protection system shall be demonstrated OPERABLE:

a. At least once per 14 days by cycling each of the following valves through at least one complete cycle from the running position using the manual test or Automatic Turbine Tester (ATT):

1) Four high pressure turbine stop valves,

2) Four high pressure turbine control valves, Four"PStx low pressure turbine stop valves, and Fourb-44x low pressure turbine control valves,

b. At least once per 14 days by testing of the two mechanical overspeed devices using the Automatic Turbine Tester or manual test.

c. At least once per 31 days by direct observation of the movement of each of the above valves through one complete cycle from the running position,

d. At least once per 40 months by disassembling at least one of each of the above valves and performing a visual and surface inspection of valve seats (if applicable), disks and stems and verifying no unaccept-able flaws. If unacceptable flaws are found, all other valves of that type shall be inspected.

      *Not applicable in MODES 2 and 3 with all main steam line isolation valves and associated bypass valves in the closed position.

COMANCHE PEAK - UNIT 1 3/4 3-75 .

1 IIX-88512 ATTACHMENT 6 PAGE 100 0F 105 [ 3/4.3 INSTRUMENTATION l BASES l 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION The OPERABILITY of the Reactor Trip System and the Engineered Safety l Features Actuation System instrumencation and interlocks ensures that: (1) the associated ACTION and/or Reactor trip will be initiatec when the parameter monitored by each channel or combination thereof reaches its Setpoint (2) the specified coincidence logic and sufficient redundancy is maintained to permit a channel to be out-of-service for testing or maintenance consistent with main-taining an appropriate level of reliability of the resctor protection and engi-neered safety features instrumentation, and (3) sufficient system functional capability is available from diverse parameters. The OPERABILITY of these systems is required to provide the overall reliability, redundancy, and diversity assumed available in the facility design for the protection and mitigation of r.ccident and transient conditions. The integrated operation of each of these systems is consistent with the assumptions used in the safety analyses. The Surveillance Requirements specified for these systems ensure that the overall system functional capability is maintained com-parable to the original design standards. The periodic surveillance tests per-formed at the minimum frequencies are sufficient to demonstrate this capability. Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with WCAP-10271, "Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrumentation System", and supplements to that report as approved by the NRC and documented in , the SER (letter to J. J. Sheppard from Cecil 0. Thomas, dated February 21, 1985). The Engineered Safety Features Actuation System Instrumentation Trip l Setpoints specified in Table 3.3-4 are the nominal values at which the bistables are set for each functional unit. A Setpoint is considered to be adjusted consistent with the nominal value when the "as measured" Setpoint is within the band allowed for calibration accuracy.

To accommodate the instrument drift assumed to occur between operational tests and the accuracy to which Setpoints can be measured and calibrated, Allowable Values for the Setpoints have been specified in Table 3.3-4. Opera-tion with Setpoints less conservative than the Trip Setpoint but within the Allowable Value is acceptable since an allowance has been made in the safety analysis to accommodate this error. An y tional provision has been inoluded , l for determining the OPERABILITY of a channel when its Trip Setpoint is found i to exceed the Allowable Value. The methodology of this option utilizes the l

     "as measured" deviation from the specified calibration point for rack and sensor components in conjunction with a statistical combination of the other uncertainties of the instrumentation to measure the process variable and the

uncertainties in calibrating the instrumentation. In Equation 3.3-1,

! Z + R + S < TA, the interactive effects of the errors in the rack and the l sensor, and the "as measured" values of the errors are considered. Z, as specified in Table 3.3-4, in percent span, is the statistical summation i of errors assumed in the analysi: excluding those associated with the sensor i and rack drift and the accuracy of their measurement. TA or Total Allowance

O is the difference, in percent span, R or Rack Error is the "as measured" deviation, in the percent span, for the affected channel from the specified Trip Setpoint. S or Sensor Error is either the "as measured" deviation of ;

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

l TXX-88512 ATTACHMENT 6 PAGE 102 0F 105 INSTRUMENTATION BASES REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued) the sensor from its calibration point or the value specified in Table 3.3-4, in percent span, from the analysis assumptions. Use of Equation 3.3-1 allows for a sensor draft factor, an increased rack drift factor, and provides a threshold value for REPORTABLE EVENTS. The methodology to derive the Trip Setpoints is based upon combining all of the uncertainties in the channels. Inherent to the determination of the Trip Setpoints are the magnitudes of these channel uncertainties. Sensor and rack instrumentation utilized in these channels are expected to be capable of operating within the allowances of these uncertainty magnitudes. Rack drif t in excess of the Allowable Value exhibits the behavior that the rack has not met its allowance. Being that there is a small statistical chance that this will happen, an infrequent excessive drift is expected. Rack or sensor drift, in excess of the allowance that is more than occasional, may be indicative of-more serious problems and should warrant further investigation. The measurteent of response time at the specified frequencies provides assurance that the Reactor trip and the Engineered Safety Features actuation associated with each channel is completed within the time limit assumed in the p safety analyses. No credit w&s taken in the analyses for those channels with Q response times indicated as not applicable. Response time may be demonstrated by any series of sequential, overlapping, or total channel test measurements provided that such tests demonstrate the total channel response time as defined. Sensor response time verification may be demonstrated by either: (1) in place, onsite, or offsite test measurements, or (2) utilizing replacement sensors with certified response time. The Engineered Safety Features Actuation System senses selected pl, ant parameters and determines whether or not predetermined limits are being exceeded. If they are, the signals are combined into logic matrices sensitive to combinations indicative of various accidents events, and transients. Once the reqe. fred logic combination is completed, the system sends actuation signals to those Engineered Safety Features components whose aggregate function best serves the requirements of the condition. As an example, the following actions may be initiated by the Engineered Safety Feature: Actuation System to mitigate the consequences of a steam line break or loss-of-coolant accident: (1) ECCS pumps start and automatic valves position, (2) Reactor trip, (3) feed water isolation, (4) startup of the emergency diesel generators, (5) containment spray pumps start and automatic valves position (6) containment isolation, (7) steam line isolation, (8) turbine trip, (9) auxiliary feedwater pumps start and automatic valves position, (10) station service water pumps start and automatic valves position, (11) Control Room Emergency Recirculation starts, and (12) essential ventilation systems (safety chilled water, electrical area fans, primary plant ventilation ESF exhaust fans, battery room exhaust fans, and UPS ventilation) start. O COMANCHE PEAK - UNIT 1 B 3/4 3-2 ,

TXX-88512 ATTACHMENT 6 F ME 102 0F 105

 -    INSTRUMENTATION V   BASES REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued)

To satisfy the recommendations set forth in Section 4.7 of IEEE 279-1971, in the event that one of the three channels of high steam generator level protection is used for level control that channel shall be placed in the tripped condition until level control is returned to its normal channel. The Engineered Safety Features Actuation System interlocks perform the following functio'4: P-4 Reactor tripped - Actuates Turbine trip, closes main feedwater valves on T,yg below Setpoint, prevents the opening of the main feedwater valves which were closed by a Safety Injection or High Steam Generator Water level signal, allows Safety Injection block so that components can be reset or tripped. Reactor not tripped prevents manual block of Safety Injection. P-11 On increasing pressurizer pressure, P-11 automatically reinstates Safety Injection actuation on low pressurizer pressure and low steam line pressure. On decreasing pressure, P-11 allows the manual block of Safety Injection actuation on low pressurizer pressure and low O steam line pressure. 3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING FOR PLANT OPERATIONS The OPERABILITY of the radiation monitoring instrumentation for. plant operations ensures that: (1) the associated action will be initiated when the radiation level monitored by each channel reaches its Setpoint, and (2) suffi-cient redundancy is maintained to permit a channel to be out-of-service for testing or maintenance. The radiation monitors for plant operations senses COMANCHE PEAK - UNIT 1 B 3/4 3-3 ,

l TXX-88512 ATTACHMENT 6 PAGE 103 0F 105

   -          INSTRUMENTATION                                                                 DP.*.:W.. .

I 1 V BASES 3/4.3.3.1 RADIATION MONITORING FOR PLANT OPERATIONS (Continued) radiation levels in selected plant systems and locations and determines wnether or not predetermined limits are being exceeded. If they are, the system sends actuation signals to initiate alarms or actuate Control Room Emergency Recir-culation or actuate Containment Ventilation Isolation. 3/413.2 MOVABLE INCORE DETECTORS "[ IDI 0936 The OPERASILITY of the movable incore detectors with the specified minimum complement of equi) ment ensures that the measurements obtained from use of this system accurately Fepresent the spatial neutron flux distribution of the core. The OPERABILITY of this' system is demons,trated by irradiating each detector used and determining theleceptability t of its voltage curve. For the purpose of measuring F (Z) or F AH full incore flux map is used. 9 Quarter-core flux maps, as defined in WCAP-8648, June 76, may be used in recalibration of the Excore Neutron Flux Detection System, full incore flux maps or symmetric incore thimbles may be used for monitori he QUADRANT POWER TILT RATIO when one Power Range channel is inoperable. 3/ h 3J SEISMIC INSTRUMENTATION p d The OPERAB capability is availabl the seismic instrumentation ensures that sufficient romptly determine the magnitude of a seismic event and evaluate the response of features important to safety. This capa-bility is required to permit compar1 f the measured response to that used in the design basis for the facility to de ne if plant shutdown is required pursuant to 10 CFR 100 Appendix A. The instrumen on is consistent with the recommendations of Regulatory Guide 1.12, "Instrumentat or Earthquakes,"

           . April 1974.

3/4 ETEOR0 LOGICAL INSTRUMENTATION The OPERABILI the meteorological instrumentation ensures that suffi-cient meteorological data a available for estimating potential radiation doses to the public as a result outine or accidental release of radioactive materials to the atmosphere. This cap ity is required to evaluate the need

for initiating protective measures to protec health and safety of the i public and is consistent with the recommendations oposed Revision 1 to Regulatory Guide 1.23 "Meteorological Programs in suppo Nuclear Power Plants," September 1980.

3/4.3.3.Ir REMOTE SHUTOOWN SYSTEM l adenld.m buske wW ~d calrcIs l TheOPERABILITYofthegemote hutdown 4ystem ensures that sufficient l l capability is available to permit s[afe shutdown of the facility from location l outside of the control room. This capability is required in the event control room habitability is lost and is consistent with General Design Criterion 19 l O of 10 CFR 50. s COMANCHE PEAK - UNIT 1 B 3/4 3-4 ,

IXX-88512 ATTACHTENT 6 PAGE 104 0F 105 INSTRUMENTATION DBfT BASES ID 1: 0466 REMOTE SHUTOOWN SYSTEM (Continued) gg_g%m Mder sWWs %L The OPERABILITY of the emota hutdown $yMenkensures that a fire will notprecludeachievingsafe[shutdow[n. The jfemote ghutdown SyMem instrumenta-tion, control 5, :.d pe' a circuits and transfer switches necepsary to eliminate effects of the fire and allow operation of instrumentation / Tontrolsend p;ter cir:;its required to achieve an~d maintain a safe shutdown condition are inde-pendent cf areas where a fire could damage systems normally used to shut down the reactor. This capability is consistent with General Design Criterion 3 of 10 CFR 50. 3/4.3.3. d ACCIDENT MONITORING INSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters for which pre-planned manually controlled operator actions are required to accomplish safety functions for recovery from Design Basis Accidents, as defined by the plant safety analysis. This capability meets the intent of the recommendations of Regulatory Guide 1.97, Revision 2, "Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident," December 1980 and those requirements of NUREG-0737, "Clarification of TMI Action Plan Requirements," November 1980 that apply to CPSES. m ._ , 4 , 4

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                                                                                   *:tc4.Uh:    N :55. *5b mmm.4...      m4.s u EE l e UEEed b C . 2 5!'1.i l 5 b d!b UN 7 Icr 5EEE EghEE:E 5 EEEENIed NetEb

_ . , , . m om The specific calibration provisions for the Containment Radiation (High Range) Monitor are in accordance with the provisions of NUREG-0737, Item II.F.1. 3/4.3.3.7 CHLORINE DETECTION SYSTEMS The OPERABILITY of the Chlorine Detection Systems ensures that sufficient capability is available to promptly detect and initiate protective action in the event of an accidental chlorine release. This capability is required to protect control room personnel and is consistent with the recommendations of Regulatory Guide 1.95, Revision 1, "Protection of Nuclear Power Plant Control room Operators Against an Accidental Chlorine Release," January 1977. This capability will not be required if the quantity of chlorine gas stored onsite is small (120 lbs.) and utilized for laboratory and calibration purposes. This applicability is consistent with the exclusions and recommendations of Regulatory Guide 1.95, Revision 1, "Protection of Nuclear Plant Control Room Operators Against an Accidental Chlorine Release," January 1977. ,, ., 4/4_.3.3.8' LOOSE PART DETECTION SYSTEM . ma TDPERASILITY of the Loow Part Detection System ensures that sufficient capability is avathble to detect loose metallic parts in the Reactor System and avoid or mitigate dama W (o Tseactor System components. The allowable out-of-service times and surveilTancagequirements are consistent with the O recommendations of Regulatory Guide 1.1337 "L ose-Part Detection Program for the Primary System of Light-Water-Cooled Reactor ," a 1981. COMANCHE PEAK - UNIT 1 8 3/4 3-5 .

TXX-88512 AllACEENT 6 FACE 105 0F 105 INSTRUMENT TION BASES 7 3/4.3.3.9' RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The Alarm / Trip Setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the 00CM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR 20. The 6PERA-BILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of 10 CFR 50 Appendix A. 8. 3/4.3.3.10 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The Alarm / Trip Setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the 00CM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR 20. This instru-mentation also includes provisions for monitoring (and controlling) the con-centrations of potentially explosive gas mixtures in the WASTE GAS HOLDUP SYSTEM. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of 10 CFR 50 Appendix A. The sensitivity of any noble gas activity monitors used to show compliance with

 ,' ,/ l   the gaseous effluent release requirements of Specification 3.11.2.2 shall be such that concentrations as low as 1 x 10 6 pCi/ml are measurable.

3/4.3.4 TURBINE OVERSPEED PROTECTION This specification is provided to ensure that the turbine overspeed protection instrumentation and the turbine speed control valves are OPERABLE and will protect the turbine from excessive overspeed. Protection from turbine excessive overspeed is required since excessive overspeed of the turbine could generate potentially damaging missiles which could impact and damage safety-related components, equipment or structures. O COMANCHE PEAK - UNIT 1 B 3/4 3-6 i I

TXX-88512 ATTACHMENT 7 PAGE I 0F 66 O COMANCHE PEAX STEAM ELECTRIC STATION TECHNICAL SPECIFICATION 3/4.4 O l O l ~. .. _. -- - - -__ _ _ -__ _

ixx-88512 AliACHMENT 7 PAGE 2 0F 66 .fm CPSES Technical Specifications t V) NRC Draft 2 Markup-Section 3/4.4 Chance ID# Justification For Chance 0223 Delete Specification 3.4.2.1, Pressurizer Safety Valves, 0470 MODES 4 and 5. At low temperature the Pressurizer Safety Valves do not provide adequate overpressure protection for the Reactor Coolant System, more importantly the Reactor Vessel. This is shown by the RCS Pressure Temperature limits of Specification 3.4.9.1 based on 10CFR50 Appendix G considerations. In order to protect the RCS under these conditions, Low Temperature Overpressure Protection (LTOP) is provided by Specification 3.4.9.3. Based on the latest analysis, including additional postulated overpressure transients and additional conservative margins for instrumentation errors, time delays, etc., LTOP is applicable at all times in MODES 4, 5 and 6. Required LTOP protection is redundant and therefore single active failure proof, except for the completely passive option of an RCS vent to atmosphere.

/'              This is much more conservative than Specification 3.4.2.1

( which is not single failure proof and will not protect all portions of the RCS. It should also be noted that the Action Statement for no pressurizer Code safety valve OPERABLE includes placing an OPERABLE RHR loop in operatien which of course provides relief nrotection via the RHR relief. This Action, however, is less conservative than the normal require:nents of Specification 3.4.9.3. Note also that CPSES has proposed the RHR relief valves as an acceptable method of overpressure protection that could be used in lieu of the RCS PORVs, or RCS vent. Therefore, because this specification does not provide adequate protection, and is less conservative than existing Specification 3.4.9.3, it should be deleted. This is exactly the same line of reasoning as to why 3.4.2.1 never included MODE 6 - by definition, the reactor vessel head is removed or at least de-tensioned thus guaranteeing overpressure protection. Deletion of the specification prevents meaningless administrative requirements. Deletion also facilitates testing and maintenance in MODE 5 when it may be desirable to remove all safety valves for refurbishment or bench testing. O

g IXM'.512 { ATTACHMEWi7 rAst 3 % M

                                 )
                                 ]
                                 }
   ,q                       CPSES Technical Specifications Q                               NRC Draft 2 Markup Section 3/4.4 Change 10#      Justification For Change 0224      The applicability of this specification is for Modes 1, 2 and 3. Specification 3.0.1 states "Compliance with the LC0 contained in the succeeding specifications is required during the Operational Modes or other conditions specified therein;..." This statement is interpreted to ensure the LC0 is met anytime the Applicability Mode is entered. The "other conditions specified" are interpreted to be conditions explicitly spelled out in the Applicability Statement such as the accumulator specification that spells out tho specified condition of pressurizer pressure above 1000 psig and in Mode 3.      This change is similar to that Licensed at Millstone 3 and Vogtle.

0230 Added

note to allow leakrate testing at pressures below 2235 psig. This is permitted in ASME Section XI IWV-3423 (e) for valves where leakrate would be expected to diminish at the higher pressures as a result of leakage channel reduction due to the forces on the disc. This would not be allcwed where system pressure is exerted under the seat of a lobe valve for example. This change bO is similar to that L censed at South Texas, Vogtle, Callaway, Shearon Harris, Byron and Wolf Creek.

0233 This change is made to reflect a more meaningful test 0475 methodology and to attach appropriate testing frequencies to ensure the required safety injection flow does not bypass the reactor core. The actual setting or verification of setting is performed on an 18 month basis or following any modifications, maintenance or operation that ef fects the seal injection flow path. This is consistent with the requirements placed on the ECCS throttle valves which one would consider to be more important to the mitigation of a LOCA and by performing this test on an 18 month basis prevents having to start a

centrifugal charging pump (CCP) on a monthly basis which will prevent unnecessary CCP degradation. This rationale of preventing pump degradation was one of the major reasons for extending the test frequency of the ASME Section XI program from monthly to quarterly. In addition, these valves are locked in their throttle oosition and verified at least once per 31 days which exceeds the requirements placed on the ECCS throttle valves.

4 v

TXX-88512 AllACHMDli 7 PAGE 4 0F 66 CPSES Technical Specifications NRC Draft 2 Markup p/ y Sectic; 3/4.4 Change 10# Justification For Change 0234 This change is made to exclude tne provisicns of Specification 4.0.4 for entry into MODE 3 and 4 and make allowances for deferral of the water inventory balance during transient conditions. During normal MODE ascension the RHR system remains inservice until just before the transition from MODE 4 to MODE 3. During this period the RCS temperature is being raised as part of the controlled heatup. The results of the water inventory balance are very sensitive to the total volumes included in the defined system and to changes-in the temperatures within the defined system. With the RHR system connected to the RCS, the volume of the defined system has increased dramatically and the ability to stabilize the temperature throughout the system to enable accurate mass calculation has decreased. In addition, during this period there is normally a significant amount of inspection and testing activity being conducted in the containment and the auxiliary

              .                                                    buildings so the probability of detecting abnormal leakage

( visually is increased (thus presumably reducing the k imediate need for the indirect method for detennining leakage i.e., water inventory balance). In addition, the three primary leak detection systems are still operative. The ability to defer the surveillance requirement for a maximum of 96 hours instead of the allowed 72 hours accommodates unanticipated load swings which might occur just prior to the scheduled surveillance time without risk of unwarranted invocation of 4.0.3. The change concerning MODE entry appears to be consistent with the guidelines in Generic Letter 87-09 dated June 4, 1987. The portion concerning water inventory balance

during transient conditions was Licensed at Seabrook (no PRA required or performed).

0237 This Surveillance Requirement and Table are being 0239 relocated to the CPSES Technical Specification Improvement 0481 Program. TV Electric believes the inclusion of this Surveillance Requirement and Table is unnecessary and the rec uirements and information would be more appropriately adc ressed in the CPSES Technical Specification Improvement Prog ram. Q - V i

IXX-88512 ATTACHM N1 7 fAGE 5 Of eM p- CPSES Technical Specifications d , NRC Draft 2 Markup Section 3/4.4 Chanae ID# Justification For Chanae 0481 (cont.) Relocation of this Surveillance Requirement and Table is consistent with the guidance provided in the NRC's Interim Policy Statement (52FR3788), February 6, 1987, and the recommendations of the Westinghouse Owners Group MERITS Program. Priority is given to the relocation of this Surveillance Requirement and Table since the detailed information is not used by the Licensed Operator, but is purely used to determine the effect of radiation on the vessel which is used to update the Pressure and Temperature limits. The information and requirements currently in this Surveillance Requirement and Table are more appropriately maintained in a document subject to TV Electric administrative control and 10CFR50.59 review under the CPSES Technical Specification Improvement Program. This change is similar to that Licensed at Seabrook, Shearon Harris and Vogtle. O s_/ 0238 See ID# 0240 0239 See 10# 0237 0240 The maximum auxiliary spray differential temperature limit 0238 should be deleted since during normal operation this 0478 differential temperature can not be reached. This is based on plant design allowing the coldest water to be injected which would be 40 degrees Fahrenheit water from the RWST and the maximum temperature coming from saturation temperature that corresponds to the relief setting of the first PORV which is 2335 psi (658.05 degrees Fahrenheit). Therefore the maximum differential temperature is 618.05 degrees Fahrenheit. This LC0 and i Surveillance Requirement has no significant impact on l normai operations and therefore should be deleted. L 0242 This change adds the allowance to use the RHR relief 0244 valves as a viable alternative to meet the low 0245 temperature overpressure requirements. The RHR relief 0480 valves have sufficient relief capacity to envelope the PORVs currently used to meet this requirement and do not have any of the processing delay that are associated with the PORVs. Technical information is presently not available and will be supplied upon completion of the ! required analysis. The change to the LC0 (including l l l l

TXX-88512 l AriACMENT 7 l PAGE 6 0F 66 l 1

: l

    )                      CPSES Technical Specifications NRC Draft 2 Markup Section 3/4.4 (cont.)         Actions) and Surveillance Requirements has been made to appropriately implement this change. This change is simi!ar to that licensed at Millstone, Byron, Seabrook, Callaway, Vogtle at.d Wolf Creek.

0470 See 10# 0223 0475 See 10# 0233 0478 See 10# 0240 0480 See 10# 0242 0481 See 10# 0237 0906 This Action Statement is revised to allow following the appropriate Action Statement for the plant condition being addressed. This change in no-way lessens the protection afforded by any of the present Action Statements. Based on 3.0.2, without this change once the Action Statement is entered it cannot be deviated from until the LC0 is met. O This is considered to be overly restrictive in this Q situation. 0937 This Technical Specification is being relocated to the CPSES Technical Specification Improvement Program. TV Electric believes the inclusion of this Specification is unnecessary and the information would be more appropriately addressed in the CPSES Technical Specification Improvement Program. Relocation of this Specification is consistent with the guidance provided in the NRC's Interim Policy Statement (52FR3788), February 6,198i, and the recommendations of the Westinghouse Owners Group MERITS Program. Priority is given to the relocation of this Specification since the detailed information is not used by the Licensed Operator, and requires no immediate action from the Licensed Operator if the Action Statement is applied. The information currently in this Specification is more appropriately maintained in a document subject to TV Electric administrative control and 10CFR50.59 review under the CPSES Technical Specification Improvement Program. 0

TH C 512 LiiACHMOli 7 FA E 7 0F 66 n V CPSES Technical Specifications 5RC Draft 2 Markup Section 3/4.4 Change 10# Justification For Change 0937(cont.) This change is similar to that Licensed at Shearon Harris, Seabrook and Vogtle. 0938 The Valves 8705A/B should be deleted from Table 3.4-1. Table 3.4-1 contains a list of RCS pressure isolation valves for which operational leakage shall be limited to 1 gpm during MODES 1,2,3 and 4. The NRC's intent in this specification is to limit intersystem loss of coolant from the RCS to major depressurized subsystems. In particular, the concept was to limit leakage past the SAFETY INJECTION SYSTEM check valves in addition to the RHR suction isolation valves. Valves 8705A/B are 3/4-inen check valves which provide a means of pressure relief from the volume of trapped water between the RHR suction isolation valves. valves 8705A/B are not classified as Reactor Coolant System pressure isolation valves despite the fact they are in parallel with the innermost RHR suction

isolation valves. This is due to the fact that valves 8705A/B are located in a flowpaths which have orifices to prevent RCS leakage in excess of the makeup system. If this valve failed to isolate RCS pressure, the potential mass flow into the RHR system could be handled by the RHR suction reliefs. a J

i 11188512 ' AllACHMENT7 PAGE 8 0F 66 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION aI STARTUP AND POWER OPERATION LIMITING CONDITION FOR OPERATION 3.4.1.1 All four (4) reactor coolant loops shall be in operation. APPLICABILITY: MODES 1 and 2. ACTION: With less than the above required reactor coolant loops in operation, be in at least HOT STANOBY within 6 hours. O SURVEILLANCE REQUIREMENTS 4.4.1.1 The above required reactor coolant loops shall be verified in operation and circulating reactor coolant at least once per 12 hours. i O COMANCHE PEAK - UNIT 1 3/4 4-1 .

r IXX88512

     . ATTACHMENT 7 PAGE 9 0F 66 REACTOR COOLANT SYSTEM
       ' HOT STANDBY LIMITING CONDITION FOR OPERATION 3.4.1.2 At least two of the reactor coolant loops listed below shall be OPERABLE with at least two reactor coolant loops in operation when the reactor trip breakers are closed and at least one reactor coolant loop in operation when the reactor trip breakers are open:"

a. Reactor Coolant Loop 1 and its associated steam generator ano reactor coolant pump,

b. Reactor Coolant Loop 2 and its associated steam generator and reactor coolant pump,

c. Reactor Coolant Loop 3 and its associated steam generator and reactor coolant pump, and

d. Reactor Coolant. Loop 4 and its associated steam generator and reactor coolant pump.

APPLICABILITY: MODE 3.** ACTION: O a. With less than the above required reactor coolant loops OPERABLE,

   \                   restore the required loops to OPERABLE status within 72 hours or be in HOT SHUTOOWN within the next 12 hours,

b. With only one reactor coolant loop in operation and the reactor trip breakers in the closed position, within 1 hour restore two loops to operation or open the reactor trip breakers.; c. With no reactor coolant loop in operation, open the reactor trip breakers and suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate i corrective action to return the required reactor coolant loop to
operation.

SURVEILLANCE REQUIREMENTS 4.4.1.2.1 At least the above required reactor coolant pumps, if not in operation, shall be determined OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability, i

         *All reactor coolant pumps may be deenergized for up to I hour provided:

l (1) no operations are permitted that would cause dilution of the Reactor l Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10*F below saturation temperature.

       **See special test exceptions Specification 3.10.4.

COMANCHE PEAK - UNIT 1 3/4 4-2 ,

r l TX1-88512 j f ! AllACHMENT7 PAGE 10 0F M REACTOR COOLANT SYSTEM HOT STANOBY SURVEILLANCE REQUIREMENTS (Continued) 4.4.1.2.2 The required steam generators shall be determined OPERABLE by " verifying secondary side water level to be greater than or equal to 10% (narrow range) at least once per 12 hours. 4.4.1.2.3 The required reactor coolant loops shall be verified in operation and circulating reactor coolant at least once per 12 hours. O l i l COMANCHE PEAK - UNIT 1 3/4 4-3 .

1X1 88512 . AllACHMENI7 PAGE 11 Of 66 REACTOR COOLANT SYSTEM wHnlI C HOT SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.1.3 At least two of the loops listed below shall be OPERABLE and at least one of these loops shall be in operation:"

a. Reactor Coolant Loop 1 and its associated steam generator and reactor coolant pump,**

b. Reactor Coolant Looo 2 and its associated steam generator and reactor coolant pun..,"* .

c. Reactor Coolant Loop 3 and its associated steam generator and reactor coolant pump,"*

d. Reactor Coolant Loop 4 and its associated steam generator and reactor coolant pump,**

e. RHR Loop A, or

f. RHR Loop B.

O( APPLICABILITY: MODE 4. ACTION:

a. With less than the above required loops OPERABLE, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible; if the remaining OPERABLE loop is an RHR loop, be in COLD SHUTOOWN within 24 hours.

                  "All reactor coolant pumps and RHR pumps may be deenergized for up to I hour provided: (1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10'F below saturation temperature.
                 **A reactor coolant pump shall not be started in Mode 4 unless the secondary O                 water temperature of each steam generator is less than 50*F above each of the Reactor Coolant System cold leg temperatures.

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

TXX88512

AllACHMENT1 FAGE 12 0F 66 REACTOR COOLANT SYSTEM HOT SHUTDOWN LIMITING CONDITION FOR OPERATION

b. With no loop in operation, suspend all operations involving a reduction in boron c_oncentration of the Reactor Coolant System and immediately initiate corrective action to return the required loop to operation. -

SURVEILLANCE REQUIREMENTS 4.4.1.3.1 The required reactor coolant pump (s), and/or RHR pump (s) if not in operation, shall be determined OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability. 4.4.1.3.2 The required steam gen m tor (s) shall be determined OPERABLE by verifying secondary side water level to be greater than or equal to 10% (narrow range) at least once per 12 hours. 4.4.1.3.3 At least one reactor coolant or RHR loop shall be verified in operation and circulating reactor coclant at least once per 12 hours. l l [ l l i l 4 l l 1 i l l 1O t l j COMANCHE PEAK - UNIT 1 3/4 4-5 ' 1 i . _ . . - _ _ _ - , - - _ - - _

TXX-88512 ATTACHMENT 7 FAGE 13 0F 66 REACTOR COOLANT SYSTEM COLD SHUT 00WN - LOOPS FILLE 0 LIMITING CONDITION FOR OPERATION 3.4.1.4.1. At least one residual heat removal (RHR) loop shall be OPERAC! E and in operation *, and either:

a. One additional RHR loop shall be OPERABLE **, or

b. The secondary side water level of at least two steam generators shall be greater than or equal to 10% (narrow range).

APPLICABILITY: MODE 5 with reactor coolant loops filled ***. ACTION:

a. With one of the RHR loops inoperable or with less than the required steam generator water level, immediately initiate corrective action to return the inoperable RHR loop to OPERABLE status or restore the required steam generator water level as soon as possible,

b. With no RHR loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR O locp to operation.

SURVEILLANCE REQUIREMENTS 4.4.1.4.1.1 The secondary side water level of at least two steam generators when required shall be determined to be within limits at least once per 12 hours. 4.4.1.4.1.2 At least one RHR loop shall be determined to be in operation and circulating reactor coolant at least once per 12 hours.

      *The RHR pump may be deenergized for up to I hour provided:        (1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10*F below saturation temperature.
     **0ne RHR loop may be inoperable for up to 2 hours for surveillance testing provided the other RHR loop is OPERABLE and in operation.
   ***A reactor coolant pump shall not be started in Hode 5 unless the secondary O       water temperature of each steam generator is less than 50*F above each of the Reactor Coolant System cold leg temperatures.

COMANCHE PEAK - UNIT 1 3/4 4-6

II M 8512 ATTACHMENT 7 I

PAGE 14 0F 66 l REACTOR COOLANT SYSTEM i COLO SHUTOOWN - LOOPS NOT FILLE 0 LIMITING CONDITION FOR OPERATION 3.4.1.4.2 Two residual heat r2moval (RHR) loops shall be OPERABLE

and at least one RHR loop shall be in operation.**

APPLICABILITY: MODE 5 with reactor coolant loops not filled. ACTION:

a. With less than the above required RHR loops OPERABLE, immediately initiate corrective action to return the required RHR loops to OPERA 8LE status as soon as possible.

b. With no RHR loop in operation. 'uspend all operations involving a reduction in boron concentration of the Reactor Coolant System and -

immediately initiate corrective action to return the required RHR loop to operation. SURVEILLANCE REQUIREMENTS O I 4.4.1.4.2 At least one RHR loop shall be determined to be in operation and , circulating reactor coolant at least once per 12 hours. l

\
        "One RHR loop may be inoperable for up to 2 hours for surveillance testing provided the other RHR loop is OPERABLE and in operation.
      **The RHR pump may be deenergized for up to 1 hour provided:               (1) no opera-tions are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10*F below saturation temperature.

COMANCHE PEAK - UNIT 1 3/4 4-7

                                                                       - - - - -      -~  ,

I IIX-88512 ATTACHMENT 7 ' I PAGE 15 0F 66 l

l [m)

%J
                  -*EAeT04-t00 TANT-SMTEM---                                                                                                                         nnR
                  -4/'.4.2                    ',AFETV-VALVM -                                                                                                                  ,

1 i MM'Wt 101 b i 1 41MITI'C CONDITION-FOR-OPERATIM-

3. 4. 2.1 A MiniRUS-Of ON! p"?S$UriIer C0de $8f9ty valua chall ha ODEDARf[ with a lif t-sett4ng of 2'"5 p;ig 1%."

A B D 1. T. P..A

                                     -t.D i t T. T. v .

nee v., A. ,. ,

                                                                                                ,A, ,E ,

A.P.T. T. A. M..

                  -W+th-no-presstreizer-Cobefety-velve-CPERABLE, imedtetely suspend ai; _

eperat4+ns-involv4ng-peshiv: caet M ty ch:nges ;nd piece en OPERABLE mi" W ato-opw ation-fr the Shet h e ::eling ::d:. \ (

                   ,...........,m                                               m .m.,

J V n Y 6 4 6 6fiab b nb rwrI6 315 J 4.4.2.1 "O ddR4cn:1 reqdrements--ether-than-thee; required by 4pec4f4catter 2.0.5.

                   "The lift-setting pressttre-sheH-eerrespond to ;;tient-conditions-of t': v:!ve
                   -- ,t n; .in:1 Oper:'Ag t:g:ratwe and pres:ure.

COMANCHE PEAK - UNIT 1 3/4 4-8 .

TIX-88512 ' ATTACMENT7 PA E 16 0F 66 _ - en REACTOR COOLANT SYSTEM . O W 4. 2. s anw vnws V f OPERATING LIMITING CON 0! TION FOR OPERATION 3.4.2// All pressurizer Code safety valves shall be OPERABLE with a lift setting of 2485 psig

1%.*

APPLICABILITY: MODES 1, 2, and 3. ACTION: With one pressurizer Code safety valve inoperable, either restore the inoperable valve to OPERABLE status within 15 minutes or be in at least HOT STANDBY within 6 hours and in at least H0! SHUTDOWN within the following 6 hours. SURVEILLANCE REQUIREMENTS , ___ e OI 4.4.2/JI No additional requirements other than those required by Specification 4.0.5.

                              "The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

O CCKANCHE PEAK - UNIT 1 3/4 4-9

IIX88512 ' i

  ,              AffACMENT 7 PAGE 17 0F 66
                                  ~

p p--

  "3            REACTOR COOLANT SYSTEM                                                      y ,,

(V 3/4.4.3 PRES $URIZER LIMITING CON 0! TION FOR OPERATION 3.4.3 The pressurizer shall be OPERABLE with a water volume of less than or equal to 1662 cubic feet (92% of span), and at least two groups of pressurizer heaterseachhavingacapacityofatleast[150/kW. APPLICABILIU: MODES 1, 2, and 3. ACTION:

a. With only one group of pressurizer heaters OPERABLE, restore at least two groups to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTOOWN within the following 6 hours.

b. With the pressurizer otherwise inoperable, be in at least HOT STANOBY with the Reactor Trip System breakers open within 6 hours and in HOT SHUTDOWN within the following 6 hours.

( l SURVEILLANCE REQUIREMENTS 4.4.3.1 The pressurizer water voluipe shall be determined to be within its limit at least once per 12 hours. 4.4.3.2 The capacity of each of the above required groups of pressurizer l heaters shall be verif8ed by energizing the heaters and measuring circuit i current at least once per 92 days. ( i I l l l COMANCHE PEAK - UNIT 1 3/4 4-10 ,

IXX-88512 ! AliACHM(NT? l PAGE 18 0F 66 l REACTOR COOLANT SYSTEM ()J

 'w 3/4.4.4 RELIEF VALVES LIMITING CON 0! TION FOR OPERA (10N 3.4.4 All power-operated relief valves (PORVs) and their associated block valves shall be OPERABLE.

APPLICABILITY: H0JES 1, 2, and 3. ACTION:

1. With one or more PORV(s) inoperable, because of excessive seat leak-age, within 1 hour either restore the PORV(s) to OPERABLE status or close the associated block valve (si; otherwise, be in at least HOT STANDBY within the next 6 hours and in (Ob0 SHUTDOWN within the fol-lowing M hours. HOT G

b. With one PORY inoperable due to causes otter than excessive seat leakage, within 1 hour either restore the PORV to OPERABLE status or close the associated block valve and remove power from the block valve; restore the PORV to OPERABLE status within the following 72 hours or be in HOT STANDBY within the next 6 hours and in 60te. 10 I: 0224 SHUT 00WNwithinthefollowingghours. HcT I c. With both PORV(s) inoperable due to causes other than excessive seat leakage, within 1 hour either restore each of the PORV(s) to OPERABLE status or close their associated block valve (s) and remove power from the block v4.1ve(s) and be in HOT STANDBY within the next 6 hours and in GOte- SHUT 00WN within the following M hours.

Her 4 t. e.-- 101: 0906

d. With one or more block valve (s) inoperable, within 1 hour (1) restore the block valve (s) to OPERABLE status or close the block valve (s) and remove power from the block valve (s); or close the PORV and remove power from its associated solenoid valve; and (2) apply ACTION b above, as appropriate, for the isolated PORV(s).

e. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS l 4.4.4.1 In addition to the requirements of Specification 4.0.5, each PORV shall be demonstrated OPERABLE at least once per 18 months by: l a. Operating the valve through one complete cycle of full travel, and

b. Performing a CHANNEL CALIBRATION of the actuation instrumentation.

Feil q rest.c Wm 4m oeAV, f.UA det;aa Saad

b. aiw, N

r. sp.in J .G Aei. %kud 6 b ..J h 4.= M +4 Za;l; l Poar war har d i.eper.Wt. -

l l COMANCHE PLAK - UNIT 1 3/4 4-11 ,

 - - . ..                                                         -          =              _ .   .

TXX-88512 I AfiACHMENT? l PAGE 19 0F 66 i

                                                                                        ~
     /'   REACTOR COOLANT SYSTEM 3/4.4.4 RELIEF VALVES                                                                 U SURVEILLANCE REQUIREMENTS 4.4.4.2 Each block valve shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel unless the block valve is closed in order to meet the requirements of ACTION a and b in Specification 3.4.4.

4 lO J 6 l t I i O COMANCHE PEAK - UNIT 1 3/4 4-12 ,

1 I TXX 88512 AliACHMENT7 ~ ' PAGE 20 0F 66

                       ~    '

N REACTOR COOLANT SYSTEM

                                                                                    ,W 4.4.5    STEAM GENERATORS D Nd LIM     NG CONDITION FOR OPERATION 3.4.5 Eac steam generator shall be OPERABLE.

APPLICA81 LIT MODES 1, 2, 3, and 4. ACTION: With one or more s am generators inoperable, restore the inoperable generator (s) to OPERABLE status ior to increasing T,yg above 200*F. SURVEILLANCE REQUIREMEN

4.4.5.0 Each steam generato shall be demonstrated OPERABLE by performance of the following augmented inserv e inspection program and the requirements of Specification 4.0.5.

4. 4. 5.1 Steam Generator Sample Se etion and Inspection - Each steam generatar shall be determined OPERABLE during hutcown by selecting and inspecting at

( least the minimum number of steam gen rators specified in Table 4.4-1. 4.4.5.2 Steam Generator Tube Sample Se etion and Inspection - The steam generator tuce minimum sample size, inspe ion result classification, and the corresponding action required shall be as ecified in Table 4.4-2. The inservice inspection of steam generator tube shall be performed at the fre-quencies specified in Specification 4.4.5.3 a the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 4.4.5.4. The tubes selected for each inservice inspection sha 1 include at least 3% of all the expanded tubes and at least 3% of the remaini number of tubes in all steam generators; the tubes selected for these insp ctions shall be selected on a random basis except:

a. Were experience in similar plants with simil water chemistry indicates critical areas to be inspected, then t least 50% of the tubes inspected shall be from these critical are s;

b. The first sample of tubes selected for each inserv inspection i

(subsequent to ^.he preservice inspection) of each s ta generator i shall include: 1 O \ \ l l COMANCHE PEAK - UNIT 1 3/4 4-1? ,

111-88512 ATTACHMENT 7 i PAii?!0F66 _ , REACTOR COOLANT SYSTEM (AM GENERATORS flN 7

                                                           >{'muJf61    d, s 09y s

x SURVLILLANCE REQUIREMENTS (Continued)

                      )     All nonplugged tubes that previously had detectable wall penetrations (greater than 20%),

2) Tubes in those areas where experience has indicated potential

       .                       roblems, and

3) A t be inspection (pursuant to Specification 4.4.5.4a.8) shall be p formed on each selected tube. If any selected tube does not p it the passage of the eddy current probe for a tube inspect on, this shall be recorded and an adjacent tube shall be selec d and subjected to a tube inspection.

c. Th6 tubes select as the second and third samples (if required by Table 4.4-2) duri each inservice inspection may be subjected to a partial tube inspec ion provided:

1) The tubes selecte for thest samples include the tubes from those areas of the ube sheet array where tubes with imperfections were p eviously found, and b

d 2) The inspections includ those portions of the tubes where I imperfections were previ sly found. The results of each sample inspection shal be classified into one of the following three categories: Category Inspec ion Results C-1 Less than 5% of the total tubes inspected'are degraded tubes and n e of the inspected tubes are defective. C-2 One or more tubes, but no more than 1% of the total tubes inspected are factive, or between 5% and 10% of the total tub inspected are degraded tubes. C-3 More than 10% of the total tubes inspected are degraded tubes or more than 1% of he inspected tubes are defective. Note: In all inspections, previously degraded tubes a t exhibit significant (greater than 10%) further wall penet sations to be included in the above percentage calculations o N COMANCHE PEAK - UNIT 1 3/4 4-14 ,

F TIX-88512 ATTACHMENT 7 PAbE 22 0F 66- - REACTOR COOLANT SYSTEM ST GENERATORS { g{" g om SURVE\LANCEREQUIREMENTS(Continued) 4.4.5.3 I pection Frequencies - The above required inservice inspections of steam gener or tubes shall be performed at the following frequencies:

a. The Jrst 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.

Subseque inservice inspections shall be performed at intervals of not less t an 12 nor more than 24 calendar months after the previous inspection. If two consecutive inspections, not including the preser-vice inspect n, result in all inspection results falling into the C-1 category or i tw consecutive inspections demonstrate that previously observed degrad ion 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 cesult: of the inservice inspection of a steam generator conducted f ri sceordanc with Table 4.4-2 at 40-month intervals fall in Category C-3, the in action frequency shall be increased to at fm least once per 20 months. The increase in inspection frequency Q ,

shall apply until the subs vent inspections satisfy the criteria of Specification 4.4.5.3a.; the\ interval may then be extended to a maximum of once per 40 months nd

c. Additional, unscheduled inservic inspections shall be performed on each steam generator in accordanc with thw first sample inspection specified in Table 4.4-2 during th shutcown subsequent to any of the following conditions:

1) Primary-to secondary tubes leak (ngt including leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.4.6.2, or

2) A seismic occurrence greater than the Operating Basis Earthquake, or

3) A loss-of-coolant accident requiring actuation of the Engineered Safety Features, or

4) A main steam line or feedwater line break., \ s l

COMANCHE PEAK - UNIT 1 3/4 4-15 ,

III 88512 ATTACHMENT 7 PAGE 23 0F 66

 /7    \  REACTOR C0OLANT SYSTEM Dgn gf 3 m EAM GENERATOR k h ] I!! 0 6
          $UR ILLANCE REQUIREMENTS (Continued) 4.4.5.4        cceptance Criteria

a. As used in this specification:

1) meerfection means an exception to the dimensions, finish, or c tour of a tube from that required by fabrication drawings or spe ifications. Eddy-current testing indications below 20% of the reinal tube wall thickness, if detectable, may be

    .                        consi red as imperfections;

2) Degracat n means a service-induced cracking, wastage, wear, or general co' osion occurring on either inside or outside of a tube;

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

4) % Dearadation means e percentage of the tube wall thickness affected or removed b degradation; I

5) Defect means an imperfect on of such severity that it exceeds the plugging limit. A tub containing a defect is defective;

6) Pluaaina Limit means the impe faction depth at or beyond which the tube shall be remove:1 from ervice and is equal to 40%*

of the nominal tube wall thickne ;

7) Unserviceable describes the conditi of a tube if it lea'ks or contains a defect large enough to af ct its structural integ-rity in the event of an Operating Basi Earthquake, a loss-of-coolant accident, or a steam line or fe ater line break as specified in Specification 4.4.5.3c., abo e; 1

i 8) Tube Inspection means an inspection of the steam generator tube l from the point of entry (hot leg side) comple ly around the U-bend to the top support of the cold leg; and i 1 "Value to be determined in accordance with the recommendations of Reg latory Guide 1.121, August 1976. I COMANCHE PEAK - UNIT 1 3/4 4-16

IXX69512 AffACHMENT7 PAGE 24 0F 66 O \EACTORCOOLANTSYSTEM v t STEM GENERATOR RE.'0CA"" W MU SURVE LANCE REQUIREMENTS (Continued)

                       \ 9      Preservice Inspection means an inspection of the full length of each tube in each steam generator performed by eddy current techniques prior to service to establish a baseline condition the tubing. This inspection shall be performed prior to in tial POWER OPERATION using the equipment and techniques exp ted to be used during subsequent inservice inspections.

b. The steam g erator shall be determined OPERABLE af ter completing the correspon ing actions (plug all tubes exceeding the plugging limit and all bes containing through-wall cracks) required by Table 4.4-2.

4.4.5.5 Reports

a. Within 15 days follow gg the completion of each inservice inspection' of steam generator tubet the number of tubes plugged in each steam generator shall be repor d to the Commission in a Special Report pursuant to Specification .9.2;

b. The complete results of the s eam generator tube inservice inspec-O I tionshallbesubmittedtothe3ommissioninaSpecialReportpur-suant to Specification 6.9.2 wit in 12 months following the completion of the inspection. Th Special Report shall include:

1) Number and extent of tubes insp ted,

2) Location and percent of wall-thick ss penetration for each indication of an imperfection, and

3) Identification of tubes plugged.

c. Results of steam generator tube inspections wh h fall into Category

C-3 shall be reported to the Commission pursuant to 10 CFR Part 50.72

                     ,_- within four hours of initial discovery, and pursu'eqt to Specifica-i                 G.9.2, tionTr-het within 30 days and prior to resumption ofs plant operation.

' Thisreportshallprovideadescriptionofinvestigat(onsconducted to determine cause of the tube degradation and corrective measures

      .                 taken to prevent recurrence, i
                                                                                        \s O

COMANCHE PEAK - UNIT 1 3/4 4-17 l l l

O _ O _ O - TABLE 4.4 1 NININUM NU MER OF STEAM GENERATORS TO BE 333 m; o ~9m g INSPECTED DURING INSERVICE INSPECTION {g$

 =                                                                                                                      =~

g , Preservice Inspection ro z , U No. of Steam Cenerators per Unit our e* First Inservice Iispection [ Two Second & Subsequent Inservice Inspections [ One8 TABLE NOTATIONS

o. e ; n emh ces pec k vdg.

z* 1. s 4hJ yerc Each ef-the other two steam generators not ins ted during the first inservice inspectiory( 9 shall be inspected during the second and thi inspectionsj for the fourth and subsequent

  "         inspections, the inservice inspection may       limited to one steam generator on a rotating schedule encompassing 12% of the tubes      the results of the-f4en or previous inspections d the kaor dC*m indicate that all steam generators a     performing in a like manner. Note that under some        T"* "

circumstances, the operating cond ions in one or more steam generators may be found to be more severe than those in other team generators. Under such c.ircumstances, the sample sequence shall be modified t inspect the most severe conditions. M D C"3 O h

                                                                                                                       ---(

m O E U ca

O O TABLE 4.4-2 O -

STEAM GENERATOR TU8E INSPECTION

;            o
                                                                                                                                      /,/ y *=2 e=
                                                                                                                                   /        MEB g                      IST SAMPLE INSPECTION                      2ND SAMPLE INSPECTION 3RDSAMPLEINpCTION         a3G
\            x gSample Size       Result      Action Required            Result                                               /               &~

i Action Required Result ActjenRequired m gA minimum of C-1 None N.A. N.A. N.A. ! N.A. j o 55 Tubes per 75.G. C-2 Plug defective tubes C-1 None N.A[ N.A. and inspect additional '

j $

a 25 tubes in this S.G. Plug defective tubes [ C-1 None i C-2 and inspect additional l 45 tubes in thly S.G. C-2 Plug defective tube: l

                                                                                                 /
                                                                                                   /

i Perform action for i / C-3 C-3 result of first

                                                                                        ,/                             sample 1

Perform action for l , C-3 C-3 result of first N.A. N.A. g f sample i C-3 Inspect all tubes in All other 5 this S.G., plug de- S.G.s ere None N.A. N.A. fective tubes and , C-1 inspect 25 tubes,in each other S. ./ Some S.G.s Perform action for N.A. N.A. m C-2 but no C-2 result of second m sample j Notification to NRC additional 5.G. are 5 cs pursuant to S50.72 (b)(2) of 10 CFR C-3 m

                                       , ' Part 50 Additional    Inspect all tubes in
                                    /                                  S.G. is       each S.G. and plug                              Q
                                                                                                                                              =
                                ,/                                     C-3           defective tubes.                                         g Notification to NRC    N.A.            N.A.             O,
                         '                                                           pursuant to 550.72 (b)(2) of 10 CFR Part 50 5=            Where n is the number of steam generators inspected during an inspection

IXXC512 ATTACHM(WI7 PAGE 27 0F R . REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.6.1 The following Reactor Coolant System Leakage Detection Systems shall be OPERABLE:

a. The Containment Atmosphere Particulate Radioactivity Monitoring System,

b. The Containment Sump Level and Flow Monitoring System, and

c. Either the containment air cooler condensate flow rate or the Con" tainment Atmosphere Gaseous Radioactivity Monitoring System.

APPLICA8ILITY: MODES 1, 2, 3, and 4. ACTION: With only two of the above required Leakage Detection Systems OPERABLE, y operation may continue for up to 30 days provided grab samples of the contain-ment atmosphere are obtained and analyzed at least once per 24 hours when the required Gaseous or Particulate Radioactive Monitoring System is inoperable; otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUT 00WN within the following 30 hours. The provisions of Specification 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS 4.4.6.1 The Leakage Detection Systems shall be demonstrated OPERABLE by:

a. Containment Atmosphere Gaseous and Particulate Monitoring Systems-performance of CHANNEL CHECK, CHANNEL CALIBRATION, and DIGITAL CHANNEL OPERATIONAL TEST at the frequencies specified in Table 4.3-3,

b. Containment Sump Level and Flow Monitoring System ~ performance of CHANNEL CALIBRATION at least once per 18 months, and

c. Containment Air Cooler Condensate Flow Rate Monitoring System -

performance of CHANNEL CALIBRATION at least once per 18 months. CCMANCHE PEAK - UNIT 1 3/4 4-P0 ,

in 88512 ' AffACHMiWI? PAGE 28 0F 66. .

                         .                                                                     n

() REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE L I LIMITING CONDITION FOR OPERATION 4 3.4.6.2 Reactor Coolant System leakage shall be limited to: No PRESSURE BOUNDARY LEAKAGE, ! a.

b. 1 GPM UNIDENTIFIED LEAKAGE,

c. 1 GPM total reactor-to-secondary leakage through all steam generators not isolated from the Reactor Coolant System and 500 gallons per day through any one steam generator not isolated from the Reactor Coolant System,

d. 10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System,

e. 40 GPM CONTROLLED LEAKAGE at a Reactor Coolant System pressurs of 2235 1 20 psig, and

f. 0.5 GPM leakage per nominal inch of valve size up to a maximum of 10 1: 0230 5 GPM at a Reactor Coolant System pressure of 2235 2 20*psig from any l Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1.

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

a. With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANOBY within 6 hours and in COLD SHUTDOWN within the following 30 hours,

b. With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE and leakage from Reactor Coolant System Pressure Isolation Valves, reduce the leakage rate to within limits within 4 hours or be in at least HOT STANOBY within the next 6 hours and in COLD SHUT 90WN within the following

    .                      30 hours.

c. With any Reactor Coolant System Pressure Isolation Valve leakage greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours by use of at least two closed manual or deactivated automatic valves, or be in at least HOT STANOBY within the next 6 hours and in COLD SHUT 00WN within the following 30 hours.

         #r T e u v esso r e s            less Nn 2.2.35                  bd      gruter bn i60 psig o. c e a.\\ o we d dec V s\p                  ge5g d <rc QOM         k G od. +o O m'. w e hi ghf.c prosof  e c

ow su na. te. a o .ge s u a ne \ a c k a.c e, z b\ sa.w nc\ c yc n s ,

            "to 2. 7. D T5 3 o.%um                        be  aa   p + ea,            ahi % ,,

m +. a COMANCHE PEAK - UNIT 1 s -s%m o *%g% 3/4 4-21 hk

                                                                                   -b b e

_&..sd eP (u-1

in 88512 ' AITACHM(Ni7 FAGE290F66 _ _ REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE SURVEILLANCE REQUIREMENTS 4.4.6.2.1 Reactor Coolant System lerkages shall be demonstrated to be within each of the above limits by:

a. Monitoring the Reactor Coolant System Leakage Detection System required by Specification 3.4.6.1 at least once per 12 hours;

b. Measurement of the CONTROLLED LEAKAGE to the reactor coolant pump seals when the Reactor Coolant System pressure is 2235 1 20 psig at 10 8: 0233 least once per 31 d y: wRh-th: ::dulating valve-fully ^aea_ + The '

provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4; c. r o x e_ r A 10 h 0234 Perf ormanc e- o f- e- R e ac t o e-C oo ient-Gys tessea te-4*ventory-be14ac+-4t le :t enee per 72 hour ; and i

d. Monitoring the Reactor Head Flange Leakoff System at least once per :

24 hours. l A 4.4.6.2.2 Each Reactor Coolant System Pressure Isolation Valve specified in i () Table 3.4-1 shall be demonstrated OPERABLE by verifying leakage to be within its limit:

a. At least once per 18 months,

b. Prior to entering MODE 2 whenever the plant has been in COLD SHUTDOWN for 72 hours or more and if leakage testing has not been performed in the previous 9 months, except for valves 8701A, 87018, 8702A, and 87028.

c. Prior to returning the valve to service following maintenance, repair or replacement work on the valve, and

d. FolSowing check valve actuation due to flow through the valve,

e. As outlined in the ASME Code, Section XI, paragraph IWV-3427(b).

The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4. ig en onW o c b b w.o q a. e e.A % W k n.kI c ba. A gtbcm c & 1 ; e.c d i m , m a.i norh M l,M . C t e n.\ go. A -qe c( h R (m

c. M. oA \eA st. o n c e. pc S\ Aq V o.\ V e s 9%9 A , 3, a n d. b o. r e Q <ac;I g; %g !

J C.

                                                                      \ t t k e.i.         gej gg

p. u en . sn 1

     "T u c,    he iq      C     gt b k)g \tS % th ei n             3   E [ hedr.

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

F. , TrX C 512 ! AttacMMilli 7 . PAGE 30 0F M i INSERT A l Performance of a Reactor Coolant System water inventory balance within 12 hours after achieving steady state operation

and at least once per 72 hours i thereafter during steady state operation, except that no more than 96 hours j shall elapse between any two successive inventory balances. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4; and ,

i 8 , f l h e O 1 i O l l l

TXX-88512 ATTADWO(T 7

FAGE 31 0F 66 !

                          .     -                                                                  l TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES VALVE NUMBER                      FUNCTION 8948 A, B, C, O            Accumulator Tank Discharge 8956 A, B, C, D'           Accumulator Tank Discharge 8905 A, B, C, O            SI Hot leg Injection 8949 A, B, C, O            SI hot Leg Injection 8818 A, B, C, O            RHR Cold Leg Injection 8819 A, B, C, O            SI Cold leg Injection 8701 A, B                  RHR Suction Isolation 8702 A, B                                                        g RHR Suction Isolation
                            --C705 A, S                   Rl? Section Is;1ation R;44ef-   1080938 g

8841 A, B RHR Hot Leg Injection 8815 CCP Cold Leg Injection 8900 A, B, C, O CCP Cold Leg Injection I

 \~,) t 4

COMANCHE PEAK - UNIT 1 3/4 4-23 ,

TXX-88512 AlikCHMENT) PAGE 32 of 66~ REACTOR COOLANT SYSTEM M 3/4.4.7 CHEMISTRY a LIMITING CONDITION FOR OPERATION 3.4.7 The Reactor Coolant System chemistry shall be maintained within the limits specified in Table 3.4-2. APPLICA8ILITY: At all times. ACTION: MODES 1, 2, 3, and 4:

a. With any one or more chemistry parameter in excess of its Steady-State Limit but within its Transient Limit, restore the parameter to within its Steady-State Limit within 24 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours; and i

b. With any one or more chemistry parameter in excess of its Transient Limit, be in at least HOT STANDBY within 6 hours and in COLD SHUT 00WN within the following 30 hours.

Ai 'll Other Times: With the concentration of either chloride or flucride in the Reactor Coolant Systas in excess of its Steady-State Limit for more than 24 hours i or in excess of its Transient Limit, reduce the pressurizer pressure to less than or equal to 500 psig, if applicable, and perform an engineering l evaluation to determine the effects of the out-of-limit condition on the l structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation p'rior to increasing the pressurizer pressure above 500 psig or prior to proceeding to MODE 4. SURVEILLANCE REQUIREMENTS l 4.4.7 The Reactor Coolant System chemistry shall be determined to be within the limits by analysis of those parameters specified in Table 3.4-2 at least once per 72 hours." l -- i O O *SampleandanalgsisfordissolvedoxygenisnotrequiredwithT**9 less than l or equal to 250 F. i COMANCHE PEAK - UNIT 1 3/4 4-24 ,

h1512 . ATTACHMENT 7 PAGE 33 0F 66 e TABLE 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS STEADY-STATE TRANSIENT PARAMETER LIMIT LIMIT Dissolved Oxygen

1,0.10 ppm i 1.00 ppm Chloride 5,0.15 ppm i 1.50 ppm Fluoride 1 0.15 ppm i 1.50 ppm i

r

Limit not applicable with T,yg less than or equal to 250'F.

O COMANCHE PEAK - UNIT 1 3/4 4-25 ,

TXX-88512- ' i ATTACHMENT 7 PAGE 34 0F 66 ( REACTOR COOLANT SYSTEM 3/4.4.8 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.8 The specific activity of the reactor coolant shall be limited to:

a. Less than or equal to 1 microcurie per gram OOSE EQUIVALENT I-131,

        .          and

b. Less th or equal to 1004 microCuries per gram of gross radioactivity.

APPLICABILITY: MODES 1, 2, 3, 4, and 5. ACTION: MODES 1, 2 and 3*:

a. With the specific activity of the reactor coolant greater than 1 microcurie per gram DOSE EQUIVALENT I-131 for more than 48 hours during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in at least HOT STANDBY with T less than 500'F within 6 hours; and avg

b. With the specific activity of the reactor coolant greater than I 100 4 microcuries per gram, be in at least HOT STANDBY with T less avg than 500*F within 6 hours.

l MODES 1, 2, 3, 4, and 5: With the specific activity of the reactor coolant greater than 1 microcurie per gram DOSE EQUIVALENT I-131 or greater than 1004 micro-Curies per gram, perform the sampling and analysis requirements of item 4.a) [ of Table 4.4-4 until the specific activity of the reactor coolant is l restorea to within its limits. i SURVEILLANCE REQUIREMENTS l 4.4.8 The specific activity of the reactor coolant shall be determined to be within the limits by performance of the sampling and analysis program of l Table 4.4-4. l l l

     *With T,yg greater than or equal to 500*F.

COMANCHE PEAK - UNIT 1 3/4 4-26 .

TH 88512 , ATTACHMENT 7 PAGE 35 0F 66 l

                                                                            .I O(

l p FIGURE 3.4-1 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERWil POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY >l pCi/ gram DOSE EQUIVALENT I-131 COMANCHE PEAK - UNIT 1 3/4 4-27 ,

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

R f U d b ' TABLE 4.4-5 ;=- i 8 9EACTOR COOLANT SPECIFIC ACTIVITY SAMPLE *E2 J

n r AND ANALYSIS PROGRAM [hk

m E" u TYPE OF MEASUREMENT SAMPLE AND ANALYSIS MODES IN nAilCH SAMPLE 9

x AND ANALYSIS FREQUENCY AND ANALYSIS REQUIRED

    . 1. Gross Radioactivity                      At least once per 72 hours.       1,2,3,4 g           Determination                                                                                       '

M 2. Isotopic Analysis for DOSE EQUIVA- 1 per 14 days. 1

  -          LENT I-131 Concentration

3. Radiochemical f er E Determination

1 per 6 months ** 1

4. Isotopic Analysis for Iodine a) Once per 4 hours, l#, 2#, 3#, 4#, 5# -

Including I-131, I-133, and I-135 whenever tre specific activity exceeds 1 pCi/ gram DOSE - w EQUIVALENT I-131 D or 100/E pCi/ gram of 9 gross radioactivity, and S$ b) One sample between 2 .1, 2, 3 and 6 hours following a THERMAL POWER change exceeding 15% of the RATED THERMAL POWER within a 1-hour period. b 21

TXX 88512 ATTACHMENT 7 PAGE 37 0F 66 j [ 3 TABLE 4.4-t (Continued) h TABLE NOTATIONS

                    *A radiochemical analysis for                                    shall consist of the quantitative measurement of the specific activity for each radionuclide, except for radionuclides with half-lives less than 10 minutes and all radiciodires, which is identified in the reactor coolant. The sp'ecific activities fgr these individual radio-nuclides shall be used in the determination of E for the reactor coolant sample. Determination of the contributors to E shall be based upon those energy peaks identifiable with a 95% confidence level.
                  ** Sample to be taken after a minimum of 2 EFP0 and 20 days of POWER OPERATION have elapsed since reactor was last subcritical for 48 hours or longer.
                   #Until the specific activity of the Reactor Coolant System is restored within its limits.

O t e9 O COMANCHE PEAK - UNIT 1 3/4 4-29 ,

TXX-88512 ATTACHMG i 7 PAGE 38 0F 66 . j REACTOR COOLANT SYSTEM

                                                                                    ,at 3/4.4.9 PRESSURE / TEMPERATURE LIMITS                                                         l REACTOR COOLANT SYSTEM                                                                        I
                                                                                                   )

LIMITING CONDITION FOR OPERATION i 3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figures 3.4-2 and 3.4-3 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:

a. A maximum heatup of 100'F in any 1-hour period,

b. A maximum cooldown of 100*F in any 1-hour period, and

c. A maximum temperature change of less than or equal to 10'F in any 1-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.

APPLICABILITY: At all times. ACTION:

 , With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation or be in at least HOT STANOBY within the next 6 hours and reduce the RCS T,yg and pressure to less than 200*F and 500 psig, respectively, within the following 30 hours.

SURVEILLANCE REQUIFEMENTS 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operations.

4. 4. ^ .1. 2 Th: r:::ter v::::1 ::teri:1 der:di: tion :urvei'!:n : :pecimens ID 8: 0237 sheli b r==:d :nd ::::in:d, te dete--ine :h:nge: 4- e:teri:1 pr perti g, a: r: uir:d by 10 CF" P:rt 50, ^.ppendix 4, i :::Ord:nce with the : hedule -

in T;bi; 4.4-5. Th; result; cf th;;; ;xeminations ; hell be used te ;-date Ti ure; 3. 4-2 :M 2. '-3. O COMANCHE PEAK - UNIT 1 3/4 4-30 ,

 ..e...        . ., , . . . . . - _ - . - .              __- -  -.--.--.......-~.--.. - . -              -.

, TXX-80$lt . ., ,

            .                             ATTACHMfNT 7 -                                                                         !
                                      ' PAGE 39 0F.66--

t

                                                                                                                               .4
                                                                                                                               .t o                                                                                                                                 .

t 1 ) i i i li PRINT-READY FIGURE

TO BE PROVIDED LATER r ,

i P i I 6 i

i r

I l l P t

                                                                                                                                 ?
                                                                                                                                 ?

l FIGURE 3.4-2 l

                                               ' REACTOR COOLANT SYSTEM HEATUP LIMITATIONS - APPLICABLE UP TO      EFPY f

4 l 1 COMANCHE PEAK - UNIT 1 3/4 4-31 , i

W T'FF . - . . w* 't gm MN mewn_ _ . _

IXX-88512 ! ' i

        +
                . ATTACHMENT 7:                i' PAGE 40 0F 66               j
      @                                                                             DRAFT 1

i ! PRINT-READY FIGURE l TO BE PROVIDED LATER .k r n i i FIGURE 3.4-3 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS - APPLICABLE UP TO EFPY COMANCNE PEAK - UNIT 1 3/4 4-32 , l l b

IXX-88512 !- 8 AITACHMENT 7 ! PAGE 41 0F 66 I 1 W l O ' _ .,CI

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COMANCHE PEAK - UNIT 1 3/4 4-33 .

l l TXX-88512 .

       .               ATTACHMENT 1 PAGE 42 0F 66-l REACTOR COOLANT SYSTEM                                                                                                                                                j PRESSURIZER LIMITING CONDITION FOR OPERATION                                                                                                                                      ;

i 3.4.9.2 The pressurizer temperature shall be limited to:

a. A maximum heatup of 100*F in any 1-hour period, an(L

b. A maximum cooldown of 200*F in any 1-hour periodf 4a4 .

c. ^ =i:= pr:y .;;t;r t;;perature dif ferer,tisi cf 020*i. 10 1: 0238 APPLICABILITY: At all times.

ACTION: With the pressurizer temperature limits in excess of any of the above limits, restore the temperature to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the pressurizer; determine that the pressurizer remains acceptable for continued operation or be in at least HOT STAN0BY within the next 6 hours and reduce the pressurizer pressure to less than 500 psig within the following 30 hours, o SURVEILLANCE REQUIREMENTS 4.4.9.2 The pressurizer temperatures shall be determined to be within the limits at least once per 30 minutes during system heatup or cooldown. De- 10 : 0240

                    ,m..o    m.+.. + _ ........ a4<<....                                4.i    .w ,i w.               a... 4m.a +- w. m4+w4-                           +w.

1 b. _ . +.. . . E. h. . . . . . . . r.. (..h. k. . .. ... .. ... 4,..m, ..-m k,1 2. . . ,.,r..,.. .r . . . E. 2. . . . . O COMANCHE PEAK - UNIT 1 3/4 4-34 ,

IIX88512 ATTACHMENT 7 PAGE 43 0F 66

 /m     i                                                                                                                    F'T REACTOR COOLANT SYSTEM V

i OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION

b. Lo resdoal 'hea.t r e m ooc \ L RR R 3 soc.H on n.\ '. eh vah e.s so.ch w a h c. 9t N d oI 45o 3 si g I 3 ), 8, o y.

3.4.9.3 At least one of the following Overpressure Protection Systems shall be OPERABLE:

a. Two power-operated relief valves (PORVs) with lif t settings which vary with RCS temperature and which do not exceed the limits established in Figure 3.4-4, or ID I: 0242 c.4h The Reactor Coolant System (RCS) depressurized with an RCS vent of greater than or equal to 2.98 square inches.

APPLICABILITY: MODE 4, MODE 5 and MODE 6 with the reactor vessel head on. 0 0244 oc one rey;<ed RHR sock t on cettel elve ACTION: rego;<ep e.dke r +wo ?cRVs of ho RRR so chon <*

a. With one PORV inoperable,Vrestore the inoper:M : PORV to OPERABLE status within 7 days or depressurize and vent the RCS thr:;gh :t 1:::t : 2.98 :qu:re inch v:nt3 within the next 8 hours.

   ]                                                       a.s spec; t;eA in speca;c a;ow 3 4.g . 3 c. , a bo9 e ,

1 i (d

b. With beth PORV5 inepei eble, depi e55wrize end went trie ES trirewgh et e 1e:St 2.99 :qu:r: inch vent within 0 hsw,..

the KRR soeben ceuel va\xs,

c. IntheeventeitherthePORVsfortheRCSvent(s)areusedtomitigate

an RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days. The report shall describe the circumstances initiating the transient, the effect of the PORVs or RCS vent (s) on the transient, l and any corrective action necessar o prevent recurrence.

l ne RRR socbon cellei va\ ve s ,

d. The provisions of Specification 3.0.4 are not applicable.

l t.c i W bou yegot <cd. TcRh 6.nA bdh cegoi red. DR socAton reh e.E n \ qc.s W 9co.bie , cAe_presso c ite a nd

          .                       V eM the R C.S                          o.s s ge.c.Niel    (n     3.4 ,q ,3 ,, , c,        ,'

w% B boars. i b COMANCHE PEAK - UNIT 1 3/4 4-35 .

            - ,.              y     _ --      v- - , , - .

IXX 88512

      ' ATTACHMENT 7-PAGE 44 0F 66 PRINT-READY FIGURE TO BE PROVIDE 0 LATER 1

l L@ l i l FIGURE 3.4-4 PORT SETPOINTS FOR OVERPRESSURE MITIGATION-APPLICABLE UP TO 10 EFPY COMANCHE PEAK-UNIT 1 3/4 4-35a

IXX 88512 AllACHMEHi7 PAGE 45 0F 66 REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEM f SURVEILLANCE REQUIREMENTS 4.4.9.3.1 Each PORV shall be demonstrated OPERABLE by:

a. Performance of an ANALOG CHANNEL OPERATIONAL TEST on the PORV actuation channel, but excludit,J valve operation, within 31 days prior to entering a condition in which the PORV is required OPERABLE and at least once per 31 days thereafter when the PORV is required OPERABLE;

b. Performance of a CHANNEL CALIBRATION on the PORV actuation channel at least once per 18 months; and

c. Verifying the PORV isolation valve is open at least once pe. 72 hours when the PORV is being used for overpressure protection.

2WSERT h in h 0245 4.4.9.3.73The RCS vent (s) shall be verified to be open at least once per 12 hours

when the vent (s) is being used for overpressure protection,

, l

             *Except whe.1 the vent pathway is provided with a valve which is locked, sealed, l              or otherwise secured in the open position, then verify these valves open at least once per 31 days.

l l

l O COMANCHE PEAK - UNIT 1 3/4 4-36 ,

IXX-88512 ATTACHMENT 7 PAGE 46 0F 66 l ~Ci' INSERT 8 U 4.4.9.3.2 Each RHR suction relief valve shall be demonstrated OPERABLE when the RHR suction relief valves are being used for cold overpressure protection as follows:

a. For RHR suction relief valve 87088

1) By verifying at least once per 31 days that RHR RCS Suction Isolation valve (RRSIV) 87018 is open with power tc the valve operator removed, and

2) By verifying at least once per 12 hours that RRSIV 87028 is open.

b. For RHR suction relief valve 8708A:

1) By verifying at least once per 31 days that RRSIV 8702A is open with power to the valve operator removed, and

2) By verifying at least once per 12 hours tnat RRSIV 8701A is open.

c. Testing pursuant to Specification 4.0.5.

O

I TXX-88512 .

 . ATTACHMENT 7
                                           )                                                 i PAGE 47 0F 66                                                                           l
                                           }'                                                l

_ . i REACT 0R COOLANT-SYSTEM 3/4.4.10 STRUCTURAL INTEGRITY s... . LIMITING CONDITION FOR OPERATION l 3.4.10 The structural integrity of ASME Code Class 1, 2, and 3 components shall be maintained in accordance with Specification 4.4.10. APPLICABILITY: All MODES. ACTION:

a. With the structural integrity of any ASME Code Class 1 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature more than 50*F above the minimum temperature required by NOT considerations.

b. With the structural integrity of any ASME Code Class 2 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolatt the affected component (s) prior to increasing the Reactor Coolant System temperature above 200*F.

l l O, c. With the structural integrity of any ASME Code Class 3 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or irsolate the affected component (s) from service.

d. The provisions of Specificasion 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.4.10 In addition to the requirements of Specification 4.0.5, each reactor coolant pump flywheel shall be inspected per the recommendations of Regulatory Position C.4.b of Regulatory Guide 1.14, Revision 1, August 1975. i l i O COMANCHE PEAK - UNIT 1 3/4 4-37 L

TIX-88512 . AliACHMENT7 PAGE 48 0F 66 REACTOR COOLANT SYSTEM 3/4.4.11 REACTOR COOLANT SYSTEM VENTS LIMITING CONDITION FOR OPERATION 3.4.11 At least one Reactor Coolant System vent path consisting of two vent valves in series powered from emergency busses shall be OPERABLE and closed at 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 path to OPERABLE status within 30 days, or, be in HOT STANOBY within 6 hours and in COLD SHUT 00WN within the follow-

 @                  ing 30 hours.

b. With both Reactor Coolant 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 at least one of the vent paths to OPERABLE status within 72 hours or be in HOT STANOBY within 6 hours and in COLD SHUT 00WN within the fol-lowing 30 hours.

l SURVEILLANCE REQUIREMENTS l 4.4.11./ 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,

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

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

rO V COMANCHE PEAK - UNIT 1 3/4 4-38 ,

IXX-88512 .

  .      AliACMENT 7 PAGE 49 0F 66,     ,
                     ~    ~

od 3/4.4 REACTOR COOLANT SYSTEM DRMT BASES 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with all reactor coolant loops in operation and maintain DNBR above 1.30 during all normal operations and anticipated transients. In MODES 1 and 2 with one reactor coolant loop not in operation this specification requires that the plant be in at least HOT STANOBY within 6 hours. In MODE 3, two riactor coolant loops provide sufficient heat removal capability for removing core decay heat even in the event of a bank withdrawal accident; however, a single reactor coolant loop provides sufficient heat removal capacity if a bank withdrawal accident can be prevented, i.e., by opening the Reactor Trip System breakers. Single failure considerations require that two loops be OPERABLE at all times. In MODES 3, 4, and 5, the operability of the required steam generators is based on maintaining a sufficient level to guarantee tube coverage to assure heat transfer capability. In MODE 4, and in MODE 5 with reactor coolant loops filled, a single l reactor coolant loop or RHR loop provides sufficient heat removal capability I for removing decay heat; but single failure considerations require that at least two loops (either RHR or RCS) be OPERABLE. In MODE 5 with reactor coolant loops not filled, a single RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations, and the unavailability of the steam generators as a heat removing component, require that at least two RHR loops be OPERABLE. The operation of one reactor coolant pump (RCP) or one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System. The reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognition and control. l The restrictions on starting an RCP with one or more RCS cold legs less I than or equal to 350*F are provided to prevent RCS pressure transients, caused by energy additions from the Secondary Coolant System, which could exceed the limits of 10 CFR 50 Appendix G. The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 50' above each of the RCS cold leg temperatures. 3/4.4.2 SAFETY VALVES I l The pressurizer Code safety valves operate to prevent the RCS from being ! pressurized above its Safety Limit of 2735 psig. Each afety valve is designed i to relieve 420,000 lbs per hour of saturated steam at the valve Setpoint. The-V -relief- capacity-of-a-single-safety valve is-adequate-to-ceMeve =y overpec :urc 10 0 000 COMANCHE PEAK - UNIT 1 B 3/4 4-1 , I - - _- _ - _ . - -- - . _ _ _ _ -- . - .

I TXX-88',12 -

   .         ATIACHMENT7                                                                  .

PAGE 50 Of 66 REACTOR COOLANT SYSTEM BASES h.. SAFEf4 VALM t (Cod'moeO 9EACTOR C00LANT LOOPS ^NO COOLANT CIRCULATION (Continued) 108:0223 c0ndition which-could Occur during shutdcen. In th: event that n :sfety valves _a re -OS E RASLE r-an-ope ra t i+ R HR-l ooprconnectsd to the RCS, provides

           -overpressure--relief-capablitty-and-wt14. prevent P.CS cverpre::uri:atier During operation, all pressurizer Code safety valves must be OPERABLE to prevent the RCS from being pressurized above its Safety Limit of 2735 psig.

The combined relief capacity of all of these valves is greater than the maximum surge rate resulting from a complete loss-of-load assuming no Reactor trip until the first Reactor Trip System Trip Setpcint is reached (i.e. , no credit is taken for a direct Reactor trip on the loss-of-load) and also assuming no operation of the power-operated relief valves or steam dump valves. Demonstration of the safety valves' lift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Code. 3/4.4.3 PRESSURIZER l The maximum water volume-also ensures that a steam bubble is formed and l thus the RCS is not a hydraulically solid system. The 12-hour periodic surveil-lance is sufficient to ensure that the parameter is restored to within its limit following expected transient operation. The requirement that a minimum l number of pressurizer heaters be OPERABLE enhances the capability of the plant i to control Reactor Coolant System pressure and establish natural circulation. Pressurizer heater groups are powered from sources that meet the requirements of Item II.E. 3.1 of NUREG-0737. l 1 ( l O COMANCHE PEAK - UNIT 1 B 3/4 4-2 ,

TXX 88512 1 ATTACHMENT 7 FAGE 51 0F W REACTOR COOLANT SYSTEM BASES l 3/4.4.4 RELIEF VALVES The power-operated relief valves (PORVs) and steam bubble function to relieve RCS pressure during all design transients up to and including the design step load decrease with steam dump. Operation of the PORVs minimizes t'he undesirable opening of the spring-loaded pressurizer Code safety valves. Each PORV has a remotely operated block valve to provide a positive shutoff capability should a relief valve become inoperable. 3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be main-tained. The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1. Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion. Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause l of any tube degradation so that corrective measures can be taken. Selected tubes in the preheater section of each 04 and 05 steam generator have been modified to correct the tube vibration degradation phenomenon experi-enced by certain Westinghouse steam generators. The modification consisted of expanding these tubes in the vicinity of the support plates and is designed to limit the amplitude of vibration. These expanded tubes are subject to a

     .special inspection whenever the steam generators are opened for inservice eddy current testing.                                                           .

The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes. If the secondary coolant chemistry is not maintained within these limits, local; zed corrosion may likely result in stress corrosion cracking. The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the Reactor Coolant System and the Secondary Coolant System (reactor-to-secondary leakage = 500 gallons per day per steam generator and a total ( leakage of 1 GPM to all steam generators). Cracks having a reactor-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents. Operating plants have demonstrated that reactor-to-secondary leakage of 500 gallons per day per steam generator can readily be detected by radiation monitors of steam generator blowdown. Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, O during which the leaking tubes will be located and plugged. COMANCHE PEAK - UNIT 1 B 3/4 4-3 .

TXX-88512 ATTACMENT 7 PAGE 52 0F 66 - - i ( REACTOR COOLANT SYSTEM BASES L-STEAM GENERATORS (Continued) z secondary coolant. Wastage-type defects are unlikely with proper will be found during scheduled inservice steam geneHo ce, it Plugging will be required for all tubes withrator imtube examinations. plugging limit of 40% of the tube nominal wall thi kperfections ex tube inspections of operating plants have demo c ness. Steam generator reliably detect degradation that has penetratec 20%nstrated the ca thickness. of the original tube wall fall into Category C-3, these results will beWhenever ng inservice inspection Special and Report pursuant to S pursuant to 10 CFR 50.72 within 4 hreported to the plant operation. pecification 6.9.2 within 30 days and prior to Such cases will be considered by the Commission o case tests, tions, basisadditional and may result eddy- in a requirement for analysis n a case-b y-Specifications, if necessary. current inspection, and revision ofcalthe T 3/4.4.6 3/4.4.6.1 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS

   '              provided to monitor and detect leakage boundary.

Regulator specification arefrom the These Detection Systems are consistent or coolant withpressure the rec Systems,"y Guide 1.45, "Reactor Coolant ommendations May 1973. PressureofBound for restoration since two diverse, 30and remain OPERA 8LE days areredundant permitted R ment gaseous or. particulate monitoring system, grab as a backup to the single remaining atmospheric n-monitoring ormed s s 3/4.4.8.2 ystem. OPERATIONAL LEAKAGE be indicative of an impendingressure gross ceptable sincefailure it may of the the presence of any PRESSURE BOUNDARY LEAXAGE placed in COLD SHUT 00WN. boundary. reqfore, There i u res the unit to be promptly expected from the RCS, the unidentified to a threshold value of less than 1 gpm . of leakage is e can be reduced port low to ensure early detection of. additional leakageThis threshold The total steam generator tube leakage limit of 1 gpm f erators not isolated from the RCS ensures that tube leakage will be limited to a small fraction of 10 or allthe r steamdosage gen-CF from the cont ution line values in the event of either a steam generator t b R 44 A 100 d break. of these accidents.The 1 gpm limit is consistent with the assumptio O steam rupture generator or under tube integrity is maintained in the even LOCA conditions. n the analysis main steam line COMANCHE PEAK - UNIT 1 B 3/4 4-4

10 88512

     ,              ATTACHMENT 7 fAGE $3 of 66 REACTOR COOLANT SYSTEM BASES OPERATIONAL LEAKAGE (Continued)

The 10 gpm IDENTIFIED LEAKAGE limitation provides allowance for a linited amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the Leakage Detection Systems. The CONTROLLEO LEAKAGE limitation restricts operation when the total flow supplied to the reactor coolant pump seals exceeds 40 gpm with the modulating (Sev.m) valve sin the supply line fully operkat a nominal RCS pressure of 2235 psig. This limitation ensures that in the event of a LOCA, the safety injection f1 w will not be less than assumed 3rvthe safety analyses. , 10 :: 04n The leakage from any RCS pressure isolation valve is sufficiently low to ensure early detection of possible in-series valve failure. It is apparent that when pressure isolation is provided by two in-series valves and when failure of one valve in the pair can go undetected for a substantial length of time, verification of valve integrity is required. Since these valves are important in preventing overpressurization and rupture of the ECCS low pressure piping which could result in a LOCA, these valves should be tested periodically to ensure low probability of gross failure. The Surreillance Requirements for RCS pressure isolation valves provide O added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA. Leakage from the RCS pressurt isolation valve is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit. 3/4.4.7 CHEMISTRY The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reactor Coolant System leakage or failure due to stress corrosion. Maintaining the chemistry within the Steady-State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System , over the life of the plant. The associated effects of exceeding the oxygen, chloride, and fluoride limits are time and temperature dependent. Corrosion l i studies show that operation may be continued with contaminant concentration l levels in excess of the Steady-State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System. The time interval permitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concen-trations to within the Steady-State Limits. I The Surveillance Requirements provide adequate assurance that concentrations ( in excess of ths limits will be detected in sufficient time to take corrective action.

                ) noena l le % F 1m u d chaq:}\oa leem one CC A C en 4rotIc d by HCV'I1L lo pchicA p<ssu r/ae r Iwei COMANCHE PEAK - UNIT 1                  8 3/4 4-5      ,

L .- _ _ _ _ . - _ _ _. . ~ . . _ _. _. _ _._.__ _ _ . . _ _ . _ _ _ __.

r TXX-88512

     ,        ATTACHMENT 7 PAGE540Fy       ,

REACTOR COOLANT SYSTEM BASES 3/4.4.8 SPECIFIC ACTIVITY The limitations on the specific activity of the reactor coolant ensure that the resulting 2-hour doses at the EXCLUSION AREA BOUNDARY (EAB) will not exceed an appropriately small fraction of 10 CFR 100 dose guideline values following a steam generator tube rupture accident in conjunction with an assumed steady- state reactor-to-secondary steam generator leakage rate of 1 gpm. The values for the limits on specific activity represent limits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative in that specific site parameters of the CPSES site, such as EAB location and meteorological conditions, were not considered in this evaluation. The ACTION statement permitting POWER OPERATION to continue for limited time periods with the reactor coolant's specific activity greater than 1 microcurie / gram DOSE EQUIVALENT I-131, but within the allowable limit shown on Figure 3.4-1, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER. The sample analysis for determining the gross specific activity and I can l O exclude the radioiodines because of the low reactor coolant limit of 1 microcurie / gram DOSE EQUIVALENT I-131, and because, if the limit is exceeded, the radio-iodine level is to be determined every 4 hours. If the gross specific activity level and radioiodine level in the reactor coolant were at their limits, the ! radioiodine contribution would be approximately 1%. In a release of reactor l coolant with a typical mixture of radioactivity, the actual radioiodine contri-

bution would probably be about 201. The exclusion of radionuclides with l half-lives less than 10 minutes from these determinations has been made for l several reasons. The first consideration is the difficulty to identify short-

! lived radionuclides in a sample that requires a significant time to collect, transport, and analyze. The second consideration is the predictable delay time between the postulated release of radioactivity from the reactor coolant to its relea:e to the environment and transport to the EAB, which is relatable to at a least 30 minutes decay time. The choice of 10 minutes for the half-life cutoff was made because of the nuclear characteristics of the typical reactor coolant radioactivity. The radionuclides in the typical reactor coolant have half-li' es of less than 4 minutes or half-lives of greater than 14 minutes, which allows a distinction between the radionuclides above and below a half-life of 10 minutes. For these reasons the radionuclides that are excluded from consideration are expected to decay to very low levels before they could be transported from the reactor coolant to the EAB under any accident condition. l The activity levels allowed by Figure 3.4-1 increase the 2-hour thyroid dose at the EAB by a factea of up to 20 following a postulated steam generator tube rupture. Therefore, operation with specific activity levels exceeding the limits of Specification 3.4.8 requires additional sampling per Table 4.4-4 and l l \ reporting of operational and sample information in the Annual Report pursuant to l Specification 6.9.1.4. This is in conformance with Generic Letter 85-19 to allow NRC evaluation . COMANCHE PEAK - UNIT 1 B 3/4 4-6 ,

IXX-88512 - AliACHMENT 7 PAGE 55 0F 66_ , me 3 O U p"a[' ' Q REACTOR COOLANT SYSTEM BASES SPECIFIC ACTIVITY (Continued) Reducing T to less than 500*F prevents the release of activity should asteamgeneratN9 tube rupture since the saturation pressure of the reactor coolant is below the lift pressure of the atmospheric steam relief valves. The Surveillance Requirements provide adequate assurance that excessive specific activity levels in the reactor coolant will be detected in sufficient time to take corrective action. A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained. 3/4.4.9 PRESSURE / TEMPERATURE LIMITS The temperature and pressure changes during heatup and cooldown are limited to be consistent with the requirements given in the ASME Boiler and Pressure Vessel Code, Section III, Appendix G and 10 CFR 50 Appendix G.

1. The reactor coolant temperature and pressure and system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with Figures 3.4-2 and 3.4-3 for the service period specified tnereon:

a. Allowable combinations of pressure and temperature for specific temperature change rates are below and to the right of the limit lines shown. Limit lines for cooldown rates between those presented may be obtained by interpolation; and

b. Figures 3.4-2 and 3.4-3 define limits to assure prevention of non-ductile failure only. For normal operation, other inherent plant characteristics, e.g. , pump heat addition and pressurizer heater capacity, niy limit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges.

O COMANCHE PEAK - UNIT 1 B 3/4 4-7 .

IIX-88512

AITACHMENT7 PAGE 56 0F 66 -

                  ~    ~

o\ REACTOR COOLANT SYSTEM DRAFT BASES PRESSURE / TEMPERATURE LIMITS (Continued)

2. These limit lines shall be calculated periodically using methods provided below,

3. The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the steam generator is below 70*F,

4. The pressurizer heatup and cooldown rates shall not exceed 100 F/h and 200*F/h, respectively) N: :pr y 0h:l' net be u::d if the t::peratur l di'ference bete::r the pre:Suri:er :nd th: :pr:y 'luid i: gr::ter th:n 9 3: 008 M625, and l

S. System preservice hydrotests and inservice leak and hydrotests shall be performed at pressures in accordance with the requirements of ASME Boiler and Pressure Vessel Code, Section XI. The new 10 CFR 50, Appendix G rule addresses the metal temperature of the closure head flange and vessel flange regions. This rule states that the mini-

 -  mum metal temperature of the closure flange region should be at least 120*F higher than the limiting RT NDT for these regions when the pressure exceeds 20%

of the preservice hydrostatic test pressure (621 psig for Westinghouse plants). For Comanche Peak Unit 1, the minimum temperature of the closure flange and the ves,el flange regions is 160*F since the limiting RT NOT is 40*F (see Table B 3/4.4-1). The Comanche Peak Unit 1 cooldown curves shown in Figure 3.4-3 are impacted by this new rule. The fracture toughness properties of the ferritic materials in the reactor vessel are datermined in accordance with the NRC Standard R(view Plan,. ASTM E185-73, and in accordance with additional reactor vessel requirements. These properties are then evaluated in accordance with Appendix G of the 1976 Summer Addenda to Section III of the ASME Boiler and Pressure Vessel Code and the calculation methods described in WCAP-7924-A, "Bcsis for Heatup and Cooldown Limit Curves," April 1975. I Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RTNOT, at the end of 16 effective full power years (EFPY) of service life. The 16 EFPY service

life period is chosen such that the limiting RT NDT at the 1/4T location in the core region is greater than the RT NDT of the limiting unirradiated material.

The selection of such a limiting RT NDT assures that all components in the Reactor Coolant System will be operated conservatively in accordance with applicable Code requirements. COMANCHE PEAK - UNIT 1 B 3/4 4-8 .

Txx-88512 ATTACHMENT 7 PAGE $7 0F 66 l b LJ A

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l O D O i 'l TABLE S 3/4.4-1 = i g; ; Y \ m.12w E REACIOR VESSEL TourJ00E55 g3U i , 3: " ! I eBdB 188 3 .. 35 Mll $HELF $NELF CODE Cu P fdDI RT

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i-40 100 40 -

119.0 Vessel Flange A588 C1. 2 R1101-1 - - le 70 to - - 97.0 l Inlet Ilezzle A588 C1. 2 R1305-1 - -

                                                                                                  -10     50

Inlet Blezzle -10 - 147.0 l A508 Cl. 2 R1105-2 - -

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                                                                                                 -10     50 I

Inlet slezzle A500 C1. 2 134.0 ch - R1105-4 - a0-so4650 sju-156.5 Outlet stezzle Outlet Ilezzle A508 C1. 2 A508 C1. 2 E1106-1 R1106-2 (W 40

T - 135.0

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outlet Nozzle Outlet Beezzle A50s C1. 2 R1106-3 - -

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135.5 A508'C1. 2 R1106-4 + -

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83.0 j Upper $4nell A5338, Cl. 1 R1104-2 . 08 .011 -50 100 40 - 75.0

Upper Satell A5338, C1. 1 R1304-3 .05 .010 -20 70 10 -

107.5 I Inter Senell A5338, C1. 1 R1107-1 .06 .010 -20 70 10 Inter Stiell A5338, C1. 1 111.5 93.5 ; R1107-2 .06 .010 -20 50 -10 123.5

103.6 Inter Satell A5338, Cl. 1 RI107-3 .05 .007 -20 70 10 131.0* 88.0 l Lower Stiell A5338, Cl. I R1108-1 .08 .008 -20 60 0 Lower 58eell 119.0 85.0 ! A5338 Cl. 1 R1100-2 .05 .006 -30 00 .?G 124.5 78.0 Lower SeneII A5338, Cl. 1 R1188-3 .07 .008 -30 60 0 Setten led. Terus 122.0 98.0 A5338, Cl. 1 R1112-1 - -

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                                                                                             '23                         l FIGURE 8 3/4.4-2                                M EFFECT OF FLUENCE AND COPPER CONTENT ON SHIFT OF RT         T FOR REACIOR VESSELS EXPOSED TO 550*F                                                         .

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TU-88512 - AllACMENT 7 PAGE 61 0F 66 REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) The reactor vessel materials have been tested to determine their initial RTNOT; the results of these tests are shown in Table B 3/4.4-1. Reactor opera-tion and resultant fast neutron (E greater than 1 MeV) irradiation can cause an increase in the RT N D I.. Therefore, an adjusted reference temperature, based upon the fluence, copper content, and phosphorus content of the material in question, can be predicted using Figure B 3/4.4-1 and the largest value of ART NOT computed by either Regulatory Guide ).99, Revision 1, "Effects of Residual Elements on Predicted Radiation Damage to Recctor Vessel Materials," or the Westinghouse Copper Trend Curves shown in Figure B 3/4.4-2. The heatup and cooldown limit curves of Figures 3.4-2 and 3.4-3 include predicted adjust-ments for this shift in RT NOT at the end of 16 EFPY as well as adjustments for possible errors in the pressure and temperature sensing instruments. Yalues of ART NDT determined in this manner may be used until the results from the material survsillance program, evaluated according to ASTM E185, are available. Capsules will be removed in accordance with the requirements of ASTM E185-73 and 10 CFR-Pert 50, Appendix H. ID1:l0481 4 ...; sche el is ;ha n in Teble 4.4 5. The lead factor represents the rela- l tionship between the fast neutron flux density at the location of the capsule and the inner wall of the reactor vessel. Therefore, the results obtained from the surveillance specimens can be used to predict future radiation damage to the reactor vessel material by using the lead factor and the withdrawal time of the capsule. The heatup and cooldown curves must be recalculated when the ART NOT determined from the surveillance capsule exceeds the calculated ART NOT f r the equivalent capsule' radiation exposure. Allowable pressure-temperature relationships for various heatup and cooldown rates are calculated using methods derived from Appendix G in Sec-tion III of the ASME Boiler and Pressure Vessel Code as required by 10 CFR 50 Appendix G, and these methods are discussed in detail in WCAP-7924-A. The general method for calculating heatup and cooldown limit curves is based upon the principles of the Linear Elastic Fracture Mechanics (LEFM) technology. In the calculation procedures a semielliptical surface defect with a depth of one quarter of the wall thickness, T, and a length of 3/2T is assumed to exist at the ir. ide of the vessel wall as well as at the outside of the vessel wall. The dimensions of this postulated crack, referred to in Appendix G of ASME Section III as the reference flaw, amply exceed the current capabilities of inservice inspection techniques. Therefore, the reactor operation limit curves developed for this reference crack are conservative and provide sufficient safety margins for protection O against nonductile failure. To assure that the radiatian embrittlement COMANCHE PEAK - UNIT 1 8 3/4 4-12 , i

                  ~                  .

TXX 88512 AlfAC19 0 I 7 PAGE 62 0F 66 REACTOR COOLA SYSTEM 8ASES PCSSURE/ TEMPERATURE LIMITS (Continued) effects are accounted for in the calculation of the limit curves, the most limiting value of the nil-ductility reference temperature, RTNOT, is used and this includes the radiation-induced shift, ARTHOT, corresponding to the end of the period for whi'ch heatup and cooldown curves are generated. The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, Kg , for the combined thermal and pressure stresses at any tima during heatup or cooldown cannot be greater than the reference stress intensity factor, K IR' for the metal temperature at that time. K yg is obtained from the reference fracture toughness curve, defined in Appendix G to the ASME Code. The K IR curve is given by the equation: KIR = 25.78 + 1.22,3 exp (0.0145(T-RTNOT + 160)] (1) Where: K IR is the reference stress intensity factor as a function of the metal temperature T and the metal nil-ductility reference temperature RT NOT. Thus, O the governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows: CKgg + kit IKIR (2) Where: K;g = the stress intensity factor caused by membrane (pressure) stress, K

                            !t
                               = the stress intensity fa-tor caused by the thermal gradients, KIR = constant provided by the Code as a function of temperature relative to the RT NOT    of the material, C = 2.0 for level A and B service limits, and C = 1.5 for inservice hydrostatic and leak test operations.

At any time during the heatup or cooldown transient, K IR is determined by the metal temperature at the tip of the postulated flaw, the appropriate value for RTNOT, and the reference fracture toughness curve. The thermal stresses resulting from temperature gradients through the vessel wall are calculated and then the corresponding thermal stress intensity factor, KIT, f r the reference flaw is computed. From Equation (2) the pressure stress intensity factors ara obtained and, from these, the allowable pressures are calculated. O COMANCHE PEAK - UNIT 1 8 3/4 4-13

                                                                            -                              \

IXX-88512 ATTACHMENT 7 PAGE 63 0F 66 l REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) C00LOOWN For the calculation of the allowable pressure versus coolant temperature during cooldown, the Code reference flaw is assumed to exist at the inside of the 'inssel wall. During cooldown, the controlling location of the flaw is always at the inside of tha wall because the thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates. Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations. From these relations, composite limit curves are constructed for each cooldown rate of interest. The use of the composite curve in the cooldown analysis is necessary because control o the cooldown procedure is based on measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the 1/4T vessel location is at a higher temperature than the fluid adjacent to the vessel ID. This condition, of course, is not true for the steady-state situa-tion. It follows that at any given reactor coolant temperature, the AT developed during cooldown results in a higher value of K at the 1/4T locatior, IR for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist such that the increase in K yg exceeds kit, the calculated allowable pressure during cooldown will be greater than the steady-state i value. The above procedures are needed because there is no direct control on temperature at the 1/4T location; therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and assures conservative operation of the system for the entire cooldown period. HEATUP Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4T defect at the inside of the vessel wall. The thermal gradients during heatup produca compressive stresses at the inside of the wall that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the K gg for the 1/4T crack during heatup is lower than the K IR for the 1/4T crack during steady-state l l conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist such that the effects of compressive j thermal stresses and different K IR 's for steady-state and finite heatup rates COMANCHE PEAK - UNIT 1 B 3/4 4-14 .

          ,-,      ,n.    -
                            ,n    , - , _ , . _ , , _ . - - --

' l TXX-88512 ATTACHMENT 7 PAGE640F66 , REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) do not offset each other and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/4T flaw is considered. Therefore, both cases have to be analyzed in order to assure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained. The second portion of the heatup analysis concerns the calculation of pressure-temperature limitations for the case in which a 1/4T deep outside surface flaw is assumed. Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and thus tend to reinforce any pressure stresses present. These thermal stresses, of course, are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramo. Furthermore, since the thermal stresses at the outside are tensile and increase with increasing heatup rate, a lower bound curve cannot be defined. Rather, each heatup rate of interest must be analyzed on an individual basis. Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced as follows. A composite curve is constructed based on a point-by-

 -./   point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration.

The use of the composite curve is necessary to set conservative heatup limitaticns because it is possible for conditions to exist such that over the course of the heatup ramp the controlling condition switches from the inside to the outside and the pressure limit must at all times be based on analysis of the most critical criterion. Finally, the composite curves for the heatup rate data and the cooldown rue data are adjusted for possible errors in the pressure and temperature sensing instruments by the values indicated on the respective curves. Although the pressurizer operates in temperature ranges above those for which there is reason for concern of nonductile failure, operating limits are provided to assure compatibility of operation with the fatigue analysis performed in accordance with the ASME Code requirements. 10 8: 0480 LOW TEMPERATURE OVERPRESSURE PROTECTION "O % The OPERABILITY of two PORVsjor an RCS vent opening of at least 2.98 square inches ensures that the RCS will be protected from pressure transients which could exceec' the limits of 10 CFR 50 Appendix G when one or more of the RCS cold legs are less than or equal to 350*F. Either PORV+has adequate relieving capa-bility to protect the RCS from overpressurization when the transient is limited to either: (1) the start of an idle RCP with the secondary water temperature of O the steam generator less than or equal to 50'F above the RCS cold leg tempera-tures, or (2) the start of charging pumps and their injection into a water-solid RCS. g 4 B 3/4 4-15 oc e Mar EM _ COMANCHE PEAK - UNIT 1 gg g ID I: 0/42

TXI-88512 ATTACHMENT 7

                PAGE 65 0F 66 REACTOR COOLANT SYSTEM BASES The maximum Nominal Allowed PORV Setpoint curve is derived from analyses b

which model the performar.ca of the overpressure protection system for a range of mass input and heat input transients. Figure 3.4-4 is based upon this analysis including cons *deration of the maximum pressure overshoot beyond the PORV setpoint which can occur as a result of time delays in signal processing and valve opening, instrument uncertainties, and single failure. For ehe~tran-sients noted, the resulting pressure will not exceed the nominal 10 Effective Full Power Years (EFPY) Appendix G reactor vessel NOT limits and the forces gen-erated due to PORY cycling do not exceed PORV piping and structural limitations. To ensure that mass and heat input transitents more severe than those assemed cannot occur Technical Specifications require the lockout of all safety

               - injection pumps and one charging pump while in MODES 4,- 5 and 6 with the reactor vessel head installed, and disallow start of an RCP if secondary temperature is
             ,  more than 50'F above primary temperature.

Operation below 350*F but greater than 325'F with charging and safety injec-tion pumps OPERA 8LE is allowed for up to 4 hours. 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 .O 1 ouring testing of a redundant channel are not considered to be credible accidents. The Maximum Allowed PORY Setpoint for the LTOPS will be updated based on the results of examinations of reactor vessel material irradiation surveillance 1 specimens.m. __mm performed,_as required by 10 CFR fen 50, Appendix H, . .: 'n ; ;;-d;n;; a.

                - ... m. . . . . . . . . . . ......m....i....,,._e..

3/4.4.10 STRUCTURAL INTEGRITY l 4 i The inservice inspection and testing programs for ASME Code Class 1, 2 , and 3 components ensure that the structural integrity and operational readiness l 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 reitef has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(1).

Components of the Reactor Coolant System were designed to provide access

to pensit inservice inspections in ac i Boiler anc' Pressure Vessel Code. h97(gordance with Section XI of the /9753 ASMEJ

                                                                                                                   . Edition l               3/4.4.11 REACTOR COOLANT SYSTEM VENTS l                         Reactor Coolant System vents are provided to exhaust noncondensible gases

and/or steam from the Reactor Coolant System that could inhibit natural circul- ;

ation core cooling. The OPERA 8ILITY of least one Reactor Coolant System vent path from the reactor vessel head, and the pressurizer steam space, ensures that the capability exists to perform this function. COMANCHE PEAK - UNIT 1 B 3/4 4-16 ,

TII 88512 AffACHMENT7 PAGE 66 0F 66 REACTORC0blAN[ SYSTEM O BASES The valve redundancy of the Reactor Coolant System vent paths serves to minimize the probability of inadvertent or irreversible actuation while ensuring that a single failure of a vent valve, pcwer supply, or control system does not prevent isolation of the vent path. The function, capabilities, and testing requirements of the Reactor Coolant System ven'.s are consistent with the requirements of Item II.B.1 of NUREG-0737, "Clarification of TMI Action Plant Requirements," November 1980. O l O COMANCHE PEAK - UNIT 1 B 3/4 4-17 L

_m . TEX 88512 ATTACHMENT $ FASE 1 Of 15 O i i COMANCHE PEAX STEAM ELECTRIC STATION TECHNICAL SPECIFICATION 3/4.5 O l O

IXI-88512 AllACHMDi8 CPSES Technical Specifications M 2 # 15 NRC Draft 2 Markup - Section 3/4.5 e .

  '( )' Change 10#        Justification For Change 0253       This version separates out the response to a failure to meet the boron concentration requirements and increases the A0T from I hour to 72 hours. Justification for this change is as follows:

a) "During the injection phase of a LOCA analysis, the boron concentration of the accumulator wat~er is not accounted for (i.e., conservatively assumed to be 0 ppm). In addition, WCAP-8471-P-A, "The Westinghouse ECCS Evaluation Model: Supplementary Information", April 1975, indicates that, for breaks larger than 3 square feet, no credit is taken for rod insertion and shutdown is due to voiding. This implies that the boron concentration of the accumulator water during the injection phase of a LOCA is not required for shutdown. However, after the injection phase (i.e., accumulators and RWST are emptied), the minimum tech spec boron concentration for the accumulators is assumed in the long-term post-LOCA subcriticality computation in order to ensure that the core remains subcritical. (See Westinghouse Technical Bulletin NSID-TB-86-08, "Post-LOCA Long-p Term Cooling: Boron Requirements", October 31,

l. v 1936.)"

l l b) If the boron content of the accumulators is determined to be out of spec, it is not possible to change the boron content and confirm that the new value is within spec within I hour, or even 8 hours. Therefore, a more reasonable time of 72 hours has been allowed. The 72 hour time period has been compared to the time allowed for one of

two ECCS trains to be inoperable, which seems to l be more serious than one of four accumulators.

I c) The most probable way that an accumulator boron l content could be out of spec is because of a slow leak either into or out of the accumulator. Thus, when the out of spec condition is discovered, the actual boron content will probably be close to the required condition. l This change is similar to that Licensed at South l Texas. l l l

IXX 88512-8 35 CPSES Technical Specifications NRC Oraft 2 Markup Section 3/4.5 Justification For Change v(n) Change 10# 0256 Remove the requirement for ACOT and channel calibration on the accumulator level and pressure instruments. This change is inconsistent with the majority of system related Technical Specifications to detail the instrumentation requirements and associated Surveillance requirements. This is even more true for monitoring instrumentation as opposed to control instrumentation. The accumulator level and pressure instrumentation have no control functions and provide indication / alarm only. The normal calibration program and quality requirements are sufficient to ensure the reliability and accuracy of this instrumentation. In addition, Surveillance Requirement 4.5.la requires that tank level and pressure be verified once per 12 hours; this same instrumentation is thus used at this frequency to perform the Surveillances. It is particularly inconsistent to require an Analog Channel Operational Test on this instrumentation. This test is typically specified on instruments that perform a tri ) or control function, not on monitoring instruments 4 Eac1 of these instruments is redundant on each tank, thus providing an added degree of reliability. The shiftly readings taken for Surveillance 4.5.la provide a check of IQ one instrument against the other (equivalent to a channel check) thus minimizing the change of an undetected V instrument malfunction. Another aspect is that the indicated Technical Specification Limits and alarm settings are established bssed on conservative instrument draft values consistent with the calibration cycle of 18 months. There is no justification for a monthly check. Should the calibration determine that unplanned instrument drift has occurred, the normal calibration program requires evaluation for corrective action (more frequent calibration, modification, etc.). This change is similar to that Licensed at Farley, Waterf,rd, Summer and Palo Verde. In addition, Calloway, Byron, Wolf Creek and Vogtle have only specified a channel calibration requirement. l 0266 The current note tied to this Surveillance Requirement is changed to ensure there is no confusion as to what is required to isolate the pump from the RCS. Without this change, it could be interpreted that the pump discharge l valve has to be shut which is one way to accomplish l isolation but not the only way. This change was approved in the meeting on 1/11/88 for Specification 3.5.3.1 and was over looked for this Specification. iO l l

IXX-88512 i AliACHMNT 8 1 -- 2- -- - --- 1 PAGE 4 0F 15 f l EMERGENCY CORF COOLING SYSTEMS 3/4.5.1 ACCUMULATORS k COLO LEG INJECTION LIMITING CONDITION FOR OPERATION 3.5.1 Each cold leg injection accumulator shall be OPERABLE with:

a. The discharge isolation valve open with power removed,

b. A contained borated water volume of between 6253 gallons ([Later]%

of span) and 6465 gallons ([Later]% span)

c. A boron concentration of between (1900) and (2100] ppm, and

d. A nitrogen cover pressure of between 605 and 655 psig.

ID 3: 0253 APPLICABILITY: MODES 1, 2, and 3*. or the here te.nk rd io 9 odside ACTION: .t.h e. re d'i rc k cVe \0es

a. With one cold leg injection accueulator inoperable, except as a result of a closed isolation valv , restore the inoperable accumulator to OPERABLE status within 1 hour or be in at least HOT STAND 8Y within O, the next 6 hours and reduce pressurizer pressure to less than 1000 psig within the following 6 hours,

{ b. With one cold leg injection accumulator inoperable due to the isolation valve being closed, either immediately open the isolation valve or be in at least HOT STANDBY within 6 hours and reduce pressurizer pressure to less than 1000 psig within the following 6 hours. l r ID 3: 0253 SURVEILLANCE REQUIREMENTS I l 4.5.l g Each cold leg injection accumulator shall be demonstrated 1 OPERABLE: I

a. At least once per 12 hours by:

1) Verifying the contained borated water volume and nitrogen cover pressure in the tanks, and

2) Verifying that each cold leg injection accumulator isolation valve is open.

c. (0; O tQ boven _en cc M rd '.c n e.k e nt coli \er in ( th en

_, a.cc o m o l oAer eatsde. %e fe \.m,t, ye)stey % bc<en t.o n te nt r oki e s. % dee'.reA W n. -tk te d.t W.w. il hears o< b,e in aA Ra s+ \o; tek ( .vn ;ts

                                                             % e. nex_t (o hoocs o. w l red o c.erot esTo ;g                     eV I
                                     *kk^

l ( "Pressurizer pressure above 1000 psig. ~ Pre 55 ore h ,less t Q

                                                                                            %o co\hp'gn3. 3 4.t%  lo      %

hoors. COMANCHE PEAK - UNIT 1 3/4 5-1 \___ _ - - - _ . .-- - - - _ .

IXX-88512 :

       '                                                                     - ~'               ' ~ ~ ~ ~ ~

AITACHMENT0 FABE 5 0F 15 ; _ . l EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

b. At least once per 31 days and within 6 hours after each indicated solution volume increase of greater than or equal to 101 gallons

([Later]% of span) by verifying the boron concentration of the solution in the water-filled accumulator;

c. At least once per 31 days when the RCS pressure is above 1000 psig by verifying that power to the isolation valve operator is removed.

4. 5.1. 2 Each-accumulator-water-level-and-pressure channel sh !' be &=n-

                              +trate& OPERABLE.                                                                       10 1: 0256
                                    -e.          At-least-once-pe&31-days -be- the-performance of en ANAt40-CifANNEtn
                                              -OPERATIONAL TEST, and h      At_}6a&t-once-per 18 Mnth! by thO p0rf9Pm4Mcc Of ; CM',P"El -

{At1BRAHON-O e F I v COMANCHE PEAK - UNIT 1 3/4 5-2

e

  ,.s IIX-88512                                        . -. -   ..   ._ ._

g, PAGE 6 0F 15 EMERGENCY CORE COOLING SYSTEMS , f 3/4.5.2 ECCS SUBSYSTEMS - T,yg > TO 350*F LIMITING CON 0iTION FOR OPERATION 3.5.2 Two independent Emergency Core Cooling System (ECCS) subsystems shall be OPERABLE with each subsystem comprised of:

a. One OPERA 8LE centrifugal charging pump,

b. One OPERABLE Safety Injection pump,

c. One OPERABLE RHR heat exchanger,

d. One OPERA 8LE RHR pump, and

e. An OPERA 8LE flow path capable of taking suction from the refueling water storage tank on a Safety Injection signal and automatically opening the containment sump suction valves during the recirculation phase of operation.

APPLICABILITY: MODES 1, 2, and 3". ACTION:

a. With one ECCS subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 72 hours or be in at least HOT STAN0PY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours,

b. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describ-ing the circumstances of the actuation and the total eccumulated actuation cycles to date. The current value of the usage factor for each affected Safety Injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.

i i

        *The provisions of Specification 3.0.4 and 4.0.4 are not applicable for entry into Mode 3 for the centrifugal charging pumps and the safety injection pumps declared inoperable pursuant to Specification 3.5.3 provided the centrifugal charging pumps and the safety injection pumps are restored to OPERABLE status within 4 hours or prior to the temperature of one or more of the RCS cold legs exceeding 375'F, whichever comes first.

l O COMANCHE PEAK - UNIT 1 3/4 5-3 I

TXX-88512 } ATTACHMENT 8 PACE 7 0F 15 u EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE:

a. At least once per 12 hours 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, O Accum. Discharge Open* 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 9

b. 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.

c. By a visual inspection which verifies that no loose debris (rags, trash, clothing, etc.) is present in the containment which could be transported to the containment sump and cause restriction of the pump i

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 Reactor Coolant System pressure signal greater than or equal to [425] psig the interlocks prevent the valves from being opened, and b) With a simulated or actual Reactor Coolant System pressure signal less than or equal to 750 psig the interlocks will cause the valves to automatica1.y close.

         "Surveillance Requirements covered in Specification 4.5.1[.

COMANCHE PEAK - UNIT 1 3/4 5-4

I , IXX-88512 ~~ ATTACMENT 8 PAGE 8 0F 15 EMERGENCY CORE COOLING SYSTEMS k SURVEILLANCE REQUIREMENTS (Continued)

2) A visual inspection of the containment sump and verifying that the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens, etc.) show no evidence of structural distress or abnormal corrosion.

e. 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 Safety Injection actuation 4nd-test signals, and

2) Verifying that each of the following pumps start automatically upon receipt of a Safety Injection actuation test signal:

a) Centrifugal charging pumps, b) Safety Injection pumps, and-c) RHR pumps.

f. By verifying that each of the following pumps develops the indicated differential pressure on recirculstion flow when tested pursuant to O- (- Specification 4.0.5:

1) Centrifugal charging pump 1 2370 psid,

2) Safety Injection pump 1 1440 psid, and

3) RHR pump > 170 psid.

g. By verifying the correct position of each mechanical position stop for the following ECCS throttle valves:

1) Within 4 hours following completion of each valve stroking operation or maintenance on the valve when the ECCS subsystems I are required to be OPERABLE, and

2) At least once per 18 months.

l CCP/SI System Valve Number SI System Valve Number ~ SI-8810A SI-8822A S'!-8816A SI-8810B SI-88228 SI-8816B SI-8810C SI-8822C SI-8816C i SI-88100 SI-88220 51-88160 Oc COMANCHE PEAK - UNIT 1 3/4 5-5 l l

     '   IIX-88512                                   ..      . _ . . _ l. . I ATTACHMENT 8 FAGE 9 of 15 EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

h. By performing a flow balance test, during shutdown, following com-plation of modifications to the ECCS subsystems that alter the subsystem flow characteristics and verifying that:

1) For centrifugal charging pump lines, with a single pump running:

a) The sum of the injection line flow rates, excluding the highest flow rate, is greater than or equal to 333 gpe, and b) The total pump flow rate is less than or equal to 555 gpa.

2) For Safety Injection pump lines, with a single pump running:

a) The sum of the cold leg injection line flow rates, excluding the highest flow rate, is greater than or equal to 437 gom, and b) The total pump flow rate is less than or equal to 660 O gpa.

3) For RHR pump lines, with a single pump running, the sum of the cold leg injection line flow rates is greater than or equal to 4652 gpe.

i. Prior to entering MODE 3 and following any maintenance or operations activity which drains porticns of the system by venting the ECCS

     -                pump casing and accessible discharge piping high points.

l O l COMANCHE PEAK - UNIT 1 3/4 5-6 ' l

111 88512 1 ATTACHMENT 8 FAGE 10 0F 15 EMERGENCY CORE COOLING SYSTEMS 3/4.5.3 ECCS SUBSYSTEMS - T,yg < 350*F ECCS SUBSYSTEMS LIMITING CONDITION FOR OPERATION 3.5.3.1 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:

a. One OPERABLE centrifugal charging pump,*

b. One OPERABLE RHR heat exchanger,

c. One OPERABLE RHR pump, and

d. An OPERABLE flow path capable of taking suction from the refueling water storage tank upon being manually realigned and transferring suction to the containment sump during the recirculation phase of operation.

APPLICABILITY: MODE 4. O ACTION:

a. With no ECCS subsystem OPERABLE because of the inoperability of either the centrifugal charging pump or the flow path from the refueling water storage tank, restore at least one ECCS subsystem to OPERABLE status within 1 hour or be in COLD SHUTDOWN within the next 20 hours,

b. With no ECCS subsystem OPERABLE because of the inoperability of either the residual heat removal heat exchanger or RHR pump, restore at least one ECCS subsystem to OPERABLE status or maintain the Reac-tor Coolant System T avg less than 350'F by use of alternate heat removal methods,

c. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describ-ing the circumstances of the actuation and the total accumulated actuation cycles to date. The current value of the usage factor for each affected Safety Injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.

A maximum of two charging pumps shall be OPERABLE whenever the temperature of one or more of the RCS cold legs is less than or equal to 350*F.

COMANCHE PEAK - UNIT 1 3/4 5-7

i IIX-88512

      ..a    gggg g g g                                . - - - - - . .        . . . - - . .     .

PAGE 11 OF 15

                         ~     ~

EMERGENCY CORE COOLING SYSTEMS DRAFT SURVEILLANCE REQUIREMENTS j 4.5.3.1.1 The ECCS subsystem shall be demonstrated OPERABLE per the applicable requirements of Specification 4.5.2. 4.5.3.1.2 A maximum of two charging pumps shall be OPERABLE; one charging pump shall be demonstrated inoperable

by verifying that the motor circuit breaker is secured in the open position within 4 hours after entering MODE 4 fromv3 or prior to whichever(temperature of one occurs first and at or more least of the once perRCS cold legs 31 days decreasing below 325'F, thereafter.

4hc O

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

COMANCHE PEAK - UNIT 1 3/4 5-8 *

   .   . IXX-88512                                 - . . _ .         .                       .

AllACHMENT 8 PAGE 12 0F 15 EMERGENCY CORE COOLING SYSTEMS k 3/4.5.3 ECCS SU8 SYSTEMS - T,yg < 350*F SAFETY INJECTION PUMPS LIMITING CONDITION FOR OPERATION _ 3.5.3.2 All Safety Injection pumps shall be inoperable. APPLICABILITY: Modes 4, 5, and 6 with the reactor vessel head o.3 ACTION: With a Safety Injection pump OPERABLE, restore all Safety Injection pumps to an inoperable status within 4 hours. SURVEILLANCE REQUIREMENTS 4.5.3.2 All Safety Injection pumps shall be demonstrated inoperable

by r verifying that the motor circuit breakers are secured in the open position

( ( within 4 hours after entering MODE 4 from MODE 3 or prior to the temperature of A one or more of the RCS cold legs decreasing below 325'F, whichever occurs first and at least once per 31 days thereafter. I

         *An inoperable pump may be energized for te                      v for fiMing acmaulators 10 h 02%

pravided the discharge at the pump has bee' frox .he RCS of a closed isolation valveNwith power removed from t - ofG or by a manual O( isolation valva 6hecured in the closed po' COMANCHE PEAK - UNIT 1 3/4 5-9

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

III 88512 - ~ - ~ ~ ATIACHMENT8 FAGE130F15 t

 .             BORON INJECTION SYSTEM 3/4.5.4 REFUELING WATER STORAGE TANK LIMITING CON 0! TION FOR OPERATION                                                       _

3.5.4 The refueling water storage tank (RWST) shall be OPERABLE with:

a. Aminimumcontainedboratedwatervolumeof(479,900 gallons

((Later]% of span),

b. Aboronconcentrationofbetween2000and(2200 ppm of boron,

c. A minimum solution temperature of 40*F, and A maximum solution temperature of 120'F.

d. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With the RWST inoperable, restore the tank to OPERABLE status within 1 hour or be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.5.4 The RWST shall be demonstrated OPERAB'.E:

a. At least once per 7 days by:

1) Verifying the contained borated water volume in the tank, and

2) Verifying the boron concentration of the water.

b. At least once per 24 hours by verifying the RWST temperature when the outside air temperature is less than 40*F or greater than 120*F.

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

III-88512 AliACHMENT8 PAGE 14 Of 15 [] v 3/4.5 EMERGENCY CORE COOLING SYSTEMS 1.. 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 below the pressure of the accumulators. This initial surge of water into the core provides the initial cooling mechanism during large RCS pipe ruptures. The limits on accumulator volume, boron concentration and pressure ensure that the assumptions used for accumulator injection in the safety analysis are met. The accumJlator power operated isolation valves are Considered to be "operating bypasses" in the context of IEEE Std. 279-1971, which requires that 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 BT ICSB 18. This is accomplished via key-lock control board cut-offswitch(es.\ The limits for operation with an accumulator inoperable for any' reason 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 immediately 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 acceptabli 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 j recirr:ulation mode during the accident recovery period. With the RCS temperature below 350*F, one OPERABLE ECCS subsystem is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the limited core cooling requirements. i The limitation for a maximum of two charging pumps to be OPERABLE and the I requirement to verify one charging pump and all safety injection pumps O COMANCHE PEAK - UNIT 1 B 3/4 5-1 ,

TXX-88512

               .                           AliACHMENT8 PAGE 15 0F 15 EMERGENCY CORE COOLING SYSTEMS BASES ECCS SU8 SYSTEMS (Continued) 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 requirement to remove power from certain valve operatces is in accord-ance with Branch Technical Position ICSB-18 for valves that fail to meet single failure considerations. Power is removed via key-lock switches on the control board. The Surveillance Requirements provided to ensure OPERABILITY of each component ensures that at a minimum, the assumptions used in the safety analyses are met and that subsystem OPERABILITY is maintained. Surveillance Requirements for throttle valve position stops-and flow balan:;e testing provide assurance that proper ECCS. flows will be maintained in the event of a LOCA. Maintenance of proper flow resistance and pressure drop in the piping system to each injection point is necessary to: (1) prevent total pump flow from exceeding runout conditions when the system is in its minimum resistance configuration, (2) provide the proper flow split between injection points in accordance with I the assumptions used in the ECCS-LOCA analyses, and (3) provide an acceptable , level of total ECCS flow to all injection poir^.s equal to or above that assumed in the ECCS-LOCA analyses. 3/4.5.4 REFUELING WATER STORAGE TANK The OPERABILITY of the refueling water storage tank (RWST) as part of the ECCS ensures that a sufficient supply of borated water is available for injec-i tion by the ECCS in the event of a LOCA. The limits on RWST minimum volume and

boron concentration ensure that: (1) sufficient water is available within l

containment to permit recirculation cooling flow to the core, (2) for small break LOCA and steam line breaks, the reactor will remain suberitical in the i cold condition following mixing of the RWST and the RCS water volumes with all

control rods inserted except for the most reactive control assembly,-end-(3)

;                                        fon large break LOCAs, the reactor will remain subcritical in the cold condition following mixing of the RWST and the RCS water volumes with all shutdown and control rods fully withdrawn, and (4) sufficient time is available for the operator to take manual action and complete switchover of ECCS and containment spray suction to the containment sump without emptying the RWST or losing suction.

The contained water volume limit includes an allowance for water not usable I because of tank discharge line location or other physical characteristics. l The limits on contained water volume and boron concentration of the RWST ! also ensure a 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 . 1 iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components. . COMANCHE PEAK - UNIT 1 B 3/4 5-2 , l

IIX 88512 AllACHMENT 9 PAGE I 0F 42 O COMANCHE ?EAK STEAM ELECTRIC STATION TECHNICAL SPECIFICA.TIO.N 3/4.6 O l O

TIX 90512 ATTAtHMENT 9 nGE 2 # 4 CPSES Technical Specifications NRC Draft 2 Markup Section 3/4.5,

    . Change 10#   Justification For Change
 '(VI       0268   Replace / Delete reference to Table 3.6-1 with the CPSES 0288   Technical Specification improvement Program (TSIP).

~ 0290 This change is based on the relocation of Table 3.6-1 0292 to the TSIP which is justified under Specification 0294 3.6.3. 0296 0279 A note has been added to the Action Statement to allow for the OPERABLE air' lock door that is required to be closed by the Action Statement to be opened for short periods of time. This note only applies to the outer air lock door and for a specified time of 15 minutes per individual entry. This change would allow the inner door to be repaired and returned to an OPERABLE status with minor impact of safety and plant operations. By following this Action there is presently no provision to allow opening the air lock door that is required to be closed except to voluntarily enter 3.0.3 which is strictly not allowed. This change is similar to that licensed at Vogtle, Shearon Harris and Seabrook. 0281 The Action time requirement of one hour is not consistent with the eight hours allowed by the containment temperature Specification 3/4.6.1.5. Both of these parameters are used in the same accident analyses as a set (q/

    ,              of initial conditions. This additional time allowance would provide adequate time to perform all required samples and fulfill all administrative requirements.

Without this change the entire process would have to be started so early in the pressure band that the operators would be unduly burdened which would distract them from peforming their normal duties of maintaining the plant in ; a stable and safe condition. 0282 This Surveillance Rcquirement has been revised 0485 to reflect the plant specific system design for average containment temperature. The average containment temperature is detennined by the volume weighted adjusted average temperature of two temperatures of which one is from elevation 1000'6" or above and the other is from elevation 860' or above. This adjusted average will be determined by correlating the preoperational test data to detennine the correction factor to be used for the temperatures taken at the two different locations. This will allow remote or local measurements of the containment temperature as well as the flexibility of being able to use any location (s) above the specified elevations since one of the two temperatures will always be determined using the correction factor for elevation 860. If, a higher elevation is used for this reading, the calculated average temperature would be higher than the actual

average temperature since as the reading elevation is increased,the temperature increases.

in se512 ArtAcment9 CPSES Technical Specifications PAGE 3 0F 42 NRC Draft 2 Markup 4 Section 3/4.6

-(]

v Change 10# Justification for Change 0285 The LC0 Statement and Action Statement for 0286 the 18 inch pressure relief valves have been changed to allow for these valves to be opened as necessary to relieve containment pressure or for required surveillance testing. This is based on the fact that the valve has been qualified to shut in the time prescribed in the accident analysis. This is accomplished by physically limiting the valve opening to ensure that the valves meet isolation time requirements. This plant specific change is similar to that licensed at South Texas, Shearon Harris, Vogtle and Seabrook. 0288 See 10# 0268 0290 See 10# 0268 0291 Added new Action e. to take exception to Specification 3.0.4. If a containment isolation valve is inoperable, - proper isolation of the affected penetration per Action Statements b. or c. will maintain containment integrity and allow operations to continue in Modes 1, 2, 3, and 4 with no impact on plant safety or operation. (~ 0292 See 10# 0268 0293 Delete the phrase "during COLD SHUTDOWN er REFUELING MODE" removes an unnecessarily resrictive requirement for the desired testing. Many of the valves will or could be tested for Phase "A", Phase "B" isolation, or Containment Ventilation Isolation during the quarterly Slave Relay Testing. This change is similar to that Licensed at Diablo Canyon and Waterford. 0294 See ID# 0268 0296 See ID# 0268 0485 See ID# 0282 0575 The addition of a 3.0.4 exemption is consistent with the other post accident Technical Specifications. The allowance to be able to change Modes while complying with the 30 day Action Requirement is the same as that allowed by Technical Specifications 3.3.3.6, Accident Monitoring Instrumentation, and 3.6.4.2, Hydrogen Monitors, which both have 3.0.4 exemptions. O

IXX88512 Ali&CHMDI9 W 4 # 42 CPSES Tcchnical Specifications

~ '

NRC Ordft 2 Markup Section 3/4.6 .

(} Change 10# Justification For Change 0940 This Table is being relocated to the CPSES Technical S)ecification Improvement Program. TV Electric believes t1e inclusion of this Table is unnecessary and the information would be more appropriately addressed in the CPSES Technical Specification Improvement Program.

Relocation of this Table is consistent with the guidance provided in the NRC's Interim Policy Statement (S2FR3788), February 6,1987, and the recommendations of the Westinghouse Owners Group MERITS Program. Priority is given to the relocation of this Table _since the detailed information is not used by the Licensed Operator, but is purely used as testing criteria provided by the venaor supplier. The information currently in this Table is more appropriately maintained in a document subject to TV Electric administrative control and 10CFR50.59 review under the CPSES Technical Specification Improvement Program. > ' - - ) This change is similar to that Licensed at Shearon llarris, Seabrook and Vogtle. The specific information supplied in Table 3.8-1 has not completed the validation process and is expected to change. As soon as this information is available it will be transmitted to the NRC under separate letter.

TXX-88512 , AffACHMENT 9 FAGE 5 0F 42 O' 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION

3. 6.1.1 Primary CONTAINMENT INTEGRITY shall be maintained.

APPLICABILITY: H0 DES 1,.2, 3, and 4. ACTION: Without primary CONTAINMENT INTEGRITY, restore CONTAIhMENT INTEGRITY within I hour or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.

SURVEILLANCE REQUIREMENTS

4. 6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated:

a. At least once per 31 days by verifying that all penetrations

not capable of being closed by OPERA 8LE containment automatic isolation l valves and required to be closed during accident conditions are '

closed by valves, blind flanges, or deactivated automatic valves secured in their positions, except as provided in Tele 3.5-1 cf S;::ific;ti;. 3.0.4.1;Iw.J w ivJ s cd MW %ve + %N l C P% S ID1:0268 I

b. By verifying that each containment air lock is in compliance,with the requirements of Specification 3.6.1.3; and

c. After each closing of each penetration subject to Type B testing, except the containment air locks, if opened following a Type A or B test, by leak rate testing the seal with gas at a pressbre not less .

than P. 48.3 paig, and verifying that when the measured leakage rate for th8se seals is added to the leakage rates determined pursuant to Specification 4.6.1.2d. for all other Type B and C penatrations, the combined leakage rate is less than 0.60 L,.

      *Except valves, blind flanges, and deactivated automatic valves which are located inside the containment and are locked, sealed or otherwise secured in the closed position. These penetrations shall be verified closed during each COLD SHUTDCWN except that such verification need not be performed more O      often than once per 92 days.          The blind flange on the fuel transfer canal need not be verified closed except after each drainage of the canal.

COMANCHE PEAK - UNIT 1 3/4 6-1

IXX88512 i

   .   .   .        AliACH S T 9                                              --

PAGE 6 0F 42 (m CONTAINMENT SYSTEMS CONTAINMENT LEAKAGE ( LIMITING CONDITION FOR OPERATION 3.6.1.2 Containment leakage rates shall be limited to:

a. An overall integrated leakage rate of:

1) Less than or equal to L,, 0.10% by weight of the containment air per 24 hours at P,, @8.3) psig, or

2) Less than or equal to L t, 0.030% by weight of the containment air per 24 hours at a reduced pressure of Pt .[24.05]psig. l

b. A combined lakage rate of less than 0.60 L, for all penetrations and valves subject to Type B and C tests, when pressurized to P,.

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

 /3 I

g With either the measured overall integrated containment leakage rate exceeding O.75 L, or 0.75 L t, as applicable, or the measured combined leakage rate for al' penetrations and valves subject to Types B and C tests exceeding 0.60 L,,

restore the overall integrated leakage rate to less than 0.75 L, or less th;.n 0.75 Lg , as applicable, and the combined leakage rate for all penetrations l subject to Type B and C tests to less than 0.60 L, prior to increasing the I . Reactor Coolant System terperature above 200*F.

l SURVEILLANCE REQUIREMENTS

     -            4. 6.1. 2 The containment leakage rates shall be demonstrated at the following test schedule and shall be dttermined in conformance with the criteria speci-fied in Appendix J of 10 CFR Part 50 using the methods and provisions of ANSI N45.4-1972:                                                             .

l a. Three Type A tests (Overall Integrated Containment Leakage F. ate) shall be conducted at 40110 month,jntervals during shutdown at j a pressure not less than either P,,l_48.3]psig, or at Pt 24.05 psig, l during each 10 year service period. The third test of each set shall be conducted during the shutdown for the 10 year plant i inservice inspection; O( COMANCHE PEAK - UNIT 1 3/4 6-2, ,

DN8512 ) ' i

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ATTACHMENT 9 FAGE 7 0F 42 l CONTAD64ENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

b. If any periodic Type A test fails to meet either 0.75 L, or 0.75 Lt '

the test schedule for subsequent Type A tests shall be reviewed and approved by the Conunission. If two consecutive Type A tests fail to meet either 0.75 L, or 0.75 L t, a Type A test shall be perfonned 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 test which:

          .                       1)    Confirms the accuracy of the test by verifying that the supple-mental test result, I.c, is in accordance with the appropriate following equation:

1L c ~ Il am

Lo ) I 1 0.25 La or l Lc ' (lta + l o) l 5 0.25 Lt where L,,or Ltm is the measured Type A test leakage and L o is the superimposed leak; .

2) Has a duration sufficient to establish accurately the change in leakage rate between the Type A test and the supplemental test; and

3) Requires that the rate at which gas is injected into the contain-ment or bled from the containment during the supplemental tsst is between 0.75 L, and 1.25 L,; or 0.75 Lt and 1.25 Lg.

d. Type 8 and C tests shall be conducted with gas at a pressure not less than Pa ,[48.3]psig, at intervals no greater than 24 months l except for tests involving:

l

1) Air locks,

2) Containment ventilation isolation valves with resilient material seals, l 3) Safety Injection Valves 1-8802A, 1-88028 and 1-88d0, and

4) Containment Spray Valves 1HV-4776, 1HV-4777, 1CT-142, and ICT-145.

e. Air locks shall be tested and demonstrated OPERABLE by the require-l ments of Specification 4.6.1.3; l

I f. Containment ventilation isolation valves with resilient uaterial seals shall be tested and demonstrated OPERABLE by the requirements ofSpecification4.6.1.7.{or4.6.1.7.pasapplicable;

g. Safety Injection Valves 1-8802A, 1-88028, and 1-8840 shall be

' demonstrated OPERABLE by performance of a step leakage measurement, while pressurized to a pressure not less than Pa, (48.3]psig, at X intervals no greater than 92 days.

h. Containment Spray Valves 1HV-4776, 1HV-4777, ICT-142, and 1CT-145 shall be leak tested with water, at a pressure not less than Pa, [48.3]psig, O- at intervals no greater than 24 months; and

1. The provisions of Specification 4.0.2 are not applicable.

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

a wm AITACHilENT 9

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PAGE 8 0F 42 . i i CONTAINMENT SYSTEMS CONTAINMENT AIR LOCKS LIMITING CON 0! TION FOR OPERATION 3.6.1.3 Each containment air lock shall be OPERABLE with:

a. Both doors closed except when the air 1r ' is being used for normal transit entry and exit through the conta tent, then at least one air lock door shall be closed, and

b. An over 1 air lock leakage rate of less than or equal to 0.05 L at a

P,,[48.3psig. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:

a. With one containment air lock door inoperable:

1. Maintain at least the OPERABLE air lock door closed and either restore the inoperable air lock door to OPERABLE s atus within 24 hours or lock the OPERABLE air lock door closed O Operation may then continue until performance of the next

() ( 2. required overall air lock leakage test provided that the OPERABLE air lock door is verified to be locked closed at least once per 31 days;

3. Otherwise, be in at least HOT STAN0BY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours; and

4. The provisions of Specification 3.0.4 are not applicable.

b. With the containment air lock inoperable, except as the result of an inoperable air lock d

  =                       restore the inoperabl,oor, e air lock maintain  at least one to OPERABLE         air within status  lock door  closed; 24 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.

m om > E flie m z.nh u 3 o y enin] Yb re w Oie in vier ad r kotk c(cor, d tn ozecd\e , is 7erme A\e b c. 3er:et oct

                  % aceet \ 5                    wndes per ino riha.i e.n b 3 or        u&.

O( COMANCHE PEAK - UNIT 1 3/4 6-4 - a

IXX 88512

  .       ATTACHMENT 9                                               -

PAGE 9 0F 42 (3 CONTAINMENT SYSTEMS U ( SURVEILLANCE REQUIREMENTS h 4.6.1.3 Each containment air lock shall be demonstrated OPERABLE:

a. Within 72 hours following each closing, except when the air lock is being used for multiple entries, then at least once per 72 hours, by verifying seal leakage is less than 0.01 L as determined by precision flowmeasurementswhenmeasuredforat1 aft 30secondswiththe volume between the seals at a constant pressure of greater than or equal to[48.3]psig;

b. By conducting overall air lock leakage tests at not less than P,,

[48.3]psig, and verifying the overall air lock leakage rate is within its limit:

1) At least once per 6 months,* and

2) Prior to establishing CONTAINMENT. INTEGRITY when maintenance has been performed on the air lock that could affect the air lock sealing capability.**

s c. At least once per 6 months by verifying that only one door in each

 ]C                    air lock can be opened at a time.
         *The provisions of Specification 4.0.2 are not applicable.
        **This represents an exemption to 10 CFR 50 Appendix J, paragraph III.O.2(b)(ii).

COMANCHE PEAK - UNIT 1 3/4 6-5 .

IIX-88512 ATTACHMENT 9 PAGE 10 CF 42 i i CONTAINMENT SYSTEMS . INTERNAL PRESSURE LIMITING CONDITION FOR OPERATION

3. 6.1. 4 Pri:::ary containment internal pressure shall be maintained between

[-0.5] =d [1.5] p;ig. psby (-DJu] p,., .4hk.t) ~ J (/.r/ p W.; (g,4,.] pq ind,'cated h APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:

                                                          $                       10 I: 0281 Withthecontainmentinternalpressureoutsideofkthelimitsabove,restorethe                '   inte STANOBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.

O SURVEILLANCE REQUIREMENTS

4. 6.1. 4 The primary containment internal pressure shall be determined.to be within the limits at least once per 12 hours.

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

l Tu-08512 AliACHMENT9 *

      -                                                ~ ' ~ ~ ~ ~
         -PAGE 11 0F 42 l

CONTAINMENT SYSTEMS AIR TEMPERATURE C LIMITING CONDITION FOR OPERATION

3. 6.1. 5 Primary containment average air temperature shall not exceed 120'F. 1 1

APPLICABILITY: MODES 1, 2, 3, and 4. l ACTION: With the containment average air temperature greater than 120'F, reduce the average air temperature to within the limit within 8 hours, or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. ( SURVEILLANCE REQUIREMENTS cuc gadpsted

4. 6.1. 5 The primary containment average air temperature shall be the xLttu

        = tic s average of                   temperatures at the follo ing                                                           IDI 0282 determined at least once per 24 hougI uaw.w.t s, locations,and shall be
                                                      < g ,.A . ,   .

(o f whkh ableas t on. Location lev L i s ;-cc e L J:- .s.

a. ben El. ioco'.6' o r dove.'

b. Flece it. %c' c" f.

A Net:: "i n imum-e f-three-eleva t4ons-requ i red, l O( l COMANCHE PEAK - UNIT 1 3/4 6-7 -

l TXX-88512 l

   .          ATTACHMENT 9                          - - - - - -          ----    --   -

l PAGE 12 0F 42 l 1 l ( CONTAINMENT SYSTEMS CONTAINMENT STRUCTURAL INTEGRITY ( LIMITING CONDITION FOR OPERATION 3.6.1.6 The structural integrity of the containment shall be maintained at a level consistent with the acceptance criteria in Specification 4.6.1.6.1,

APPLICABILITY: MODES 1, 2, 3, and 4. I ACTION: With the structural integrity of the containment not conforming to the above requirements, restore the structural integrity to within the limits within 24 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUT 00WN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.6.1.6.1. Containment Surfaces. The structural integrity of the exposed 4 accessible interior and exterior surfaces of the containment, including the liner plate, shall be determined during the shutdown for each Type A contain-ment leakage rate test (reference Specification 4.6.1.2) by a visual inspection of these surfaces. This inspection shall be performed prior to the Type A containment leakage rate test to verify no apparent changes in appearance or other abnormal degradation. l 4.6.1.6.2 Reports. Any abnormal degradation of the containment structure detected during the above required inspections shall be reported to the:Comis-l sion in a Special Report pursuant to Specification 6.9.2 within 15 days. l ' This report shall include a description of the condition of the concrete, the inspection procedure, the tolerances on cracking, and the corrective actions taken. 1 l l 1 O< COMANCHE PEAK - UNIT 1 3/4 6-8 . l

IXX-88512 i ATTACMENT 9 PAGE 13 0F 42 l q\.s CONTAINMENT SYSTEMS , CONTAINMENT VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.6.1.7 Each ccatainment and hydroger ventilation isolation valve shall be OPERABLE and:

a. Each 48-inch and 12-inch containment and hydrogen purge supply and exhaust isolation valve shall be locked closed, and

b. The 18-inch containment p"essure relief discharge isolation valve (s)

                      -may-be-opefefosp-to-90-hours-during c c:lendar ye:r, sha\\ be c.lcsed, tc
                       %e mo dmom c Ate.nt Trad;cd                  Lt     ma o43 be- cTen \er vent APPLICABILITY:        MODES 1, 2, 3, and 4.          S b**

c.e,n b \ c e r 'i ' en 4er pec55u re

o. A for wrvestiance te sts ACTION: -tM o pe n regn & g\ gem % be

a. With any 48-inch or 12-inch containment or hydrogen purge supply and/or exhaust isolation valve open or not locked closed, lock close that valve or isolate the penetration (s) within 4 hours, otherwise -

be in at least HOT STANDBY within the next 6 hours and in COLD SHUTOOWN within the following 30 hours. D h 0286 i f roSons c%e r bn @4e6 In % i 7. b above, 1 O b. With the 18-inch containment pressure relief discharge isolation ( valve (s) open fork': r: than 90 heur; during : ::: lender year, close the open 18-inch valve (s) or isolate the penetration (s) within 4 hours, otherwise be in at least P 7 STAN0BY within the next 6 hours, and in COLD SHUTOOWN within the following 30 hours.

c. With a containment pressure relief discharge isolation valve (s) having a measured leakage rate in excess of the limits of Specifications 4.6.1.7.3 or with the containment and hydrogen purge supply or exhaust isolation valve (s) having a measured leakage rate in excess of the l limit of Specification 4.6.1.7.+ restore the inoperable valve (s) to l

OPERABLEstatuswithin24hoursf;otherwisebeinatleastHOTSTANDBY within the next 6 hours, and in COLD SHUTDOWN within the following

30 hours.

1 2. SURVEILLANCE REQUIREMENTS 4.6.1.7.1 Each 48-inch and 12-inch containment and hydrogen purge supply and exhaust isolation valve shall be verified to be locked closed at least once per 31 days.

4. 0.1. 7. 2 The-cumdative-time-that-a%-18-inchpressura relief discharge I tso1+trion-valves-have-been-open durJng a calendar- year shall be determinad at ID : 0285 least once per 7 days, 1

l O COMANCHE PEAK - UNIT 1 3/4 6-9

TXX-88512 AliACHMENT 9 - PAGE 14 0F 42 I O V CONTAINMENT SYSTEMS i SURVEILLANCE REQUIREMENTS (Continued) 1 4.6.1.7.9 At least once per 184 days on a STAGGERED TEST BASIS, the inboard and outboard isolation valves with resilient material seals in each locked closed 48-inch and 12-inch containment and hydrogen purge supply and exhaust penetration shall be demonstrated OPERABLE by verifying that the measured leakage rate is less than 0.05 L, when pressurized to P,. 3 4.6.1.7.4 At least once per 92 W ys each 18-inch containment pressure relief discharge isolation valve with resilient material seal; shall be demonstrated OPERABLE by verifying that the measured leakage rate is less than 0.06 La when pressurized to P,. i O l

O COMANCHE PEAK - UNIT 1 3/4 6-10

TXX-88512 - I ATTACMENT 9 PAGE 15 0F 42 f ps Q CONTAINMENT SYSTEMS 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT SPRAY SYSTEM LIMITING CONDITION FOR OPERATION 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: H0 DES 1, 2, 3, and 4. ACTION: With one Containment Spray System inoperable, restore the inoperable Containment Spray System to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours; restore the inoperable Containment Spray System to OPERABLE status within the next 48 hours or be in COLD SHUTDOWN within the following 30 hours. - - - SURVEILLANCE REQUIREMENTS l 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,

(

L- 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, disc arge flow through the test header of greater than or equal toL5800 gpm with the pump eductor line open when tested pursuant to Spec fica-tion 4.0.5; '

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 I signal, and

2) Verifying that each spray pump starts automatically on a Containment Spray Actuation or Safety Injection test signal.

1

d. At.1 east once per 5 years by performing an air or smoke flow test through each spray header and verifying each spray nozzle is unobstructed.

O( COMANCHE PEAK - UNIT 1 3/4 6-11 ,

in-88512 ~ ~ ~ ~ ~ ATTACHMENT 9 PAGE 16 0F 42 CONTAINMENT SYSTEMS SPRAY ADDITIVE SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.2 The Spray Additive System shall be OPERABLE with:

a. .A spray additive tank containing a volume of between 4900 and 5167Jgallonsofbetween28and30%byweightNaOHsolution,and

b. Four spray additive eductors each capable of adding HaOH solution from the chemical additive tank to a Containment Spray System pump flow.

APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With the Spray Additive System inoperable, restore the systee to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours; restore the Spray Additive System to OPERABLE status within the next 48 hours' , or be in COLD SHUTOOWN within the following 30 hours. 1 l p J SURVEILLANCE REQUIREMENTS 4.6.2.2 The Spray Additive System shall be demonstrated OPERABLE: l a. At least once per 31 days by verifying that each valve (manual, l power-operated, or automatic) in the flow path that is not locked, sealed, or othe mise secured in position, is in its correct position;

b. At least once per 6 months by:

1) Verifying the contained solution volume in the tank, and

2) Verifying the concentration of the NaOH solution by chemical analysis,

c. At least once per 18 months during shutdown, by verifying that each automatic valve in the flow path actuates to its correct position on a Containment Spray Actuation test signal; and

(

d. At least once per 5 years by verifying:

1) The flow path through the Spray Additive supply line, and

l 2) RWSTtestwaterflowratesofbetween50]GPMand[100GPMthrough l the eductor test loop of each4 of the Jpray Additive ystem. b o.s lO COMANCHE PEAK - UNIT 1 3/4 6-12

IXX-88512 PAGE 17 0F 42 l CONTAINMENT SYSTEMS . 3/4.6.3 CONTAINMENT ISOLATION VALVES ( LIMITING CONDITION FOR OPERATION 3.6.3 The containment isolation valves specified in Table 3.5-1 shall be 10 h 0288 OPERABLE.wHh-hMatten-t4mes as shown-in4aNe-h64r APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:

         *With one or more of the containment isolation valve (s) pecified in Teble ID h 0290
       -h64 inoperable, maintain at least one isolation valve OPERABLE in each                 l affected penetration that is open and:

a. Restore the ino'perable valve (s) to OPERABLE status within 4 hours, or Isolate each affected penetration within 4 hours by use of at least '

b. one deactivated automatic valve secured in the isolation position, or (] c. Isolate each affected penetration within 4 hours by use of at least one closed manual valve or blind flange, or (/ ( - d. Be in at least HOT STAN0BY within the next 6 hours and in COLO SHUTDOWN within the following 30 hours, s ID 1: 0291 SURVEILLANCE REQUIREMENTS 4.6.3.1 The containment isolation valves spedf4cd in T:ble 3.5-1 shall be demonstrated OPERABLE prior to returning the valve to service after maintenance,ID 1: 0292 j repair or replacement work is performed on the valve or its associated actuator, control or power circuit by performance of a cycling test, and verification of isolation time.

e. The provisions of Specification 3.0.4 are not applicable provided that the affected penetration is isolated in accordance with ACTION b or c above, and provided that the associated system, if applicable, is declared inoperable and the appropriate ACTION statements for that system are perfonned.

CAUTION: The inoperable isolation valve (s) may be part of a system (s).

Isolating the affected penetration (s) may affect the use of the system (s). Consider the technical specification requirements on the affected system (s) and act accordingly. [ COMANCHE PEAK - UNIT 1 3/4 6-13 ,

TXX-88512 -l ' s AllACHMENT 9 PAGE 18 0F 42 CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 4.6.3.2 Each containment isolation valve :pecified ir, Tabis 3.5-1 shall be ID I: 0293 demonstrated OPERABLE duririg the RCiUCLING M00C cr COLD SiiUT00W-at least onc9 per 18 months by: 10 8: 0294

a. Verifying that on a Phase "A" Isolation test signal, each Phase "A" isolation valve actuates to its isolation position;

h. Verifying that on a Phase "B" Isolation test signal, each Phase "B" isolation valve actuates to its isolation position; and

c. Verifying that on a Containment Ventilation Isolation test signal, each pressure relief discharge valve actuates to its isolation position.

4.6.3.3 The isolation time of each power-operated or automatic valve of-- iD 0 op94 Teble O.C-1 shall be determined to be within its limit when tested pursuant to ~ Specification 4.0.5. l l G 1 . l l l l 1 O COMANCHE PEAK - UNIT 1 3/4 6-14 ' l l

IIX-88512 i

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AliACHMENT9 PAGE 19 Of 42 RE.DCATEw,a TABLE 3.6.1 CONTAINMENT ISOLATION VALVES MAXIMUM ISOLATION TYPE FSAR TABLE TIME LEAK TEST V .VE NO. REFERENCE NO.* LINE OR SERVICE (Seconds) REQUIREMENTS

1. Pnas "A" Isolation Valves 1HV-215 20 Feedwater Sample 5 Note 1 (FW to Stm Gen #1) 1HV-2155 22 Sample 5 Note 1 (FW to Stm Gen #2) 1HV-2399 7 Blowdown From Steam 5 Note 1 Generator #3 1HV-2398 28 Blowdown From Steam 5 Note 1 Generator #2 1HV-2397 29 Blowdown From-Steam 5 Note l' Generator #1 1HV-2400 30 llowdown From Steam 5 Note 1 nerator #4

( 1-8152 32 Let wn Line to Letd n Heat Exchanger 10 C 1-8160 32 Letdown ine to 10 C Letdown H t Exchanger 1-8890A 35 RHR to Cold g Loops 15 Note 2

                                                                     #1 & #R Test L ne 1-88908                   36              RHR to Cold Leg L ops         15       Note 2
                                                                     #3 & #4 Test Line 1-8047                    41              Reactor Makeup Water          10       C to Pressure Relief Tank & RC Pump Stand Pipe 1-8843                    42              !! to RC System Cold          10       Note 2 Leg Loops #1, #2, #3, #4 Test Line 1-8881                     43              SI to RC System Hot           10       No e,2 Leg Loops #2 & #3                          'N Test Line                                     N O

COMANCHE PEAK - UNIT 1 3/4 6-15 .

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        , , , ,     TXX-88512                                            .__. . . -             . - _ .

ATTACHMENT 9 PAGE 20 0F 42 a m o"o E0Mv

                               .     .                                              l q                                                TABLE 3.6.1 (Continued)

CONTAINMENT ISOLATION VALVES ( MAXIMUM ISOLATION TYPE FSAR TABLE TIME LEAK TEST VAL E NO. REFERENCE NO.* LINE OR SERVICE (Seconds) REQUIREMENTS

1. Phase A" Isolation Valves (Continued) 1-8824 ' 44 SI to RC System Hot 10 Note 2 Leg Loops #1 & #4 Test Line 1-8823 45 SI to RC System Cold 10 Note 2 Leg Loops #1, #2, #3,

                                                          & #4 Test Line 1-8100               51            Seal Water Return                  10         C and Excess Letdown 1-8112               51            Sea Water Return                   10         C and Excess Letdown                                     -

1-7136 52 RCOT Heat Exchanger 10 C to Waste Hold Up Tank LCV-1003 52 RCQT Heat Exchanger O' ( to Waste Hold Up Tank 10 C 1HV-5365 60 Demin\ eralized Water 5 C Supply 1HV-5366 60 Demineralifed Water 5 C l Supply . 1HV-5157 61 Containment Sue Pump 5 C Discharge

      =               1HV-5158              61           Containment Sump Pump              5           C Discharge

! 1HV-3487 62 Instrument Air to s5 C l Containment ' l 1-8825 63 RHR to Hot Leg Loops 15' s Note 2 l #2 & #3 Test Line 1 1HV-2405 73 Sample from Steam 5 Note 1 Generator #1 1HV-4170 74 RC Sample From Hot 5 C Legs COMANCHE PEAK - UNIT 1 3/4 6-16 , i

IXX 88512 . . . _ _ . .. -- .._ l ATIACHMENT9 PAGE 21 0F 42 JE,.0CATEmm TABLE 3.6.1 (Continued) CONTAINMENT ISOLATION VALVES MAXIMUM ISCLATION TYPE CSAR TABLE TIME LEAK TEST ALVE NO. REFERENCE NO.* LINE OR SERVICE (Seconds) REQUIREMENTS

1. Pha\e"A"IsolationValves(Continued) 1HV-41 8 74 RC Sample From Hot 5 C Leg #1 1HV-4169 74 RC Sample From Hot 5 C Leg #4 1HV-2406 76 Sample from Steam 5 Note 1 Generator #2 1HV-4167 7 Pressurizer Liquid 5 C Space Sample '

1HV-4166 77 Pressurizer.. Liquid 5 C - Space Sample 1HV-4176 78 Pressurizer Steam 5 C Space Sample ( 1HV-4165 78 essurizer Steam Sp ce Sample 5 C IHV-2407 79 Samp from Steam 5 Note 1 Genera r #3 1HV-4175 80 Accumulat rs 5 C . 1HV-4171 80 Sample frem N 5 C Accumulator # 1HV-4172 80 Sample from 5 C Accumulator #2 1HV-4173 80 Sample from 5 C Accumulator #3 1HV-4174 80 Sample from 5, C Accumulator #4 \ 1HV-7311 81 RC PASS Sample Discharge to RCOT 5 \C

1HV-7312 81 RC PASS Sample 5 CN Discharge to RCOT COMANCHE PEAK - UNIT 1 3/4 6-17 .

IXX-88512 - . . . _ _ . . ATTACHMENT 9 _ , IDI0940 2ABLE 3.6.1 (Continued) CONTAINMENT ISOLATION VALVES ( MAXIMUM ISOLATION TYPE FSAR TABLE TIME LEAK TEST VA E NO. REFERENCE NO.* LINE OR SERVICE (Seconds) REQUIREMENTS

1. Phas "A" Isolation Valves (Continued) 1HV-240 82 Sample from Steam 5 Note 1 Generator #4 1-8871 83 Accumulator Test and 10 C Fill 1-8888 83 Accumulator Test and 10 C Fill 1-8964 83 Accumulator Test and 10 C Fill 1HV-5556 84 Containment. Air PASS 5 C -

Return 1HV-5557 84 Containment Air PASS 5 C q Return O( 1HV-5544 94 R iation Monitoring S le 5 C 1HV-5545 94 Radic ion Monitoring 5 C l Sample l 1HV-5558 97 Containme Air PASS 5 C . Inlet l 1HV-5559 97 Containment A PASS 5 C Inlet 1HV-5560 100 Containment Air P S 5 C l Inlet 1HV-5561 100 Containment Air PASS 5 C Inlet 1HV-5546 102 Radiation Monitoring C Sample Return 1HV-5547 102 Radiation Monitoring 5 C Sample Return 1-8880 104 N2 Supply to 10 C ( Accumulators CCMANCHE PEAK - UNIT 1 3/4 6-18 ,

TXX-88512 ATTACHMEHI9 PAGE 23 0F 42

                                                                                                              "{IDI0940 TABLE 3.6.1 (Continued)

CONTAINMENT ISOLATION VALVES c. MAXIMUM

              \                                                                                      ISOLATION       TYPE
                 \           FSAR TABLE                                                                TIME       LEAK TEST VACVE NO. REFERENCE NO.*                             LINE OR SERVICE              (Seconds) REQUIREMENTS

1. Phase "A" Isolation Valves (Continued) 1-7126 105 H2 Supply to RC Drain 10 C Tank 1-7150 105 H2 Supply to RC Drain 10 C Tank 1HV-4710 111 CC Supply to Excess 5 Note 1 Letdown & RC Drain Tank Heat Exchanger 1HV-4711 112 CC Return from Excess 5 Note 1 Letdown & RC Orain Tank Heat Exchanger 1HV-3486 113 Service Air to 5 C Containment p 1HV-4725 114 'ontainment CCW Drain 5 C Q( nk Pumps Discharge 1HV-4726 114 Con inment CCW Drain 5 C Tank umps Discharge 1-8027 116 Nitroge Supply to PRT 10 C 1-8026 116 Nitrogen S ply to PRT 10 C 1HV-6084 120 Chilled Wate Supply 10 C to Containment oolers 1HV-6082 121 Chilled Water Ret rn 10 C From Containment Coolers 1HV-6083 121 Chilled Water Return 10 C From Containment Coolers 1HV-4075B 124 Fire Protection System 10 C Isolation 1HV-4075C 124 Fire Protection System 10 Isolation \

o< \ COMANCHE PEAK - UNIT 1 3/4 6-19

IXX-88512 ATTACHMENT 9 - -- l PAGE 24 0F 42 IM 09:0 n.. . lip A" [)"'Uun l TABLE 3.6.1 (Continued) CONTAINMENT ISOLATION VALVES (~ MAXIMUM ISCLATION TYPE FSAR TABLE TIME LEAK TEST VALV NO. REFERENCE NO.* LINE OR SERVICE (Seconds) REQUIREMENTS

2. Phase " Isolation Valves 1HV-4708 117 CC Return from ACP's 10 C Motors 1HV-4701 117 CC Return from RCP's 10 C Motors 1HV-4700 18 CC Supply to RCP's 10 C Motors 1HV-4709 119 CC Return From RCP's 10 C Thermal Barrier

~ ~

1HV-4696 119 CC Return From'RCP's 10 C Thermal Barrier

3. Containment Ventilation Isola ion Valves 1HV-5542 58 drogen Purge Supply. 10 C C.

1HV-5543 58 Hyd en Purge Supply 10 C 1HV-5563 58 Hydrog n Purge Supply 10 0 1HV-5540 59 Hydrogen urge Exhaust 10 C 1HV-5541 59 Hydrogen Pu e Exhaust 10 C

1HV-5562 59 Hydrogen Purge Exhaust 10 C l 1HV-5536 1C9 Containment Purg Air 5 C

! Supply l 1HV-5537 109 Containment Purge Ai 5 C Supoly 1HV-5538 110 Containment Purge Air 5 C Exhaust 1HV-5539 110 Containment Purge Air 5 C Exhaust l 1HV-5548 122 Containment Pressure 3 L\ Relief \ COMANCHE PEAK - UNIT 1 3/4 6-20

l

TXX-88512 ' AliACHMENT1 PAGE 25 0F 42 10gg949 ( TABLE 3.6.1 (Continued) CONTAINMENT ISOLATION VALVES MAXIMUM ISOLATION TYPE FSAR TABLE TIME LEAK TEST VAL NO. REFERENCE NO.* LINE OR SERVICE (Seconds) REQUIREMENTS

3. Contai ent Ventilation Isolation Valves (Continued) 1HV-5549 122 Containment Pressure 3 C Relief

4. Manual Valves 1MS-711 4a TDAFW Pump Warm-up N.A. Note 1, 11 Valve 1MS-390 Sa N2 Supply to Steam N. A. Note 1 Generator #1 1MS-387 9a N 2 Supply to Steam N.A. Note 1..

Generator #2 l 1MS-384 13a N2 Supply to Steam N.A. Note 1 Generator #3 1MS-712 17a AFW Pump Warm-up N.A. Note 1, 11 ( Va ve IMS-393 18a N2 Su ply to Steam N.A. Note 1 Genera or #4

1FW-106 20b N2 Suppl to Steam N.A. Note 1 l Generator l

l 1FW-104 22b N2 Supply to team N.A. Note 1 1 Generator #2 l 1FW-110 24 Secondary Samplin N.A. Note 1 1FW-102 24b N2 Supply to Steam N.A. Note 1 l Generator #3 1FW-119 26 Secondary Sampling N.A. Note 1 l 1FW-108 26b Na Supply to Steam H. . Note 1 Generator #4 1-7135 52 RCDT Heat Exchanger to N.A. C Waste Holdup Tank ISF-011 Refueling Water O I 56 Purification to Refueling Cavity N.A. C

COMANCHE PEAK - UNIT 1 3/4 6-21 ,

TH-88512 1

                                                                                                         -        - - -~ ~~

ATTACHMENT 9 - PAGE 26 0F 42

~~

kht

                                                  ~
                                                                                                           .      0940 c                                                                                                                                     DRAFI 6

TABLE 3.6.1 (Continued) CONTAINMENT ISOLATION VALVES ( MAXIMUM ISOLATION TYPE FSAR TABLE TIME LEAK TEST VA VE NO. REFERENCE NO.* LINE OR SERVICE (Seconds) REQUIREMENTS

4. Manuh Valves (Continued)

ISF-012 56 Refueling Water N.A. C Purification to Refueling Cavity 1SF-021 67 Refueling Cavity to N.A. C Refueling Watar Purification Pump ISF-022 6 Refueling Cavity to N.A. C Refueling Water Purification Pump

ISF-053 71 Refueling Cavity N.A. C Skimmer Pump Discharge ISF-054 71 Refueling Cavity N. A. C O( kimmer Pump 0 charge 1HV-2333B 2 MSI Bypass from N.A. Note 1, 6 Steam enerator #1 1HV-2334)V 7 MSIV Byp s from N.A. Note 1, 6 Steam Gen ator #2 . 1HV-2335B 11 MSIV Bypass om N. A. Note 1, 6 Steam Generat #3

      =                     1HV-2336B                  15              MSIV Bypass from                        N.A.           Note 1, 6 Steam Generator #

5. Power-Operated Isolation Valves 1HV-2452-1 4 Main Steam to Aux. FPT N.A. Note 1 From Steam Line #1 1PV-2325 5 Steam Generator #1 N. Note 1 Atmospheric Relief IPV-2326 9 Steam Generator #2 N.A. ote 1 Atmospheric Relief COMANCHE PEAK - UNIT 1 3/4 6-22

TXX-88512 t

      ~ --

ATTACHMENT 9 PAGE 27 0F 42 tt 0940 fS TABLE 3.6.1 (Continued) Q CONTAINMENT ISOLATION VALVES MAXIMUM ISOLATION TYPE FSAR TABLE A TEST VALV NO. REFERENCE NO.* LINE OR SERVICE (Seconds) P.EQUIREMENTS

5. Power-OheratedIsolationValves(Continued)

IPV-2327 13 Steam Generator #3 N.A. Note 1 Atmospheric Relief 1HV-2452-2 17 Main Steam to Aux. FPT N.A. Note 1 From Steam Lir- #4 1PV-232S 1 Steam Generator #4 N.A. Note 1 Atmospheric Relief 1HV-2491A 20a Auxiliary Feedwater N.A. Note 1 tv Steam Generator #1 1HV-24918 20a Auxiliary Feedwater N.A. Note 1" to Steam Generator #1 i 1HV-2492A 22a uxiliary Feedwater N.A. Note 1 ) N Steam Generator #2 (O ( 1HV-2492B 22a Aux liary Feedwater to S am Generator #2 N.A. Note 1 1HV-2493A 24a Auxilia Feedwater N.A. Note 1 to Steam enerator #3 1HV-2493B 246 Auxiliary F dwater N.A. Note.1 to Steam Gen ator #3 1HV-2494A 26a Auxiliary Feedw ter N.A. Note 1 l to Steam Generat #4 1HV-24948 26a Auxiliary Feedwater N.A. Note 1 to Steam Generator # 1-8701B 33 RHR From Hot Leg N.A. Note 5 Loop #4 i 1-8701A 34 RHR From Hot Leg N. Note 5 Loop #1 1-8809A 35 RHR to Cold Leg loops N.A. Note 4

                                                  #1 and #2 1-8809B           36            RHR to Cold Leg loops       N.A.       Note 4
 .h (

s

                                                  #3 and #4 f                                                                                                  s COMANCHE PEAK - UNIT 1              3/4 6-23           .

1

IXX-88512

  . ATTACMENT 9                          ---           -   - - -       -----

PAGE 28 0F 42 f

                    ~  '

in 0949 [) TABLE 3.6.1 (Continurd)

CONTAINMENT ISOLATION VALVES ( MAXIMUM ISOLATION TYPE FSAR TABLE TIME LEAK TEST VALVE NO. REFERENCE NO.* LINE OR SERVICE (Seconds) REQUIREMENTS

5. Power-Opera d Isolation Valves (Continued) 1-8801A 42 High Head Safety N.A. Note 7 Injection to Cold Leg Loops #1, #2, #3, & #4 1-88018 42 High Head Safety N.A. Note 7 Injection to Cold Leg Loops #1, #2, #3, & #4 1-8802A 43 SI Injection to Het Leg N.A. Note 8 Loops #2 and #3 1-88028 44 SI Injection to Hot Leg N.A. Note 8- -

Loops #1 and #4 1-8835 45 SixInjection to Cold N.A. Note 4 Le h toops #1, #2, #3, (9 v and #4 ( 1-8351A 47 SealI(actiontoRC Pump (Loo #1) N.A. C 1-83518 48 Seal Injec ion to RC N.A. C Pump (Loop #2)

1-8351C 49 Seal Injection'to RC N.A. C . Pump (Loop #3) \ 1-83510 50 Seal Injection to RC N.A. C Pump (Loop #4) 1HV-4777 54 Containment Spray to i N.A. Note 3 Spray Header (Tr. B)

1HV-4776 55 Containment Spray to N>A. Note 3 Spray Header (Tr. A) \ 1-8840 63 RHR to Het Leg Loops N.A.\ Note 8

                                           #2 and #3                                 \

1-8811A 125 Containment Recire. N.A. Note 1, 10 Sump to RHR Pumps (Train A) COMANCHE PEAK - UNIT 1 3/4 6-24

TXX-88$12 ATTACHMENT 9 PAGE 29 0F 42 D;'O1 - H . .U:J fa IDI 0940 pM-TABLE 3.6.1 (Continued) W b. CONTAINMENT ISOLATION VALVES C MAXIMUM ISOLATION TYPE FSAR TiBLE TIME LEAK TEST VALVE NO. REFERENCE NO.* LINE OR SERVICE (Seconds) REQUIREMENTS

5. Power-Opera ed Isolation Valves (Continued) 1-88118 126 ~ Containment Recire. N.A. Note 1, 10 Sump to RHR Pumps (Train B) 1HV-4782 127 Containment Recirc. N.A. Note 1, 10

                                                          \                                   to Spray Pumps (Train A) 1HV-4783                          128                                        Containment Recire.              N.A.      Note 1, 10 to Spray Pumps (Train B)
                                                                                                              '                                  ~

6. Check Valves 1-8818A 35 R to Cold Leg N.A. Note 2 l Loco #1 l

( 1-8818B 35 RHR t.o Cold Leg N.A. Note 2 Loop #2

1-8818C 36 RHR to Cold Leg N.A. Note 2 Loop #3 \ , 1-88180 36 RHRtoCold\eg N.A. Note 2 l Loop #4 1-8046 41 Reactor Makeup ster N. A. C to Pressurizer Relief l Tank and RC Pump l Stand Pipe i l 1-8815 42 High Head Safety N.A. Note 2 Injection to Cold Leg Loops #1, #2,

                                                                                            #3 and #4 i

ISI-8905A 44 SI to RC System Hot N.A. Note 2 Leg Loop #1 151-89058 43 SI to RC System Hot N.A. Note 2 Leg Loop #2

O ISI-8905C 43 SI to RC System Hot Leg Loop #2 N.A. Note'2 s COMANCHE PEAK - UNIT 1 3/4 6-25 .

TXX-88512 ' l

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

ATTACHMENT 9 PAGE 30 0F 42

                                                                                                              )
                                                                                    .0L toi o'4o TABLE 3.6.1 (Continued)

CONTAINMENT ISOLATION VALVES k MAXIMUM ISOLATION TYPE FSAR TABLE TIME LEAK TEST VAL E NO. REFERENCE NO.* LINE OR SERVICE (Seconds) REQUIREMENTS

6. Check Valves (Continued) 1S1-890 44 SI to RC System Hot N.A. Note 2 Leg Loop #4 ISI-8819A 45 RC System Cold N. A. Note 2 Leg Loop #1 ISI-8819B 45 SI to RC System Cold N.A. Note 2 Leg loop #2 151-8819C 45 SI to RC System Cold N.A. Note 2 Leg loop #3

151-88190 45 SI to RC Sy"st'em Cold N.A. Note 2 Leg Loop #4 1-8381 46 Charging Line to N.A. f') Regenerative Heat C v( ICS-8368A 47 Exchanger Se\ RC P al Injection to (Loop #1) N.A. C 1C5-83688 48 Seal Injection to N.A. C RC Pump (Loop #2)

1CS-8368C 49 Seal Injection to N.A. C RC Pump (Loop #3) 1C5-83680 50 Seal Injection lo N.A. C RC Pump (Loop #4) 1C5-8180 51 Seal Water Return and N.A. C Excess Letdown \ ICT-145 54 Containment Spray to . .A. Note 3 Spray Header (Tr. B) 1CT-142 55 Containment Spray to N.A. Note 3 Spray Header (Tr. A) 1C1-030 62 Instrument Air to N.A. C Containment 1-8841A 63 RHR to Hot Leg N.A. Note 2 k Loop #2 COMANCHE PEAK - UNIT 1 3/4 6-26 .

IXX-88512 '

                                           ~ - - - ' ~                            ~~

ATTACHMENT 9 PAGE 31 CF 42 _ . I i i, ijF,IDI0940 7' TABLE 3.6.1 (Continued) (N) \ I l CONTAINMENT ISOLATION VALVES ( \

           \                                                                                         i MAXIMUM
             \  s          FSAR TABLE ISOLATION TIME TYPE LEAK TEST VALVE NO. REFERENCE NO.*        LINE OR SERVICE         (Seconds) REQUIREMENTS      '

6. ChechValves(Continued) 1-8841B 63 RHR to Hot Leg N.A. Note 2 Loop #3 151-8968 104 N2 Supply To N.A. C Accumulators 1CA-016 113 Service /.ir to N.A. C Containment ICC-629 11 CC Return from RCP's N.A. C Motors ICC-713 118 CC Supply to RCP's N.A. C

~ '

Motors 10C-831 119 CC Return from RCP's N.A. C Thermal Barrier ( 1CH-024 120 Ch lied Water Supply to ntainment Coolers N.A. C

7. Steam Line Isolation Signal 1HV-2333A 1 MSIV #1 5 Note 1, 9, 12 1HV-2409 3 Drain From in 5 Note 1 Steam Line #1 1HV-2334A 6 MSIV #2 5 Note 1, 9, 12 1HV-2410 8 Drain From Main 5 Note 1 Steam Line #2 1HV-2335A 10 MSIV #3 5 Note 1, 9, 12 1HV-2411 12 Drain From Main 5 Note 1 Steam Line #3 1HV-2336A 14 MSIV #4 5 Note 1, 9, 12 1HV-2412 16 Orain From Main 5 Note 1 Steam Line #4

COMANCHE PEAK - UNIT 1 3/4 6-27 ,

TXX 88512 ' i ATTACHMENT 9 PAGI 32 0F 42 RE.0XE IM 0940 t,A] TABLE 3.6.1 (Continued) j6 . CONTAINMENT ISOLATION VALVES k , MAXIDUM N VALVE NO. FSAR TABLE REFERENCE NO.* LINE OR SERVICE ISCLA110N TIME TYPE LEAK TEST (Segoj,js),s _ REQUIREMENTS

8. Fee'dwater Line Isolation Signal

1HV-2134 19 Feedwater Isolation i Note 1, 12

                           'N                            Steam Generator #1
                              \

1FV-2193 \ 20c Feedwater Bypass 5 Note 1, 12

                                  \                      Line 1HV-2185            ' 20d              Feedwater Isolation                   5           Note 1, 12 Bypass Line 1HV-2135               21              Feedwater Isolation                  5            Note 1, 12 Steam Generator #2 IFV-2194               22c             Feedwater Bypass                     5            Note 1$12 Line 1HV-2186                22d            Feedwater Isolation                   5            Note 1, 12 Bypass Line

( 1HV-2136 23 F Ste ater Isolation Generator #3 5 Note 1, 12 IFV-2195 24c Feedwater Bypass 5 Note 1, 12 ( Line \ s 1HV-2187 24d Feedwater Isolation 5 Note.1, 12 Bypass Line \ l 1HV-2137 25 Feedwater Isol ion 5 Note 1, 12 Steam Generator #4 1FV-2196 26d Feedwater Bypass x 5 Note 1, 12 Line 'N l 1HV-2188 26e Feedwater Isolation ' 5, Hote 1, 12 Bypass Line N

                                                                                                    \
                                                                                                      \

9. Safety Injection Actuation Isolation x 1-8105 46 Charging Line to Regenerative Heat 10 '.

C Exchanger (~ COMANCHE PEAK - UNIT 1 3/4 6-28 -

IXX 88512 3773ggy 9 ---- --- -- PAGE 33 0F 42 k } IDI 0940 TABLE 3.6.1 (Continued) g UlmlI CONTAINMENT ISOLATION VALVES ISOLATION NPE FSAR TABLE TIME LEM TEST LVE NO. LINE OR SERVICE REFERENCEN0f (Seconds) As0VIREMEN!S

10. Reih f Valves '

1-8708 33 RHR From Hot Leg N.A. Note 5 Loop #4 1-87084 34 RHR From Hot Leg P.A. Note 5 Loop #1 1HS-021 Sb Main Steam Safety Valve S.G. #1 [N.A.[ Note 1,12 1MS-022 Sb Main Steam Safety [N.A./ Note 1, 12 Valve S.G. #1 1MS-023 Sb Main Steam Safety [N.A.7 Note 1,12 Valve S.G. #1 1MS-024 Sb Main Steam Safety [N.A.I Note 1, 12 alve S.G. #1 ( 1MS-025 Sb Ma Valv Steam Safety S.G. #1 (N.A./ Note 1, 12 1MS-058 95 Main St . Safety ' [N.A.I Note 1, 12 Valve S. . #2 1MS-059 9b Main Steam S fety '['N.A.] Note.1, 12 Valve S.G. #2 1MS-060 9b Main Steam Safet '

                                                                     /N.A.I Note 1, 12 Valve S.G. #2 1HS-061           9b             Main Steam Safety          /N.A.I Note 1, 12 Valve S.G. #2           '

1MS-062 9b Main Steam Safety ( .I Note 1, 12 Valve S.G. #2 ' 1MS-093 13b Main Steam Safety '(N. A. ] Note 1, 12 Valve S.G. #3 1MS-094 13b Main 3 team Safety bl. A.7 Not 1, 32 Valve S.G. #3 ' ' 1MS-095 13b Main Steam Safety [N.A./' Note 1,12 Valve S.G. #3 O' COMANCHE PEAK - UNIT 1 3/4 6-29 .

IxX-88512

                                                ~           ~                     ~

ATTACHMENT 9 PAGE 34 0F 42 Itl 09'O TABLE 3.6.1 (Continued) CONTAINMENT ISOLATION VALVES MAXIMUM ISOLATION TYPE FSAR TABLE TIME LEAK TEST VA VE NO. REFERENCE NO.* LINE OR SERVICE ( scends) REQUIREMENTS

10. Reli Valves (Continued)

IMS-096 13b Main Steam Safety [N.A. Note 1, 12 Valve S.G. #3 1MS-097 13b Main Steam Safety [N.A.j Note 1, 12 Valve S.G. #3 1MS-129 8b Main Steam Safety (N.A.fNote1,12 Valve S.G. #4 iMS-130 18b Main Steam Safety [N.A.hNote1,12 Valve S.G. #4 I IMS-131 18b Main Steam Safety (N.A.I Note 1, 12 Valve S.G. #4 1MS-132 18b ain Steam Safety (N.A.'fNote1,12 Ive S.G. #4 ' ( 1MS-133 18b Main team Safety Valve .G. #4 fN.A.fNote1,12 1RC-036 41 RMUW to T & RCP N.A. C Standpipe IWP-7176 52 RCDT HX to Wa e N.A. C Holdup Tank 100-430 60 Demineralized Wate N.A. C l Supply IVD-907 61 Cont. Sump Pump H.A. C Discharge IPS-193 80 Sample from H. C Accumulators ICC-1067 114 Containment CCW Orain N.A. l Tank Pump / Discharge ICH-271 120 Chilled Water Supply N.A. C to Cont. Coolers ICH-272 121 Chilled Water Supply N.A. C p, from Cont. Coolers V COMANCHE PEAK - UNIT 1 3/4 6-30 -

I IIX-88512

                                --#          *                                                                          ~

AffACHMENT9 PAEE 35 of 42 IN OHO TABLE 3.6.1 (Continued) TABLE NOTATIONS ; k dentification code for containment penetration and associated isolation-v ives in FSAR Tables 6.2.4-1, 6.2.4-2, and 6.2.4-3.

 <.                                             Note    -

These are closed syst*ms which meet the requirements of NUREG-0800 Section 6.2.4, II.6, paragraph o. These valves are therefore not required to be leak tested. Note 2: T se valves incide containment are part of closed systems outside con ainment which are in service post accident at a pressure in exce of containment design pressure and satisfy single failure criter on. These valves are therefore not required to be leak tested. Note 3: These are losed systems outside containment which are in service post accid t and have a water-filled loop seal on the containment j side of the lves for a period greater than 30 days following the , accident. Th e valves are therefore leak rate testej with water

;                                                            at a pressure o    P,.

Note 4: These ESF valves a normally open and remain open during post - - - accident conditions. Postaccident they are continually pressurized in excess of contai t pressure from an ESF source which meets the L single failure criterio . These valves are therefore not required to be leak tested. I i (% Note 5: An effective fluid seal on ese penetrations is provided by the suction sources to the residu 1 heat removal pumps during and fol- , lowing an accident. In additi , these containment isolation valves are non-automatic, are not requi d to operate postaccident and are located inside containment. Thes valves are therefore not required to be leak tested. j Note 6: All four MSIV bypass valves are locked losed in Mode 1. During Modes 2, 3, and 4, one MSIV bypass valve may be opened provided the other three MSIV bypass valves are lo ed closed and their associated MSIVs are closed. i Note 7: These are parallel ESF valves that are norma 11 closed, but are designed to open during post-accident condition Failure of one valve to open will not prevent system pressurizat on on both sides of both valves in excess of containment pressure. hese valves are therefore not required to be leak tested. l Note 8: These valves located outside containment are normally c sed and , j see a pressure in excess of containment pressure in pnt- ccident ' conditions. A valve stem leakage check will be per'9%d a l quarterly basis to assure no significant stem leakr% suuld ccur in post-accident conditions. Note 9: These valves require steam.to be tested are are thus not require to be tested until the plant is in MODE 3. l

\ f

! COMANCHE PEAK - UNIT 1 3/4 6-31 .

11X-88512 l AliACHMENT9 PAGE 36 0F 42

                                                )-~~'""                       ~~
                                                )

i RE . 0CA..'.

                                                                            .       :ma TABLE 3.6.1 (Continued)

TABLE NOTATIONS Note 10: These valves will hAveswater against them during post-accident conditions to preclude anysrelease of containment atmosphere to the environment. ~ Note 11: These valves are normally locked closed and are open only to warm-up the steam supply lines prior to normal surveillance testing. N Note 12: These valves are included for table completenes the requirements of Specification 3.6.3 do not apply. Instead, th equirements of Specification 3.7.1.1, 3.7.1.5 and 3.7.1.6 apply for n steam safety, valves, main steam isolation valves and feedwater lation valvef,respectively. '- S Oc l t e l . O( COMANCHE PEAK - UNIT 1 3/4 6-32 .

IXX-88512 l -- - ---- . . . - '....' PAGE 37 0F 42 } 1 CONTAINMENT SYSTEMS 3/4.6.4 COMBUSTIBLE GAS CONTROL ,, HYOROGEN MONITORS LIMITING CONDITION FOR OPERATION 3.6.4.1 Two independent containment hydrogen monitor trains (with at least one channel per train) shall be OPERABLE. APPLICABILITY: MODES 1 and 2. ACTION:

a. With one hydrogen monitor train inoperable, restore the inoperable monitor train to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 6 hours. The provisions of 3.0.4 are not applicable.

ggyh With both hydrogen monitor trains inoperable, restore at least one m'onitor b. train to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours. O SURVEILLANCE REQUIREMENTS 4.6.4.1 Each hydrogen monitor shall be demonstrated OPERABLE:

a. At least once per 31 days by performing a channel check, and

Jt)///w ess m.u

b. At least once per 92 days on a staggered, test, basi.s by performing a ;/2..,x1 sequence using sample gas in accordance with the manufac-ttJrer's recommendations and by verifying that the current calibration

                     / constants are contained in the microprocessor data base.

CAfibrafio/L i O COMANCHE PEAK - UNIT 1 3/4 6-33 '

TXX-88512 i

  .             AliACHMENT9 PAGE 38 CF 42 O             CONTAINMENT SYSTEMS ELECTRIC HYDROGEN RSCOMBINERS t

LIMITING CONDITION FOR OPERATION 3.6.4.2 Two independent Hydrogen Recombiner Systems shall be OPERABLE. APPLICABILITY: MODES L and 2. ACTION: With one Hydrogen Recombiner System inoperable, restore the inoperable system to OPCRABLE status within 30 days or be in at least HOT STANOBY within the next 6 hours. Tae. p.-wow. . ci sp a . 't . s he .3.cs = noA whbie. 10 8: 0575 SURVEILLANCE REQUIREMENTS 4.6.4.2 Each Hydrogen Recombiner System shall~b'e demonstrated OPERABLE: '

a. At least once per 6 months by verifying, during a Hydrogen Recombiner

, System functional test, that the minimum heater sheath temperature l increases to greater than or equal to 700*F within 90 minutes. I v Upon reaching 700'F, increase the power setting to maximum power ! for 2 minutes and verify that the power meter reads greater than or equal to 60 kW, and

b. At least once per 18 months by:

1) Performing a CHANNEL CALIBRATION of all recombiner instrumenta-l tion and control circuits,

2) Verifying through a visual examination that there is no

, evidence of abnormal conditions within the recombiner enclosure (i.e., loose wiring or structural connections, deposits of foreign materials, etc.), and l 3) Verifying the integrity of ali heater electrical circuits b/ i performing a resistance to ground test following the above l , required functional test. The resistance to ground for any l heater phase shall be greater than or equal to 10,000 ohms. l l \ a< COMANCHE PEAK - UNIT 1 3/4 6-34 , I

Trr 88517 AtkCHMEif 9 i PAGE 39 0F 42 3/4.6 CONTAINMENT SYSTEMS BASES 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 restriction, 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 p test' to account for possible degradation of the containment leakage barriers () between leakage tests. 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 the overall air lock leakage will not become excessive due to seal damage during the intervals between air lock leakage tests. 3/4.6.1.4 INTERNAL PRESSURE The limitations on containment internal pressure ensure that: (1) the containment structure is prevented from exceeding its design negative pressure differential of 5 psi with respect to the outside atmosphere, and (2) the containment peak press e don not exceed the design pressure of 50 psig duringaLOCA. A* d [46.3]ThemaximumosakpressureexpectedtobeobtainedfromaLOCAeventis psig. The limit of 1.5 psi for, initial positive containment pressure will limit the total pressure to 48.3Jpsig, which is less than design pressure and is consistent with the safety analyses. d COMANCHE PEAK - UNIT 1 8 3/4 6-1 ,

IXX-88512

 .            AllACHMENT9 PAGE 40 0F 42 1

1 h G CONTAINMENT SYSTEMS . BASES ID I: 0485 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. -Measurements--shal4-be-eade-at all listed iccations, whether ID I: 0485 t-by-Hxed Th e m e my tee ye.4 cc eraturetmportatde-instruments, prier tc deMcio;r,' big k'he aversge &ir-a o a \co st ?_ ok the m ec~so re m e nh maAe. aA We LMed \c cW en g 3/4.6.1.6 CONTAINMENT STRUCTURAL INTEGRITY % D '*E fM " "d 'W'M M eM tete.% c temger&re roaaw rement ace.r to e u 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 ,reqW red to ensure that the containment will withstand the maximum pressure of L48.3.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 _ O V 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 valves 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 revents power from being supplied to the valve operator. -- Sge,gogag%,q3 e taeca (em o o roovs The use of the Containment Ventilation System auring operations is restricted to the 18-inch pressure relief discharge isolation valves since_,. fjg enb3 uM4Ac the '8-inch :nd 12-inch valves, the 18-4ch valves are capable of closing during a LOCA or steam line break accident. Therefore, the -S+TEUM5te4 acA SOUN0 0 7 dose guideline of 10 CFR 100 would not be exceeded in the event of an accident during containment venting operation. Oper: tier with one pai* cf these-val-ves-open-wi44-be-14mited t0 90 heur; during : :: lend:r y;;r. The tote 4-t4me-the-containment-purg: (vent) :y; tem 1:010tien valve; m:y be Open 28 0285 during M OES 1, 2, 3, :nd A ": c:lendar ye:r 1: : function of :nticipated need and operating experfence. Only safety-related reasong e.g., containment pressure-control-or-the-reduc +ien of airborne radioeet4vity-to-feeH+ tate personnel-4ccess for-survei' lance :nd maintenance activities, iiiey be Usett-to support-the-additi0n:1 time requests. Only safety-related reasons ushould be used to justify the opening of these isolation valves due4eg,M00Es 1, 2, 3, and 4.in any calendar-year-regardless-of-the ellow ble hcues.\, ,

.n.

.O fE

                                                                      ; eg. e_e nWom M p *\ Awt e cw k tei M sur M COMANCHE PEAK - UNIT 1                        B 3/4 6-2    o.th i b i e 5,

TXX 88512

 .        AliACHMENT9                                   i PAGE 41 0F 42 t

CONTAINMENT SYSTEMS BASE 1 l 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 l seal degradation and will allow opportunity for repair before gross leakage failures could develop. The 0.60 L leakage limit of Specification 3.6.1.2b. shallnotbeexceededwhentheleak$geratesdeterminedbytheleakageintegrity tests of these valves are added to the previously determineo 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 ni,d resultant lower contain-ment leakage rate are consistent with the assump.tions used in the safety . analyses. The Containment Spray System which is composed of redundant trains, pro- ! vides post-accident cooling of the containment atmosphere. However, the Con-tainment Spray System also provides a mechanism for removing iodine from the containment atmosphere and therefore the time requirements for restoring an 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 NaOH is added to the containment spray in the event of a LOCA. The limits on NaOH l volume and concentration ensure a 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 caustic stress I corrosion on mechanical systems and components. The contained solution volume limit includes an allowance for solution not usable because of tank discharge line location or other physical characteristics. These assumptions are con-sistent 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 limits specified for those isolation valves designed to 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. O COMANCHE PEAK - UNIT 1 8 3/4 6-3 .

IXX-88512 AliACHMENT 9 PAGE 42 0F 42 i CONTAINMENT SYSTEMS BASES 3/4.6.4 COMBUSTIBLE GAS CONTROL The OPERABILITY of the equipment and systems required for the detection and control of hydrogen gas ensures that this equipment will be available to maintain the hydrogen concentration within containment below its flammable limit during post-LOCA conditions. Either recombiner unit is capable of con-trolling the expected hydrogen generation associated with: (1) zirconium-water reactions, (2) radiolytic decomposition of water, and (3) corrosion of metals within containment. These Hydrogen Control Systems are consistent with the recommendations of Regulatury Guide 1.7, "Control of Combustible Gas Concen-l trations in Containment Following a LOCA," March 1971, i J i,

e +

oi O COMANCHE PEAX - UNIT 1 B 3/4 6-4

IXX-88512 AllACHMENT 10 PAGE 1 0F $4 O COMANCHE PEAK STEAM ELECTRIC STATION TECHNICAL SPECIFICATION .- . 3/4.7 O t O

TXI-68512-ATTACHMENT 10 CPSES Technical Specifications PAGE 2 0F 54 NRC Draft 2 Markup Section 3/4.7 V,a : Change ID# Justification For Change 0300 The frequency for the Auxiliary Feedwater (AFW) System OPERABILITY should be changed from 31 days to 92 days. This is based on several items, the first being ASME Code, Section XI testing. The IST Plan approved by the NRC specifies the frequency of testing all Class 1, 2 & 3 pumps. In this plan it is explicitly specified that testing of the AFW pumps will be conducted on a quarterly basis. All other ESF pumps (i.e. safety injection pumps, charging pumps, RHR pumps, etc.) are consistent between the IST plan and the Technical Specifications in that they are all tested on a quarterly basis. CPSES plant specific design requires that the AFW pumps be used on plant startups and shutdowns. This operation allows for the flow path to be verified and adds assurance for the proper operation of the system, in the SER (pages 22-40 [GS-2] and 22-45 (GL-2]) it is ex) licitly spelled out that the monthly valve line-up not )e required for CPSES. There are no requirements in NUREG 0800, NUREG 0737 or CPSES SER that specify the frequency of testing for the AFW pumps. Due to CPSES specific design the AFW System (a single train) is inoperable during the performance of this testing. Therefore, by requiring monthly vice quarterly testing, the AFW System is inoperable 3 times more than (]

 '                     any other ESF system during surveillance testing. To minimize the amount of time required that the AFW System has to be inoperable for surveillance testing and to make the testing requirements consistent with ASME Code, Section XI, this change should be made to the CPSES Technical Specifications.

0301 This change to the Action Statement is made to allow post work testing of the MSIV in Modes 2 or 3. The Mode 2 and 3 Action Statement is interpreted to allow continued operations with one MSIV shut, presently, without this clause to allow testing the previously inoperable MSIV, you would never be allowed to enter Mode I without having to go down to Mode 4 to perform post work testing if this testing involved cycling the MSIV, which in most cases cycling the MSly would be required. If this were true there would be no need to have this separate Action Statement for Modes 2 and 3. O

III 88512

               - AllACHMENTIO.          CPSES Technical'!pecifications PAGE 3 of 54                 NRC Oraft 2 Markup Section 3/4.7 r                             -

V Change IDF Justification For Change 0302 This specification was added because the Action Requirement, Applicability and Surveillance Requirements of Specification 3/4.6.3, Containment Isolation valves, were not appropriate for the safety function of-the feedwater isolation valves, feedwater isolation bypass valves, and feedwater preheater bypass valves. These valves are designed to close on an SI signal to prevent excessive cooldown during accident conditions; on a steam generator high-high level to limit the effects of a feedwater malfunction event, and to restrict the mass addition to containment for steamline breaks. i i The Action Requirements are designed around the normal plant operating conditions for CPSES. During HOT STANDBY the steam generators are normally fed via auxiliary feedwater, with the transition flow being initiated through the feedwater preheater bypass valves, and then through the feedwater isolation valves. During plant operations, if the feedwater preheater bypass valves and feedwater isolation bypass valves are maintained closed, they are in the proper condition to meet their accident mitigation function, hence no restriction to plant operations need be applied. Likewise, low power operation with the valves inoperable but closed is permissible since feedwater can be supplied via auxiliary feedwater. Limited opening of a valve while in the action statement is allowed to facilitate post maintenance testing required to demonstrate the valves; OPERABILITY, This plant specific change is similar to that licensed at Callaway. 0305 Action Statement has been modified since the 1.5 feet of sedimentation represents a conservative threshold in terms of loss of Ultimate Heat Sink (VHS) capacity. The impact of achieving this threshold will depend on the time interval between initial operation and the time at which the threshold was exceeded and the rate of change in the sediment build up. How rapidly the corrective and/or remedial action needs to be taken to assure future safe operations can best be determined from the historical data. The existing Action Statement would, more than likely, require excessive costs for expediting and mobilizing dredging equipment and would unnecessarily preclude continued plant power operation to degrees vastly disproportionate to the time dependent threat to safety margins. Since this portion of the specification is plant specific, this should not be considered to be a change to the e standards and it still satisfies the SSER # 1 requirements to have the sediment monitoring program and will report specifically the measures that will be taken to remove the sediment.

   - - , , , e ~,           +

IXX-88512 ATTAcer 10 CPSES Technical Specifications W 4 W $4 NRC Draft 2 Markup Section 3/4.7 / . Change ID# Justification For Change 0309 The requirement for testing following pa,,iting, fire, etc. has been moved from 4.7.7b to 4.7.7c for the following reasons: Poisoning of activated carbon by gaseous contaminates has long been known as the primary reason for degradation of the carbon adsorption efficiency. Coninonly used solvents such as acetone and toluene and welding fume decomposition products are known poisons. The sffects of weathering on carbon have been studied extensively and published in NUREG/CR 2112. The concentration of these poisons required to lower the efficiency below acceptable levels is unknown. The ability of carbon to adsorb radioiodine contaminates is a chemical reaction. If a chemical release has occurred, a carbon sample should be removed from the affected housing and tested in the. laboratory to determine its efficiency and then evaluate its acceptability for use. HEPA filters are installed upstream of the adsorber beds as a filter for air-borne particulates, unaffected by

/G d              gases and with no ability to filter gaseous contaminates, The D.O.P. in-place leak test detects mechanical defects which results in bypass of the filter. Filter service life if based on pressure drop and is changed out accordingly. Gaseous contaminates do not affect filter life, therefore, D.O.P. testing after painting, fire, and/or chemical release is not necessary.

The same philosophy holds true for Halide testing the carbon beds. This test detects mechanical defects, primarily, insufficient compaction of the carbon. In-place testing is not required after a release, only the laboratory test and subsequent evaluation of results. This change is made to make the Technical Specifications consistent with the FSAR. 0310 Each filter train contains a carbon adsorber section 0314 with a HEPA filter bank on the upstream and downstream 0322 side of the adsorber. The intent of the adsorber is 0331 to trap gaseous radioiodine contaminates. The upstream HEPA filter bank is designed to trap any airborne particulate contaminates. O

TXI88512 5 CPSES. Technical Specifications NRC Draft 2 Markup 3ection 3/4.7 f) V Change 10f Justification For Change (cont.) The downstream HEPA filter bank was added to the design of cleanup trains many year ago, primarily to trap any carbon fines that may be released from the ad m rber beds. Some designers felt this would also be a good back-up to the upstream HEPA bank in the event of total failure of that bank. This has since been accepted as not a credible scenario. The installation of two filter banks in one housing was not intended to satisfy any single-failure proof criteria. A high-grade 95% roughing filter would be sufficient to trap carbon fines. The minimum carbon particulate size to be installed in a bed is governed by qualification testing prior to installation. Change-out of the HEPA filters presently installed, to roughing filters, is not being advocated due to seismic and pressure dro) considerations. The downstream HEPA filters tend to 1 ave an exceptionally long service li-fe. . and should only degrade from accidental damage during maintenance activities or loading due to a dirty filter housing. ( This change is made to make the Technical Specifications consistent with the FSAR. 0311 Added footnote to clarify which revision of ANSI N-510 0315 referenced by Regulatory Guide 1.52 will be used. 0323 This is consistent with the commitment made in the 0324 FSAR, Section 1A/B. 0328 0329 0491 0492 0314 See 10# 0310 0315 See 10# 0311 O U

T!X-88512 ' AliACHMENT 10 CPSES Techincal Spccifications PAGE 6 # 54 NRC Draft 2 Markup

Section 3/4.7 Change IDi Justification For Change 0321 The requirement for testing following painting, fire, 0327 etc. has been moved from 4.7.8b to 4.7.8c for the following reasons:

Poisoning of activated carbon by gaseous contaminates has long been known as the primary reason for degradation of the carbon adsorption efficiency. Commonly used solvents such as acetone and toluene and welding fume decomposition

                   . products are known poisons. The effects of weathering on carbon have been studied extensively and published in NUREG/CR 2112. The concentration of these poisons required to lower the efficiency below acceptable levels is unknown.

The ability of carbon to adsorb radiolodine contaminates is a chemical reaction. If a chemical release has occurred, a carbon sample should be removed from the affected housing and tested in the laboratory to determine its efficiency and then evaluate its acceptability for use. ' HEPA filters are installed upstream of the adsorber beds as a filter for air-borne particulates, unaf fected by (]" gases and with no ability to filter gaseous contaminates. The 0.0.P. in-place leak test detects mechanical defects which results in bypass of the filter. Filter service life if based on pressure drop and is changed out accordingly. Gaseous contaminates do not affect filter life, therefore, 0.0.P. testing after painting, fire, and/or chemical release is not necessary. The same philosophy holds true for Halide testing the carbon beds. This test detects mechanical defects, primarily, insufficient compaction of the carbon. In-place testing is not required after a release, only the laboratory test and subsequent evaluation of results. This change is made to make the Technical Specifications consistent with the FSAR. 0322 See ID# 0310 0323 See ID# 0311 O

x TXX-88512 K ATTACHMENT 10 CPSES Technical Specifications PAGE 7 of 54 NRC Oraft 2 Markup

- ~

tle Section 3/4.7~ [NN Change 10# Justification For Change 0324 See 108 0311 0327 See 10# 0321 0328 See 10# 0311 0329 See 10# 0311 0331' See 10# 0310 0336 Technical Specifications Surveillance Requirements 0493 are being relocated to the CPSES Technical Specification Improvement Program. TV Electric believes the inclusion of these Surveillance Requirements is-unnecessary and the information would be more appropriately addressed in the CPSES Technical Specification Improvement Program. Relocation of these Surveillance Requirements is consistent with the guidance provided in the NRC's Interim Policv Statement (52FR3788), February 6, 1987, and the reconimendations of the Westinghouse Owners Group MERITS Program. Priority is given to the relocation of these Surveillance Requirements since the detailed information is not used by the Licensed Operator. The information currently in these Surveillance Requirements is more appropriately maintained in a document subject to TV Electric administrative control and 10CFR50.59 review under the CPSES Technical Specification Improvement Program. This change is similar to that Licensed at Shearon Harris and Seabrook. 0338 Replace "installed in the core" with "prior to installation." The intent of the wording in this surveillance is to cause the leak test to be conducted after the majority of activities in the vicinity of the source / fission chamber, that could cause damage and result in leakage, have been completed. Activities such as construction and maintenance in the area shculd be essentially complete prior to this surveillance being acconplished. A 31 day "prior to" requirement is used to accomplish this. For the original CPSES design, the only fission detectors used were the incore detectors and the

TXX-88512 AffAc W Nr to CPSES Technical Specifications FAGE 8 0F 54 NRC Oraft 2 Markup Sectica 3/4.7 Change 10# AstificationForChange 0338 (cont.) "installed in the core" provision was applicable. However, new Source-Interr:ediate Range Detectors (fission chambers) have been installed in order to meet the qualification requirements of Regulatory Guide 1.97.

                                     .These detectors are ex-core detectors located external to the Reactor Vessel. Therefore, the "installed in the core" does not apply and the surveillance defaults to "31 days prior to being subjected to core flux." There can be however, a significant period of time between completion of construction or reassembly in and around the reactor vessel and the criticality of the reactor (first event of "core flux") thus exceeding 31 days. As written, this surveillance would require repetition in this event despite the fact of the relative inaccessibility of the detectors and minimal possibility of the detector having been damaged. The change as proposed complies with the original intent of this surveillance and will be applicable to the new ex-coro fission chambers at CPSES.

This change is similar to that licensed at Waterford which has ex-core fission chambers. 0340 The UPS HVAC System is designed with two 100's capacity trains that provides the cooling for both units batteries and inverters. Due to the importance of the systems cooled by the UPS HVAC system, it is being added to the ' Techni al Specification to ensure its OPERABILITY during accident conditions. The Action Requirements are based on l allowing sufficient time to perform repair while minimizing the amount of time the system is inoperable. This Action time franie of seven days is consistent with other Technical Specifications presently contained in NUREG-0452. The Surveillance Requirements are sufficient to ensure system OPERABILITY in all required MODES. 0341 The Safety Chilled Water System is a required system to ensure proper plant operation during accident conditions. Placing this system in the Technical Specifications ensures its OPERABILITY during accident conditions, which will allow the cooling of essential equipment to mitigate an accident condition. The rationale for the ACTION requirenents has been placed in the BASES and the Surveillance Requirements are sufficient to ensure system OPERABILITY in all required MODES. This plant specific change is similar to that licensed at Palo Verde, j 0491 See 10# 0311

m 4 g TXI-C512 AHAM 10 CPSES Technical Specifications N 9 # 54 NRC Draft 2 Markup ' Section N4.7 ? h Change 10# Justification For Change 0492 See 10# 0311 0493 See IDi 0336 1 0604 The Bases have been expanded to clarify which valves are required to be tested by Surveillance Requirement 4.7.1.2b.l. This flow path in the original submittal was considered totally passive. During the validation effort it was determined that the Feedwater Split-Bypass Valves V are required to be shut upon an Auxiliary Feedwater initiation in order to meet the requirements of the accident analysis. This is explained in more detail in section 10.4 of the FSAR. 0941 The applicability of this specification is for Modes 1,2 and 3. Specification 3.0.1 states "Complicance with the LC0 contained in the succeeding specifications is required during the Operational Modes or other conditions specified therein:...", this statement is interpretea to ensure the LC0 is met anytime the Applicability Mode is entered. The "other conditions specified" is interpreted to be

                  , conditions ex)1icitly spelled out in the Applicability Statement suc1 as the accumulator specification that s) ells out the specified condition of pressurizer pressure a)ove 1000 psig and in Mode 3.

0943 The steam generator atmospheric relief valves (ARVs) are required to terminate the postulated steam genertor tube rupture accident. The accident analyses indicate that two ARVs on intact steam generators are required to be available to cool the RCS in a timely manner following the occurrence of a steam generator tube rupture accident. The LC0 is written to require all four ARVs to be OPERABLE in order to allow for not being able to use the ARV on the faulted steam generator and an active failure of one of the remaining three ARVS. Because the purpose of the ARVs is to provide for removal of reactor decay heat and the Ri1R system is available in MODE 4, the applicable MODES were selected as MODES 1,2 and 3. The Action Requirements are based on the generic actions that were licensed at Seabrook and South Texas. The Surveillance Requirements are sufficient to ensure the ARVs are OPERABLE in all required MODES, This generic change is similar to that licensed at Seabrook and South Texas. j O

irr-C512 ! ArtACHENT 10 CPSES Technical Specifications ' PAGE 10 0F 54 NRC Draft 2 Markup Section 3/4.7 Chanae 108 Justification For Chaqqg 0945 This Technical Specification is being relocated to the CPSES Technical Specification Improvement Program. TU Electric believes the inclusion of this specification is unnecessary and the information would be more appropriately addressed in the CPSES Technical Specification Improvement Program. Relocation of this specification is consistent with the guidance provided in the NRC's Ir.terim Policy Statement (52FR3788), February 6, 1987, and the recommandations of the Westinghouse Owners Group MERITS Program. Priority is given to the relocation of tu s specification since the detailed information is not used by the Licensed Operator, and requires no immediate action from the Licensed Operator if the Action Statement is applied. The information currently in this specification is more appropriately maintained in a document subject to TU Electric administrative control and 10CFR50.59 review under the CPSES Technical Specification Improvement Program. This change is similar to that Licensed at Shearon Harris, Seabrook and Vogtle. 0946 Delete reference to the Detector Well Temperature since it is bound by the Reactor Cavity Exhaust Temperature which is the location inside containment where the 4ctual temperature is measured. Testing has been performed on the detector wells to prove that with the exhaust near 1500F the detector well temperature is well below 1350F.

IIX 88512 AliACHMEN!10 PAGE 11 of 54 l 1 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE : ( l SAFETY VALVES i LIMITING CONDITION FOR OPERATION l 3.7.1.1 All main steam line Code safety valves associated with each steam generator shall be OPERABLE with lift settings as specified in Table 3.7-2. APPLICABILITY: MODES 1, 2, and 3. ACTION:

a. With four reactor coolant loops and associated steam generators in operation and with one or more main steam line Code safety valves inoperable, operation in MODES 1, 2, and 3 may proceed provided, that within 4 hours, either the ihoperabl.e. valve is restored to OPERABLE .

status or the Power Range Neutron Flux High Trip Setpoint is reduced per Table 3.7-1; otherwise, be in at least HOT STANDBY within the next 6 hours and in 406& SHUTDOWN within the following g hours. nor e

b. The provisions of Specification 3.0.4 are not, applicable. El 090 i

c , SURVEILLANCE REQUIREMENTS l 4.7.1.1 No additional requirements other than those required by Specification 4.0.5. l l COMANCHE PEAK - UNIT 1 3/4 7-1 . l

TII-88512 ATTACHMENT 10 PAGE 12 0F 54 _ . TABLE 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES MAXIMUM NUMBER OF INOPERABLE MAXIMUM ALLOWABLE POWER RANGE SAFETY VALVES ON ANY NEUTRON FLUX HIGH SETPOINT OPERATING STEAM GENERATOR (PERCENT OF RATED THERMAL POWER) 1 87 2 65 3 43 TABLE 3.7-2 . - STEAM LINE SAFETY VALVES PER LOOP VALVE NUMBER LIFT SETTING (* 1%)* ORIFICE SIZE LOOP 1 LOOP 2 LOOP 3 LOOP 4 1MS-021, 058, 093, k29 1185 psig 16 in 2 1MS-022, 059, 094, 130 1195 psig 16 in 2 l 1MS-023, 060, 095, 131 1205 psig 16 in 2 1MS-024, 061, 096, 132 1215 psig 16 in 2 1MS-025, 062, 097, 133 1235 psig 16 in 2 "The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure, COMANCHE PEAK - UNIT 1 3/4 7-2 .

           ,       TXX-88512 AllACHMENT10 PAGE 13 OF 54 PLANT SYSTEMS                                                                          wInnt I

( AUXILI ARY FEED'4ATER 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 powered from separate emergency busses, and

b. One steam turbine-driven auxiliary feedwater pump capable of being powered from an OPERABLE steam supoly system.

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

a. With one auxiliary feedwater pump or associated flow path inoper .

able, restore the required auxiliary feedwater pumps or associated flow path to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours. ( b. With two auxiliary feedwater pumps or associated flow paths inoper-able, be in at least HOT STANDBY witin 6 hours and in HOT SHUTOOWN within the following 6 hours. With three auxiliary feedwater pumps or associated flow paths inop-c. erable, immediately initiate corrective action to restore at least one auxiliary feedwater pump to OPERABLE status as soon as possible. SURVEILLANCE REQUIREMENTS 4.7.1.2.1 Each auxiliary feedwater pump and associated flow path shall be demonstrated OPERABLE: QL -

a. At least once per 41 days on a STAGGERED TEST BASIS by: ID 1: 0300

1) Verifying that each motor-driven pump develops a discharge pressure of greater than or equal to [Later) psig at a flow of greater than or equai to[430]gpm;

2) Verifying that the steam turbine-driven pump develops a dis-charge pressure of greater than or eglual to (Later] psig at a O( flow of greater than or equal t(860,gpm ghen the secondary steamsupplypressureisgreaterthan,532Jpsig. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3; COMANCHE PEAK - UNIT 1 3/4 7-3

IIX-88512 ATTACHMEHi10 PAGE 14 0F 54 PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) (

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

4) Verifying that each automatic valve in the flow path is in the fully open position whenever the Auxiliary Feedwater System is in standby for auxiliary feedwater automatic initiation or when above 10% RATED THERMAL POWER.

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

1) Verifying that each automatic valve in the flow path actuates to its correct position upon receipt of an Auxiliary Feedwater 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. O-c l l . . I l l l ( l COMANCHE PEAK - UNIT 1 3/4 7-4

TXX 88512

   .        ATTACHMENTi0 PAGE 15 0F 54 p        PLANT SYSTEMS d        CON 0ENSATE STORAGE TANK LIMITING CON 0! TION FOR OPERATION 3.7.1.3 The condensate storage tank (CST) shall be OPERABLE with a contained water volume of at least 282,540 gallons (Ldtr % of span) of water.

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

     . With the CST inoperable, within 4 hours either:

a. Restore the CST to OPERABLE status or be in at least HOT STANOBY within the next 6 hours and in HOT SHUTOOWN within the following 6 hours, or

b. Demonstrate the OPERABILITY of the Station Service Weter (SSW) system as a backup supply to the auxiliary feedwater pumps and- -

restore the CST to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTOOWN within the following 6 hours. O( SURVEILLANCE REQUIREMENTS 4.7.1.3.1 The CST shall be demonstrated OPERABLE at least once per 12 hours by verifying the contained water volume is within its limits when the tank is the supply source for the auxiliary feedwater pumps. 4.7.1.3.2 The SSW system shall be demonstrated OPERABLE at least once per 12 hcurs whenever the SSW system is being used as an alternate supply source to the auxiliary feedwater pumps by verifying the SSW system operable and each motor operated valve between the SSW system and each operable auxiliary feed-I water pump is operab11. l ( COMANCHE PEAK - UNIT 1 3/4 7-5 ,

            ~     " '-                                                                    --

CHP i 10 FAGE 16 0F $4 PLANT SYSTEMS SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION , 3.7.1.4 The specific activity of the Secondary Coolant System shall be less than or equal to 0.1 microcurie / gram DOSE EQUIVALENT I-131. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With the specific activity of the Secondary Coolant System greater than 0.1 microcurie / gram DOSE EQUIVALENT I-131, be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours. O SURVEILLANCE REQUIREMENTS k 4.7.1.4 The specific activity of the Secondary Coolant System shall be i determined to be within the limit by performance of the sampling and analysis program of Table 4.7-1. O< COMANCHE PEAK - UNIT 1 3/4 7-6 ,

1 l

     . .. IXX-38512                                   - - - -

ATTACHMENT 10 PAGE 17 Cf 54 TABLE 4 7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY { { SAMPLE AND ANALYSIS PROGRAM TYPE OF MEASUREMENT SAMPLE AND ANALYSIS AND ANALYSIS FREQUENCY

1. Gross Radioactivity At least once per 72 hours.

Determination"

2. Isotopic Analysis for DOSE a) Once per 31 days, when-EQUIVALENT I-131 Concentration ever the gross radio-activity dettermination indicates concentrations greater than 10% of the allowable limit for radiciodines, b) Once par ,5 monthe, when-ever the gross radio- -

activity determination indicates concentrations less than or equal to 10% eg of the allowable limit (g for radioiodines. l l l l l

          *A gross radioactivity analysis shall consist of the quantitative sensurement of the total specific activity of the secondary coolant except for radio-nuclides with half-lives less than 10. minutes. Determination of the contributors to the gross specific activity shall be based upon those energy peaks identifiable with a 95% confidence level.

1 COMANCHE PEAK - UNIT 1 3/4 7-7 . l

                                                  ' " ' ~

TrX88512 AffACHMENT10 PAGE 18 Of 50 - PLANT SYSTEMS MAIN STEAM LINE !$0LATION VALVES ( LIMITING CONDITION FOR OPERATION

3. 7.1. 5 Each main steam line isolation valve (MSIV) shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3. ACTION: MODE 1: With one MSIV inoperable but open, POWER OPERATION may continue provided the inoperable valve is restored to OPERABLE status within 4 hours; oth'erwise be in HOT STANDBY within the next 6 hours and in HOT SHUT 00tiH within the following 6 hours. MODES 2 and 3:

                                                                                           ,101:0301 With one MSIV inoperable, subsequent operation in MODE 2 or 3 may proceed,$

provided the isolation valve is maintained closed. Otherwise, be in HOT STANOBY within the next 6 hours and in HOT SHUTOOWN within the following ( 6 hours. l l w A the a cc c @ .on t,baA t h e l ( V a.\ q e mag be openew a.co e t i ,nw r n Aci +< a. SURVEILLANCE REQUIREMENTS Yor P o d " ' A i " ^" h ^ ^p e r'. e A.h l 4.7.1.5 Each MSIV shall be demonstrated OPERABLE by verifying full closure , within 5 seconds when tested pursuant to Specification 4.0.5. The pvwisions I of Specification 4.0.4 are not applicable for entry into MODE 3. l l \ l COMANCHE PEAK - UNIT 1 3/4 7-8

IXX M 512 ' ATTACMNT 10 PAE 19 0F 54 v PLANT SYSTEMS 108:0302 [ IN FEEDWATER ISOLATION VALVES LIMITING CONDITION FOR OPERATION , 3.7.1.6 Each main feedwater line shall have OPERABLE a feedwater isolation valve, feedwater isolation bypass valve, and feedwater preheater bypass valve. APPLICABILITY: MODES 1, 2, and 3 ACTION: MODE 1: a) With one feedwater isolation valve inoperable, but open, operations may continue provided the inopernble feedwater isolation valve is' restored to OPERABLE '- - status within 4 hours, otherwise be in HOT STANDBY within the next 6 hours; With one or more feedwater isolation bypass valves O' b) inoperable, operations may continue provided each affected feedwater isolation bypass valve is restored to OPERABLE status or closed within 4 hours, otherwise be in HOT STANDBY within the next 6 hours. c) With one or more feedwater preheater bypass valves inoperable, operations may continue provided each affected feedwater preheater bypass valve is restored to OPERABLE status or closed within 4 hours, otherwise be in HOT STANDBY within the next 6 hours. MODES 2 and 3: a) With one or more feedwater isolation valves inoperable, operations may proceed provided the affected feedwater l isolation valve (s) is restored to OPERABLE status or ' closed within 4 hours, except that the valve may be opened as needed for a period of up to 1 hour for post maintenance testing; otherwise be in ROT SHUTDOWN within the next 6 hours. I b) With one or more feedwater isolation bypass valves inoperable, operations may proceed provided the t affected feedwater isolation bypass valve (s) is restored to OPERABLE status or is closed within 4 (~N hours, except the valve may be opened as needed for a l l period of up to 1 hour for post maintenance testingt otherwise be in HOT SHUTDOWN within the next 6 hours, i 3/4 7- fu

in-88512 ATTACHMENT 10 PAGE 20 M 54 N - V IDI 0302 c) With one or more feedwater preheater bypass valves l inoperable, operations may proceed provided the affected feedwater preheater bypass valve (s) is restored to OPERABLE status or closed within 4 hours, except the valve may be opened as needed for a period of up to i hour for post maintenance testing;.otherwise be in HOT SHUTDOWN within the next 6 hours. SURVEILLANCE REQUIREMENTS 4.7.1.6 Each feedwater isolation valve. feedwater isolation bypass valve, and feedwater p;2 heater bypass valve shall be demonstrated OPERABLE by verifling full closure within 5 seconds when tested pursuant to Specification 4.0.5. l lO t l i l l w sse

TXX-88512 ATTACHMENT 10 PAGE 21.06 54 V IDf 0943 PLANT SYSTEMS STEAM GENERATOR ATMOSPHERIC RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.7.1.7 At 1sast four atmospheric relief valves and associated remote manual controls shall be OPERABLE. APPLICABILITY: MODES 1, 2 and 3. ACTION: a) With one less than the required atmospheric steam relief valves OPERABLE, restore the required atmospheric steam relief valves to OPERABLE status within 7 days; or be in at least HOT STANDBY within tha next 6 hours and in HOT SHUTDOWN within the following 6 hours and place the required RCS,'RHR loops in operation for decay heat removal. O () b) With two less than the required atmospheric relief valves OPERABLE, restore at least three atmospheric l relief valves to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours and place the required RCS/RHR loops in operation for decay heat removal. SURVEILLANCE REQUIREMENTS , 4.7.1.7 Each atmospheric relief valve and associated manual controls shall be demonstrated OPERABLE by: l a) At least once per 24 hours by verifying that the nitrogen accumulator tank is at pressure greater than or equal to (80] psig. 1 b) Testing pursuant to Specification 4.0.5. ( l i G 3/4 7- S c

IIX-88512 AfiACH;;EHf 10 PAGE 22 0F 54 PLANT SYSTEMS 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION (- LIMITING CONDITION FOR OPERATION 3.7.2 The temperatures of both the primary and secondary coolants in the steam generators shall be greater than 70*F when the pressure of either coolant in the steam generator is greater than 200 psig. , APPLICABILITY: At all times. ACTION: With the requirements of the above specification not satisfied;

a. Reduce the steam generator pressure of the applicable side to less than or equal to 200 psig within 30 minutes, and

b. Perform an engineering evaluation to determine the effect of the overpressurization on the structural integrity of the stess generator. Determine that the steam generator remains acceptable for continued opt Mion prior to increasing its temperatures above 200'F.

( SURVEILLANCE REQUIREMENTS i 4.7.2 The pressure in each side of the steam generator shall be determined to be less than 200 psig at least once per hour when the tamperature of either l the primary or secondary coolant is less than 70 F. l l l l Oc COMANCHE PEAK - UNIT 1 3/4 7-9 , l

TXX-88512 ATTACHMENI10 PAGE 23 0F 54 PLANT SYSTEMS 3/4.7.3 COMPONENT COOLING WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.3 At least two independent component cooling water loops shall be OPERABLE. APPLICABILITY: H0 DES 1, 2, 3, and 4. ACTION: With only one component cooling water loop OPERABLE, restore at least two loops to OPERABLE status within 72 hours or be in at least. HOT STANOBY witain the nex*. 6 hours and in COLD SHUTOOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS

~~

4.7.3 EachcomponentcoolingwaterloopshalfbedemonstratedOPERA8LE

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

{ correct position; and

b. At least once per 18 months during shutdown, by verifying that:

1) Each automatic valve servicing safety-related equipment actuates to its correct position on its associated engineer safety feature actuation signal, and e

2) Each Component cooling Water System pump starts automatically on a safety injection test signal.

l ( s COMANCHE PEAK - UNIT 1 3/4 7-10 , l

    ,      in 88512 ATIACHMENT10 PAGE 24 0F $4 PLANT SYSTEMS 3/4.7.4 STATION SERVICE WATER SYSTEM c

LIMITING CONDITION FOR OPERATION 3.7.4 At least two independent service water loops shall be OPERABLE. APPLICABILITY: H0 DES 1, 2, 3i and 4. ACTION: With only one service water loop OPERABLE, restore at least two 1 cops to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUT 00WN within the following 30 hours. 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, f power-operated, or autcmatic) servicing safety-related equipment that

 /]
 \,;s is not locked, sealed, or otherwise secured in position is in its correct position; and

b. At least once per 18 months during shutdown, by verifying that:

1) Each automatic valve servicing safety-related equipment actuates to its correct position on a Safety Injection test signal, and

2) Each station service water ty:t:= pump starts automaticaily on a Safety Injection test signal, s

A

 %' (

COMANCHE PEAK - UNIT 1 3/4 7-11

                             *a,                                                   <

r I TXX-88512 },, , , AllACHMENT 10 ) PAGE 25 0F 54 j l PLANT SYSTEMS 3/4.7.5 ULTIMATE HEAT SINK LIMITING CONDITION FCR OPERATION 3.7.5 The ultimate heat sink (VHS) shall be OPERABLE with:

a. A minimum water level at or above elevation 770 Hean Sea Level, USGS datum, and

b. A station service water intake temperature of less than or equal to 102'F, and

c. A maximum average sediment depth of less than or equal to 1.5 feet in the service water intake channel.

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

a. With the# requirements for water level and intake temperature not satisfied, l be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the

   ~                     following 30 hours.

I b. N Mh-th: =:r g: sedi nt d:pth-in-the-seedee-wateMhanne;-greater th:n 1.5 f :t, th: chann:1 : hall be cleaned-w&tMn [30] days-to-reduce-

           / p.         4he-average ::di :nt-depth-to-less-than-015-feett                                        10 I: 0305 SURVEILLANCE REQUIREMENTS 4.7.5 The ultimate heat sink shall be determined OPERABLE:
     =

a. At least once per 24 hours by verifying the station service water intake temi erature and UHS water level to be within their limits.

l .

b. At least once per 12 months by visually inspecting the dam and verifying no abnormal degradation or erosion, and

c. At least once per 12 months by verifying that the average sediment depth in the service water intake channel is less than or equal to

             -                    1.5 feet.

w u. h 6e o.o e me e Se d ' m e d cle fth in We se(6ce & inh kt thu % c\ (.c m mac g' e A UMM rec Ate ~ La n i.5 Iect 3 r e pa vec esad sabmit et- h e 3 (D;g ; q ,y > J - sm go h g,e A paf r w ~t e r-1, s ea t-mm a a s ,a ,_J L w same g3 tg m e,& q,.s, m wwh<a mea sce<s

             , de 5d f 4 i 4 wede r b\We e.m ur w be        e

c. MMsk A -to a es,,dh 6

r. . n e_recnen i v . 9; ., ; n 6., j g Fe;t:c g.m 3/4 -12

(.0MANCHE PEAK - UNIT 1 S.o A are n.+ aphak .

i IXX-88512 I I ATTACHMENT 10 l l PAGE 26 0F 54_ , i l 1

PLANT SYSTEMS 3/4.7.6 FLOOO PROTECTION

   < LIMITING CONDITION FOR OPERATION                                        DRAFT 3.7.6 Flood protection shall be provided for all safety-related systems, components, and structures when the water level of the Squaw Creek Reservoir (SCR) exceeds 777.5 Mean Sea Level, USGS datum.

APPLICABILITY: At all times. ACTION: With the water level of SCR above elevation 777.5 Mean Sea Level, USGS datum, initiate and complete within 2 hours, the flood protect'. . measures verifying that any equipment which is to be opened or is opened for maintenance is isolated from the SCR by isolation valves, or stop gates, or is at an elevation above 790 feet. SURVEILLANCE REQUIREMENTS 4.7.6 The water level of SCR shall be determined to be within the limits by: ( a. Measurement at least once per 24 hours when the water level is below elevation 776 Hean Sea Level, USGS datum, and

b. Measurement at least once per 2 hours when the water level is equal to or above elevation 776 Mean Sea Level, USGS datum) e cL

c. With the water level of SCR above 777.0 Hean Sea Level, USGS datum, verify flood protection measures are in effect by verifying once per 12 hours that flow paths from the SCR which are open for' l maintenance are isolated from the SCR by isolation valves, or stop l gates, or are at an elevation above 790 feet.

t l [ l O< COMANCHE PEAK - UNIT 1 3/4 7-13

TXX18512 ATTACHMENT 10 Pl.GE 27 0F 54

 ^   PLANT SYSTEMS 3/4.7.7 CONTROL ROOM HVAC SYSTEM                                                 .

LIMITING CONDITION FOR OPERATION 3.7.7 Two independent control room HVAC trains shall be OPERA 8LE. APPLICABILITY: All MODES. ACTION: MODES 1, 2, 3 and 4: With one control room HVAC train inope:able, restere the inoperable train to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. MODES 5 and 6:

a. With one control room HVAC train inoperable, restore the inoperable system to OPERABLE status within 7 days or initiate and maintain operation of the remaining OPERABLE control room HVAC train in the emergency recirculation mode.

, b. With both control room HVAC trains inoperable, or with the OPERABLE l control room HVAC trains required to be in the emergency recircula-l tion mode by ACTION a., not capable of being powered by an OPERABLE l emergency power source, suspend all operations involving CORE ALTERATIONS or positive reactivity changes. SURVEILLANCE REQUIREMENTS 4.7.7 Each control room HVAC train shall be demonstrated OPERABLE:

a. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters and charcoal l 4 4 __atisorbers and verifying that the train operates for at least 10 l

continuous hours with the emergency pressurization unit heaters l operating; I l t O l COMANCHE PEAK - UNIT 1 3/4 7-14 , t

TXX-80512

   ,          ATTACHMENT 10                                                                                                               l PACE 28 0F 54                                                                                                               )
                                                                                                                                          )
 ,h          PLANT SYSTEMS O                                                                                                                                         l SURVEILLANCE REQUIREMENTS (Continued)

b. At least once per 18 months or-f1-)-after any structural maintenance on the HEPA filter or charcoal adsorber housings ne (M following_

                          -peinting ,-f4 reree-chemical-releasede-eny-ventd+atdon Icac ::= uni-cating A th the systee by:

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 practdure guidance in Regulatory

( vg ream. ku Positiodhad 5.c4 gnd C.S.d of Regulatory Guide 1.52, l

         ^g U}                   Revisions 2, March 1978, and the emergency filtration unit flow rate is 3000 cfm i 10%, and the emergency pressurization unit flow rate fs 800 cfm i 10%;

ID 1: 0410

2) Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accor-dancewithRegulatorygositionC.6.bofRegulatoryGuide1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revi-s sion 2, March 19787 for a methyl. iodide penetration of less -

than 0.2%; and 10 I: 0309

3) Verifyinghmergencyfiltrationunitflowrateof8000cfm

                                 + 10% and en emergency pressurization unit flow rate'4f 800 cfm t 10% tr%; y t= cperatica when tested in accordance with l

ANSI N510-1975.1980 or to w n O r e or c ke_rn;a!

7.c nt

c. After every 720 hours of charcoal IIMTAA adsorber operatior?, 4NR theNNW. by verYfying,wr Nc within 31 days after removal, that a laboratory ant. lysis of a repre-sentative co Position C... consampleobtainedinaccordancewithRegulatory,7 of Regulatory Guide 1.52, Revision 2, March 1978 ;

meets ths laboratory of Regulatory Guide 1.52, testing criteria2,of Revision March Regulatgry Position 1978, for C.6.a 10 : Om a methyl l iodide penetration of less than 0.2%; '

d. At least once per 18 months by:

1 1) Verifying that the pressure drop across the comb 9ed.,HEPA l filters and charcoal adsorber banks is less thanj].7 finches l Water Gauge while operating the emergency filtration unit at a flow rate of 8000 cfm + 10%, and is less than Gauge while operating the emergency pressuriza/9.25] tion unit at a inches flow rate of 800 cfm i 10%;

2) Verifying that on a Safety Injection, Loss-of-Offsite Power, Intake Vent-High Radiation, or Plant Vent-High Radiation test signal, the train automatically switches into the emergency recirculation mode of operation with flow through the HEPA filters and charcoal r.dsorber banks;

3) Verifying that the emergency pressurization unit maintains the control room at a positive pressure of greater than or equal e A uST 9 5 80- R EO Oall be used. m p\ce ok A NSI t4SID- l'115.

ID 1: 0311 COMANCHE PEAK - UNIT 1 3/4 7-15 , 1 l I

i l TXX-88512

    -        AllACHMENT10                              ;

PAGE 29 0F 54 ' I PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) I to 1/8 inch Water Gauge relative to the adjacent areas, including the outside atmosphere, at a flow rate of less than l or equal to 800 cfm during system operation;

4) Verifying that the heaters in the emergency pressurization i units dissipate 10 + 1 kW when tested in accordance with ANSI '

H510 g ; and

5) Verify ng that on a High Chlorine test signal, the train auto-matically switches into the isolation mode of operation with flow through the emergency filtration HEPA filters and char-coal adsorber banks within 10 seconds, g tk.a. o d rea m 'D 1: 0314

e. Aftereachcompleteorpartialreplacementof4HEhAfilterbankin I the emergency filtration unit (s), by arifying that the unit satis-fies the in place penetration and bypass leakage testing acceptance criteria of less than 0.05% in accordance with ANSI N510-1915 for a 00P test aerosol while operating the unit at a flow rate of(800,0 cfm i 10%;

1980 Its 0315

f. After each complete or partial replacement of a charcoal adsorber l

bank in the 6mergency filtration unit (s), by verifying that the l unit satisfies the in place penetration and bypass leakage testing acceptance criteria of less than 0.05% in accordance with ANSI N510- for a halogenated hydrocarbon refrigerant test gas while ! operating the unit at a flow rate of 8000 cfm z 10%; ,'--;$ ,. , 19 E0 w a psbem

g. After each complete or partial replacement of'a HEPA filter bank in the emergency pressurization unit (s), by verifying that the unit satisfies the in place penetration and bypass leakage testing acceptance criteria of less than 0.05% in accordance with ANSI

tfora00Ptestaerosolwhileoperatingtheunitat'aflowl N510-19]f800cfmi10%;and rate of nao a

h. After each complete or partial replacement of a charcoal allsorber bank in the emergency pressurization unit (s), by verifying that the unit satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05% in accordance with ANSI H510 for a halogenated hydrocarbon refrigerant test gas while opratin the unit at a flow rate of 800 cfm i 10%. ,

I'lP/3 l l O COMANCHE PEAK - UNIT 1 3/4 7-16 ,

IXX 88512 !

   ,           AfiACHMENT 10
                                                     -                         -- 1 PAGE 30 Of 54 PLANT SYSTEMS
 \].s                                                                                                  *"

3/4.7.8 PRIMARY PLANT VENTILATION SYSTEM - ESF FILTRATION tlNITS LIMITING CONDITION FOR OPERATION 3.7.8 Two independent ESF Filtration Units shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With one ESF Filtration Unit inoperable, restore the inoperable ESF Filtration Unit to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REOUIREMENTS 4.7.8 Each ESF Filtration Unit shall be demonstrated OPERABLE:

a. At least once per 31 days on a STAGGERED TEST BASIS by initiating, I from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that each ESF Filtration Unit operates for

( L at least 10 continuous hours with the heaters operating;

b. At least once per 18 months or -(-1-) after any structural maintenance on the HEPA filter or charcoal adsorber housingsr Or (2) f0ll0*g pahthfire, or chemita4-release-in-any-ventMetica Icne cor municat4ng with the syst^- by: 10 : 0321

1) Verifying that each ESF Filtration Unit satisfies the in place penetration and bypass leakage testing acceptance criteria of less then 1.0% by using tha test procedure guidance in Regula-(o Aer $M tory Positions C.S.a C.5 4 and C.5.d of Regulatory Guide 1.52, +

y Revision 2, March 19787 and verifying the flow rate is 15,000 cfm b i 10% per ESF Filtration Unit when tested in accordance with j ANSI N510-19M; and to : 0322 8 ggo ID 1: 0324

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

Revision 2, March 1978, T meets the laboratory testing criteria l of Regulatory Posi, tion C.6.a of Regulatory Guide 1.52, Revi-sion 2, March 1978T for a methyl iodide penetration of less than 1.0%. i 4 - At45T N 6 to - l %') shall be esel in p l ue ed ArnI easio-f 975. ( 10 :: 0323 l COMANCHE PEAK - UNIT 1 3/4 7-17 l l

IXX-88512 . _ . - -

                                                -                                    i ATTACHMEFT10 PAGE 41 0F St
                   ~       '

ko!!odi.n3 pinhIn3 ) Ice oe c.nt6c.a.\ j /LANT SYSTEMS or cd u.s e m ca.nD co <n en o n , e stiy v e.nt; ikkie rt aoNe N3 Ul0 I {V Th.a v e nta d ,,o u;gs 3stenu rt. ge p, , SURVEILLANCE REQUIREMENTS (Continued) ID 1: 0327

c. After every 720 hours of charcoal adsorber operatiorf, by verifying, within 31 days after removal, that a laboratory analysis of a repre-sentativ9 carbon sample obtained in accordance with Regulatory Position C 6.b of Regulatory Guide 1.52, Revision 2, March 1978f ID 1: 0329 meets the laboratory testing criteria of Regulat*ory Position C.6.a o'? Regulatory Guide 1.52, Revision 2, March 1978, for a methyl iodide penetration of les than 1.0%;

d. At least once per 18 months by:

1) Verifying that the pressure drop across the combi ed 4 EPA filters and charcoal adsorber banks is less than 8.25jinches Water Gauge while operating each ESF Filtration nit at a flow rate of 15,000 cfm 1 10%,

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

3) Verifying that the heaters dissipate 100 1 5 kW when tested in i -

accordance with ANSI N510 -197&.1920 g i y

e. After each complete or partial replacement of -e'HEPA filter bank, "

by verifying that the associated ESF Filtration Unit satisfies the 2 in place penetration and bypass leakage testing acceptance criteria ; of less than 1.0% in accordance with ANSI H510-19 N for a DOP test .q g e

 /G                  aerosol while operating the associated ESF Filtration Unit at a                   'l C                   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 satis-fies the in place penetration and bypass leakage testing acceptance l criteria of less than 1.0% in accordance with ANSI N510-197hfor a ,q ge l

halogenated hydrocarbon refrigerant test gas while operating'the associated ESF filtration unit y:t= at a flow rate of 15,000 cfm l 1 13. l l l l \ \ l l

        #-     ANSI         M 510 - WEo            snad       be  osed, m 7 ctce   \

cd ANSI $ i a sto- ms = l l a l l COMANCHE PEAK - UNIT 1 3/4 7-18 , l l

l

          ._..                        .. _ ._____   _ _ . . _ . -     _      . . .         . _ _ .        i IXX-88512 ATTACHP.ENT 10 PAGE 32 0F 54-O            PLANT SYSTEMS V            3/4.7.9 SNUBBERS                                                                                     f" l                  LIMITING CONDITION FOR OPERATION 3.7.9 All snubbers shall be OPERABLE. The cnly snubbers excluded from the requirements are those installed on nonsafety-r61ated systems and then only if their failure of failure of the system on which they are installed would have no adverse effect on any safety-related system.

APPLICABILITY: MODES 1, 2, 3, and 4. MODES 5 and 6 for snubbers located on systems required OPERABLE in those MODES. y proadib the a 9wed augu.cded McNice ACTION: r i w ar.c e r Aa n c.e t .i

                                          ,epe With one/ or more snu%vonbers inoperable on any systen, within 72 hours replace or re-store the inoperable snubber (s) to OPERABLE status and perform an engineering eval-uationher Specirfkation 4.7 4. on the attached component or declare the attached system inoperable and follow the appropriate ACTION statement for that system.

SURVEILLANCE REOUIREMENTS 4.7.9 Each snubber shall be demonstrated OPERABLE by performance of the#reto;nmeds following augmented inservice inspection program.in cddition tc the require ,ct,% mettts of Specificat.iei 4.0.57 aMa" N.InspectionTypes

                                  .As used in this specification, type of snubber shall mean snubbe
                                  'of the same design and manufacturer, irre:pective of capacity.

O' b. Vis'ba! Inspections ID 1: 0336 Snubberc.are categorized as inaccessible or accessible uring reactor operatf orb. Each of these groups (inaccessible and cessible) may be inspectedsindependently according to the sche e below. The first inservice visual inspection of each type f snubber shall be performed after 4 months but within 10 mont of commencing POWER OPERATION and shalk. include all snubbers. f all snubbers of each type are found OPERABLE during the fir inservice visual inspection, the second inservice visual inspecti shall be performed at the I first refueling outage. Othuvis , subsequent visual inspections shall be performed in accordan e with the following schedule: No. of Inoperable Snubbers Each Type Subsequent Visual per Inspection Period en aw sw W rt- Inspection Period * **' l 0 N 18 months t 25% 1 12 months i 25% 2 N 6 months i 25% 4 N 124 days t 25% 5,6,7 \ 62 days t 25% 8 or more \31 days 125% 1 l

                   *The inspecti            interval for each type of snubber shall not b(lengthened l                     more than ne step at a time unless a generic problem has beensidentified and cor eted; in that event the inspection interval may be lengthened one step- e first time and two steps thereafter if no inoperable snubbers of O(              t         type are found.
               /TheprovisionsofSpecification4.0.2arenotapplicable.
                                                                                                                    's COMANCHE PEAK - UNIT 1                               3/4 7-19            ,

TXX-88512 ATTACHMENT JG . PAGE 33 0F 54 s PLANT SYSTEMS tjRVEILLANCE REQUIREMENTS (Continued) ( c. Visual Inspection Acceptance Criteria IM 03u Visual inspections shall verify that: (1) there are no visible / tindications of damage or impaired OPERABILITY, (2) attachments (c the foundation or supporting structure are functional, and (3 fasten-ers for attachment of the snubber to the component and to t snubber anchorage are functional. Snubbers which appear inoperabi as a result of visual inspections may be determined OPERABLE fpr the purpose of establishing the next visual inspection interfal, provided that: (1) the cause of the rejection is clearly estab)ished and remedied for that particular snubber and for other sodbbers irrespec-tive of type that may be generically susceptible; and (2) the affected snubberisfunctionallytestedintheas-foundcon/itionanddetermined OPERABLE per Specification 4.7.9f. All snubbers connected to an inoperablecommor,hydraulicfluidreservoirshaibecountedas inoperable snubbers. N

d. Transient Event Inspection s .-

An inspection shall be performed of all nubbers attached to sections of systems that have experienced unexp ted, potentially damaging transients as determined from a revie of operational data and a m visual inspection of the systems wi in 6 months following such an event. In addition to satisfying ) e visual inspection acceptance ( i s' l criteria, freedom-of-motion of inethanical snubbers shall be verified using at least one of the folloying: (1) manually induced snubber movement;or(2)evaluationofin-klacesnubberpistonsetting;or

(3) stroking the mechanical stiubber' hrough its full range of travel,

e. Functional Tests During the first refueling shutdown and at least once per 18 months thereafter during shu%own, a representativ sample of snubbers of each type shall be tested using onw of the fo lowing sample plans.

, The sample plan for/each type shall be selecte prior to the test l period and cannot be changed during the test period. The NRC Regional l Administrator s 11 be notified in writing of the for each snubber type prior to the test periodthe or\sample ampleplanplan selected used in the p,r'ior test period shall be implemented: 1 1) At legit 10% of the total of each type of snubber hall be h func onally tested either in place or in a bench t t. For eac snubber of a type that does not meet the functional test as optance criteria of Specification 4.7.9f., an addittonal 10% pf that type of snubber shall be functionally tested unt'Q no more failures are found or until all snubbers of that typ have been functionally tested; ur ( COMANCHE PEAK - UNIT 1 3/4 7-20 ,

IXX-88512 ATTACHMENT 10 PAGE 34 0F 54 PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) g

e. Functional Tests (Continued)

              \      2)     A representative sample of each type of snubber shall be func-tionally tested in accordance with Figure 4.7-1. "C" is thef total number of snubbers of a type found not meeting the accept-ance requirements of Specification 4.7.9f. The cumulative'
                      \ s number       of snubbers of a type tested is denoted by "N". At the end of each day's testing, the new values of "N" and "C" (pre-vious day's total plus current day's increments) shall' be plottedonFigure4.7-1. If at day time the point p, lotted falls in the "Reject" region, all snubbers of that type shall be fogetionally tested. If at any time the point. plotted falls in the\"Accept" region, testing of snubbers of that type may be terminated. When the point plotted lies in the'"Continue Testing"Negion,additionalsnubbersofthat,typeshallbe tested unti,1 the point falls in the "Accept"/ region or the "Reject"reKon,orallthesnubbersofth,attypehavebeen tested; or            i                                               /
                                                   \                           .
                                                                                                /                 ~     .

3) An initial repre entative sample of 55/ snubbers shall be func-tionally tested. or each snubber type which does not meet the functional test ac(ceptance criteria / another sample of at least one-half the size o the initial 'mple shall be tested until f- the total number test is equal o the initial sample size

( multiplied by the facto , 1 + C 2, where "C" is the number of snubbers found which do t t the functional test acceptance criteria. The results fr' is sample plan shall be plotted using an "Accept" line whi follows the equation N = 55(1

                           + C/2). Each snubber poi                     hould be plotted as soon as the snubber is tested. If a poiht plotted falls on or below the "Accept" line, testing ff that                          e of snubber may be terminated.

If the point plotted 11s above he "Accept" line, testing must continue until he point fall in the "Accept" region or all the snubbers o thattypehavebgentested. Testing equipment fa)' ure during functional esting may invalidate that day's testing And allow that day's testi to resume anew at a later time providyl all snubbers tested with t failed equipment during the day of equi nt failure are ratested. The ,epresentative sample selected for t functional test sample plans sha be randomly selected from the snub rs of each type and reviewed before eginning the testing. The review s all ensure, as far as practicable, that hey are represen-tative of e various configurations, operating envirogments, range of size, and capacity of snubbers of each type. Snubbers Alaced in the same lo tion as snubbers which failed the previous funct onal test shall p6 retested at the time of the next func+.ional test ut shall not b4 included in the sample plan. If during the function 1 testing, add}(tonal sampling is required due to failure of only one ty)e of ( ( snubber, the functional test results shall be reviewed at that ime tddetermineifadditionalsamplesshouldbelimitedtothety}peof

                 /nubberwhichhacfailedthefunctionaltesting.

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

IXX-88512 ATTACHMENT 1.0 .

              \    PAGE 35 0F 54 pt 0336 5URVEILLANCE REQUIREMENTS (Continued)

C Functional Test Acceptance Criteria s The snubber functional test shall verify that:

                               '1)        Activation (restraining action) is achieved within the              /
                                   \ \

specified range in both tension and compression; /

2) \ Snubber bleed, or release rate where required, is present in b,oth tension and compression, within the specified range;

3) For mechanical snubbers, the force required to initiate or maintain motion of the snubber is within the specified range in both directions of travel; and

4) For snubbers specifically required not to displace under continuous load, the ability of the snubber,to withstand load without displacement, s /

Testing methods may be used to measure parameters indirectly or parameters other than those specified.if those results can be - . correlated to the specified parameters through established methods.

g. FunctionalTestFailuheAnalysis j An engineering evaluation \shall be',made of each failure to meet the functional test acceptance' criteria to determine the cause of the

( failure. The results of this evaluation shall be used, if applicable, in selecting snubbers to be tasted in an effort to determine the OPERABILITY of other snubbers irrespective of type which may be subject to the same failure modea ! For the snubbers found inoperable, an engineering evaluation shall

              .                 be performed on the components to which the inoperable snubbers are attached. The purpose'of this engineering evaluation shall be to l

determine if the components to which the inoperable snubbers are l attached were adversely affected by the inoperability of the snubbers in order to ensure that the component remains capable of meeting the designed servicei 'N Ifanysnubberselectedforfunctionaltestindseither fails to lock up or . fails to move, i.e., frozen-in place; the cause will be evaluated.a'nd, if caused by manufacturer or desigo deficiency, all snubbers tionally'of the same tested. Thistype testingsubject to theshall requirement same.defe~c.t be\ independentshallofbe func-the requirements stated in Specification 4.7.9e. for (nubbers not meet.ing the functional test acceptance criteria. i

                                  .s-.-                                                                \
                             ,/                                                                          \

O< COMANCHE PEAK - Uh!T 1 3/4 7-22 ,

TXX-68Sl2 ATTACHMENT 10 PAGE E OF 54 PLANT SYSTEMS O.k. 54RVEILLANCE REQUIREMENTS (Continued)

DI OD h., Functional Testing of Repaired and Replaced Snubbers S'nubbers which fail the visual inspection or the functional tes acceptance criteria shall be repaired or replaced. Replacepefit snubbers and snubbers which have repairs which might affect the functional test results shall be tested to meet the funct.ional test criteria before installation in the unit. Mechanica/snubbersshall have met the hcceptance criteria subsequent to therf'r raost recent service, and the NQeedom-of-motion test must hjve been performed within 12 months before being installed in the unit.

i. Snubber Service Program Life \

The service life of hydraul n s'e chanical snubbers shall be monitored to ensure that the s ce life is not exceeded between surveillance inspections. T maximum expected service life for various seals, springs, a other crlt(cal parts shall be deter-mined and established ba ed on engineering information and shall be, extended or shortenejk6ased on monitored t'ast results and failure history. Critica arts shall be replaced hat the maximum" sarvice life wi) not be exceeded during a per M when the snubber is required be CPERABLE. The parts replacement shall be docu-mented an he documentation shall be retained in (ac rdance with

    .                  Specificattien 6.10.3.

l l l l l COMANCHE PEAX - UNIT 1 3/4 7-23 . i

I

         +..                                                              t T11-88512 AllACHMENT10 PAGE 37 0F 54-Its 0336

( DRAFT l ( l f l l l

l i l l i -FIGURE 4. 7-1

                                -$AMPtP N 2) TOR SNU0SER TUNCTICHAL TEST COMANCHE PEAK - UNIT 1               3/4 7-24   ,

TXX-08512

      -                                             ~                            '
  .         ATTACHMENT 10 PAGE 38 0F 54 PLANT SYSTEMS

( x3/4.7.10 SEALED SOURCE CONTAMINATION [0945 LIM ING CONDITION FOR OPERATION 3.7.10 ch sealed source containing radioactive material either in excess of 100 micro ries of beta and/or gama emitting material or 5 microcuries of alpha emitting ma rial shall be free of greater than or equal to 0.005 microcurie of removable ontamination. APPLICABILITY: t all times. ACTION:

a. With a seal source having removable contamination in excess of the above limits, immediately withdraw the sealed source from use and l either:

1. Decontaminat and repair the sealed source, or

2. Dispose of the aled source in accordance with Commission Pegulations.

q b. The provisions of Specif ations 3.0.3 and 3.0.4 are not applicable. O{ SURVEILLANCE REQUIREMENTS 4.7.10.1 Test Requirements - Each sealed so ce shall be tested for leakage and/or contamination by:

a. The licensee, or

b. Other persons specifically authorized by e Comission or an Agreement State.

l The test method shall have a detection sensitivity of at least 0.005 microcurie per test sample, 4.7.10.2 Test Frequencies - Each category of sealed sources excluding startup sources and fission detectors previously subjected to ore flux) shall be tested at the frequency described below,

a. Sources in use - At least once per 6 months for all seal sources j containing radioactive saterials:

1) With a half-life greater than 30 days (excluding Hydrogeq 3),

and \ ( 2) 'In any form other than gas. COMANCHE PEAK - UNIT 1 3/4 7-25 , ! l l l

TXX-88512 _ _ . . _ *

~~

ATTACHE WT 10 PAGE 39 0F 54 I PLANT SYSTEMS

                                                                                        -        101 OMS SURVEILLANCE REQUIREMENTS (Continued)

( .

b. Stored sources not in use - Each sealed source and fission detector shall be tested prior to use or transfer to another licensee unless tested within the previous 6 months. Sealed sources and fission detectors transferred without a certificate indicating the last test date shall be tested prior to 'eing o placed into use; and pe:ec % Wdstut;en o r

c. artup sources and fission detectors Each sealed startup source an fission detector shall be tested ithin 31 days prior to being i

sub (cted to core flux or intalled ia the core and following repair i or matotenance to the source. 4.7.10.3 Reports - report shall be prepared and submitted to the Commission on an annual basis if aled source or fission detector leakage tests reveal the presence of greate han or equal to 0.005 microcurie of removable contamination. l O c l

l N l N 1 , I l t' l V) ( COMANCHE PEAK - UNIT 1 3/4 7-26 , l

I IXX-88512 - AITACHMENT 10 PAGE 40 0F 54 i j l PLANT SYSTEMS 3/4.7.11 AREA TEMPERATURE MONITORING LIMITING CONDITION FOR OPERATION 3.7.11 The maximum temperature limit for normal conditions of each area shown in Table 3.7-3 shall not be exceeded for more than 8 hours and the maximum , temperature for abnormal conditions of each area given in Table 3.7-3 shall not I be exceeded. APPLICABILITY: Whenever the equipment in an affected area is required to be OPERABLE. ACTION:

a. With one or more areas exceeding the maximum temperature limit (s) for wel -+ 27. mel conditions shown in Table 3.7-3 for more than 8 hours, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that provides a record of'the cumulative time and the amount by which the temperature in the affected area (s) exceeded the limit (s) and an analysis to demonstrate the continued OPERABILITY of the affected equipment. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

b. With one or more areas exceeding the maximum temperature limit (s) for abnormal conditions shown in Table 3.7-3, prepare and submit a Special Report as required by ACTION a. above and within 4 hours either restore the area (s) to within the maximum temperature limit (s) for abnormal conditions, or

1) Declare equipment in the affected area (s) INOPERABLE; or,

2) Verify that the qualification envelope for the affected equip-ment has not been exceeded, or declare the affected equipment which exceeded the qualification envelope INOPERABLE; or,

3) Perform an analysis that justifies continued operation.

SURVEILLANCE REQUIRENENTS 4.7.11 The temperature in each of the areas shown in Table 3.7-3 shall be determined to be within its limit at least once per 12 hours. O COMANCHE PEAK - UNIT 1 3/4 7-27

      ,    IXX-88512 ATTACHMENT 10 PAGE 41 0F 54 TABLE 3.7-3
    \                                       AREA TEMPERATURE MONITORING MAXIMUM AREA                                              TEMPERATURE LIMIT (*F)

Normal Abnormal Conditions Conditions

1. Electrical and Control Building Normal Areas 104 [122]

Control Rooms UPS/ Battery Rooms [80] 104 [80] (104]

2. Fuel Builrting Normal Areas 104 [122]

Spent Fuel Pool Cooling Pump Rooms 122 (122]

3. Safegaurds Building

                   ' Normal Areas                                     104             (122]

AF, RHR, SI, Containment Spray Pump Rooms 122 [122] Diesel Generator Area 129 [129] Day Tank Room 122 [122]

4. Auxiliary Building

[ Normal Areas 104 (122] CCW, CCP Pump Rooms 122 (122] l 5. Service Water Intake Structure 127 [127]

         . 6.      Containment Building General Areas                                     120             [120]

CRDM Platform 140 [140] Cetector 'd;il/ Reactor Cavity Exhaust [175] ItslON R.C. Pipe Penetrations +3(015o] 20 [200] l CRDM Shroud Exhaust 163 [163] l l 'O C - ceE P m . U m 1 3,4 7 2e .

l l TXX-88512 ,

ATTACHMENT 10 1 PAGE 42 0F 54 1

+.

O PLANT SYSTEMS it 3/4.7.F+ UPS HVAC SYSTEM OPERATING LIMITING CCNDITION FOR OPERATION il 3 . 7 . b& Two independent UPS HVAC trains shall be OPERABLE: APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With only one UPS HVAC train OPERABLE, restore the inoperable system to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours and in COLD " - SHUTDOWN within the following 30 hours.

   ,.      SURVEILLANCE REQUIREMENTS il 4 . 7 . Ft.1 Each UPS HVAC train shall be demonstrated OPERABLE at least once per 18 months by:

a) Verifying that each UPS HVAC train starts automatically on a Safety Injection test signal. b) Verifying that each UPS HVAC train starts automatically on a Blackout test signal. 12. 4.7.14.2 Each UPS HVAC train shall be demonstrated OPERABLE at least once per 31 days by starting the non-operating UPS HVAC train and verifying that the train operates for at least 1 hour. O jy 7 .19

IXX-88512 ATTACHMENT 10' PAGE 43 0F 54 j-

 \)                                     1 PLANT SYSTEMS 10 1: 0341 i3 3/4.7.19 SAFETY CHILLED WATER SYSTEM LIMITING CONDITION FOR OPERATION O

3.7.15 At least two independent safety chilled water trains shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With only one safety chilled water train OPERABLE: a) Within 1 hour verify that the normal HVAC system is providing space cooling to the vital power distribution rooms that depehd'on the inoperable safety chilled water train for space cooling, and I b) Within 8 hours verify OPERABILITY of the Emergency Core Cooling Systems that depend on the remaining OPERABLE safety chilled water train (one train each of high, intermediate, and low head Safety Injection and auxiliary feedwater), and c) Within 24 hours verify OPERABILITY of all required systems, subsystems, trains, components and devices that depend on the remaining OPERABLE safety chilled water train l for space cooling, and d) Restore the inoperable train to OPERABLE status within 7 days. If any of these conditions are not satisfied within the specified time, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. ! SURVEILLANCE REQUIREMENTS (3 4.7.tS.1 Each safety chilled water train shall be

demonstrated OPERABLE at least once per 31 days by verifying l that each valve (manual, power-operated, or automatic) servicing safety-related equipment that is not locked, I sealed, or otherwise secured in position, is in its correct l position. 1 3/'l 7-30 .

IXX-88512 ATTACHMENT 10 a PAGE 44 0F 54 (3 D tti Mu 13 4.7.16.2 Each safety chilled water pump shall be tested pursuant to Specification 4.0.5. I3 4.7.tS.3 Each safety chilled water train shall be demonstrated OPERABLE at least once per 18 months by: a) Verifying.that each safety chilled water train pump and vital power distribution room emergency fan coil units start on a Safety Injection test signal, b) Verifying that each safety chilled water train pump and vital power distribution room emergency fan coil units start on a Blackout test signal. O I l i i l l l l O 3/47-3l

IXX-88512 - ATTACHMENT 10 PAGE 45 0F 54 3/4.7 PLANT SYSTEMS U..... p%) BASES 3/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAFETY VALVES The OPERABILITY of the main steam line Code safety valves ensures that the Secondary System pressure will be limited to within 110% (1305 psig) of its design pressure of 1185 psig during the most severe anticipated system opera-tional transient. The maximum relieving capacity is associated with a Turbine trip from 100% RATED THERMAL POWER coincident with an assumed loss of condenser heat sink (i.e., no steam bypass to the condenser). The specified valve lift settings tnd relieving capacities are in accor-dance with the requireme.nts of Section III of the ASME Boiler and Pressure Code, 1974 Edition. The total rated relieving capacity for all valves on all of the steam lines is 18,190,884 lbs/h which is 120% of the total secondary steam flow of 15,140,106 'bs/h at 100% RATED THERMAL POWER. STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis of the reduction in Secondary Coolant System steam flow and THERMAL POWER required by the reduced Reactor Trip settings of the Power Range Neutron Flux channels. The (T Reactor Trip Setpoint reductions are derived on the following bases: For four loop operation p ,(X) - (Y)(V) X x (109) Where: ! SP = Reduced Reactor Trip Setpoint in percent of RATED THERMAL POWER, V = Maximum number of inoperable safety valves per steam line, 109 = Power Range Neutron Flux-High Trip Setpoint, X = Total relieving capacity of all safety valves per steam line in 1bs/ hour, and Y = Maximum relieving capacity of any one safety valve in lbs/ hour l lO l l COMANCHE PEAK - UNIT 1 B 3/4 7-1 , l

IXX-CBS12 ATTACHMENT 10 PAGE 46 0F 54 (] PLANT SYSTEMS 1

 \._)

BASES 3/4.7.1.2 AUXILIARY FEE 0 WATER SYSTEM The OPERABILITY of the Auxiliary Feedwater System ensures that the Reactor Coolant System can be cooled down to less than 350'F from normal operating conditions in the event of a total loss-of-offsite power. Each electric motor-driven auxiliary feedwater pump is capable of deliver-ing a total feedwater flow of[_430]gpm to two steam generators at a pressure of 1221 psig to the entrance of the steam generators. The steam-driven auxiliary feedwater pump is capable of delivering a total feedwater flow of[860Jgpm to four steam generators at a pressure of 1221 psig to the entrance of the steam .N generators. This capacity is sufficient to ensure that adequate feedwater flow is available to remove decay heat and reduce the Reactor Coolant System temp-erature to less than 350'F when the Residual Heat Removal System may be placed into operation. Thy Auxiliary Feedwater System is capable of delivering a total feedwater flow of _430Jgpm at a pressure of 1221 psig to the' entrance of at least two-steam generators while allowing for: (1) any possible spillage through the design worst ;ase break of the main feedwater lir.e; (2) the design worst case single failure and (3) recirculation flow. This capacity is sufficient to ensure that cAequate feedwater flow is available to remove decay heat and Q reduce Reactor Coolant System temperature to less than 350*F at which point the Residual Heat Removal System may be placed in operation. ( 3/4.7.1.3 CONDENSATE STORAGE TANK gfA)$& RT $ IM %D* The OPERABILITY of the condensate storage tank with the minimum water volume ensures that sufficient water is available to maintain the RCS at HOT STANOBY conditions for 18 hours with steam discharge to the atmosphere concur-rent with total loss-of-offsite power or 4 hours at HOT STANDBY followed by a cooldown to 350*F at a rate of 50 F/HR for 5 hours. The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics. HUREG-0737, Item II.E.1.1 requires a backup source to the CST which is the

   ,      CPSES Station Service Water System, which can be manually aligned, if required.

3/4.7.1.4 SPECIFIC ACTIVITY The limitations on Secondary Coolant System specific activity ensure that the resultant offsite radiation dose will be limited to a small fraction of 10 CFR 100 dose guideline values in the event of a steam line rupture. This dose also includes the effects of a coincident 1 gpm primary-to-secondary tube leak in the steam generator of the affected steam line. These values are cor.sistent with the assumptions used in the safety analyses. O COMANCHE PEAK - UNIT 1 8 3/4 7-2 ,

I t IXX-88512 j ATTACHMENT 10 INSERT B PAGE 47 0F 54 _ , The auxiliary feedwater flow path is a passive flow path based on the fact f hat valve actuation is not required in order to supply flow to the steam generators. The automatic valves tested in the flow path are the Feedwater Split Flow Bypass which are required to be shut upon initiation of the Auxiliary Feedwater System to meet the requirements of the accident analysis. O O

IIX-88512

. ATTACHMENT 10 PAGE 48 0F 54

+o P' ANT SYSTEMS DR*\FT BASES 3/4.7.1.5 MAIN STEAM LINE ISOLATION VALVES The OPERABILITY of the main steam line isolation valves ensures that no more than one steam generator will blow down in the event of a steam line rupture. This restriction is required to: (1) minimize the positive reac-tivity effects of the Reactor Coolant System cooldown associated with the blowdown, and (2) limit the pressure rise within containment in the event the steam line rupture occurs within containment. The OPERABILITY of the main steam isolation valves within the closure times of the Surveillance Require-ments are consistent with the assumptions used in the safety analyses. Dj5EKT C -> 10 :: 0302 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION I The limitation on steani generator pressure and temperature ensures that the oressure-induced stresses in the steam generators do not exceed the maximum allowable fracture toughness stress limits. The limitations of 70*F and 200 psig are based on a steam generator RT NOT f 60*F and are sufficient to prevent brittle fracture. 3/4.7.3 COMPONEPT COOLING WATER SYSTEM The OPERABILITY of the Component Cooling Water System ensuras that suf-ficient cooling capacity is available for continued operation of safety-related equipment during normal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the safety analyses. 3/4.7.4 STATION SERVICE WATER SYSTEM The OPERABILITY of the Station Service Water System ensures that suffi-cient cooling capacity is available for continued operation of safety-related equipment during normal and accident conditions. The redundant cooling capa-city of t.iis system, assuming a single failure, is consistent with the assump-tions used in the safety analyses. 3/4.7.5 ULTIMATE _ HEAT SINK The limitations on the ultim2te heat sink level and temperature ensure that sufficient cooling capacity ie available to either: (1) provide normal cooldown of the facility or (2) mitigate the effects of accident conditions within acceptable limits. O COMANCHE PEAK - UNIT 1 B 3/4 7-3 -

IXX-88512 ATTACHMENT 10 PAGE 49 0F $4

   'N (G                      2n A.er C-3/4.7.1.6      MAIN FEEDWATER ISOLATION VALVES The feedwater isolation valves, the feedwater isolation bypass valves, and the feedwater preheater bypass valves are deLigned to close on a Feedwater Isolation Signal to

1) limit the cooldown following a safety injection / reactor trip, and 2) limit the mass addition to the containment on a steamline break inside containment, and 3) limit the severity of feedwater malfunctions which result in over feeding of a steam generator. The allowed outage times and required actions are consistent with normal plant operating requirements and the safety functions of the valves.

3/4.7.1.7 STEAM GENERATOR ATMOSPHERIC RELIEF VALVES The OPERABILITY of the steam generator atmospheric relief g('%,

     /  valves (ARVs) ensures that reactor decay heat can be dissipated to the atmosphere in the event of a steam generator tube rupture and loss'of offsite power and that the Reactor Coolant System can be cooled down for Residual Heat Removal System operation. Two ARVs are required to cool the Reactor Coolant System in a time frame compatible with prevention of overfill of the raulted steam generator.

All *our ARVs are required to be OPERABLE to allow for not being able to use the ARV on the faulted steam g?nerator and an active failure of one of the remaining three ARVs. O

TXX-88512 ATTACHMENT 10 PAGE 50 0F 54 PLANT SYSTEMS ( BASES ULTIMATE HEAT SINK (Continued) The limitations on minimum water level is based on providing a 30-day cooling water supply to safety-rMated equipment without exceeding its design basis temperature and is consistent with the recommend 3tions of Regulatory Guide 1.27, "Ultimate Heat Sink for Nuclear Plants," ,rdv. 2 (January 1976). The limitation on maximum temperature is based on the naximum allowable compo-nent temperatures in the Service Water and Component Cooling Water Systems, and the requirements for cooldown. The limitation an average sediment depth is based on the possible excessive sediment buildup in the service water intake channel. 3/4.7.6 FLOOD PROTECTION The limitation of flood protection ensures that facility protective { actions will be taken in the event of flood conditions. The only credible~

food condition that endangers safety related Equipment is from water entry l into the turbine building via the circulating water system from Squaw Creek Reservoir and then only if the level is above 778 feet Mean Sea Level. This corresponds to the elevation at which water could enter the electrical and h

V r:ontrol building endangering the safety chilled water system. The surveillance requirements are designed to implement level monitoring of Squaw Creek Reservoir should it reach an abnormally high level above 776 feet. The Limiting Condition for Operation is designed to implement flood protection, by ensuring no open flow path via the Circulating Water System exists, prior to reaching the postulated flood level. 3/4.7.7 CONTROL ROOM HVAC SYSTEM The OPERABILITY of the Control Roor,i HVAC System ensures that: (1) the control room ambient air temperature does not exceed the allowable temperature per 3/4 7.11 for continuous-duty rating for the equipment and instrumentation cooled by this system, and (2) the control room will remain habitable for opera-

tions personnel during and following all credible accident conditior- Opera-tion of the syrtem 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, requirements of General Design Criterion 19 of 10 CFR 50 Appendix A. ANSI N510-1W & will be sed as a procedura? quide for surveillance testing. 196 0 10 h 001 b] COMANCllE PEAK - UNIT 1 8 3/4 7-4 . L

TXX-88512 ATTACHMENT 10 PAGE 51 0F 54

   ,   PLANT SYSTEMS V)

BASES , 3/4.7.8 PRIMARY PLANT VENTILATION SYSTEM - ESF FILTRATION UNITS The OPERABILITY of the ESF Filtration Units ensures that radioactive materials leaking frols the ECCS equipment within the safeguards and auxiliary buildings following a LOCA are filtered prior to reaching the environment. 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' 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 operation of the ESF filtration units and the resultant effect on offsite dosage calculations was assumed in the safety analyses. ANSI H510-1975 will be t. sed as a procedural guide for surveillance testing. l M EC' ID 8: 0492 3/4.7.9 SNUBBERS I 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. Snubbers are classified and grouped by design and manufacturer'but not by size. M r example, mechanical snubbers utilizing the same design features of (( the 2-kip,10-kip and 100-kip capacity manufactured 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 fram either snanufacturer. ;_ A list of individual snubbers with detailed information of snubber loca-tion and size and of system affected shall be available at the plant in accor-dance with 10 CFR 50 50.71(c). The accessibility of each snubber shall be determined and approved by the Station Operation Review Committee (50RC). The determination shall be based upon the ex1 sting radiation levels and the expected time to perform a visual inspection in each snubber location as well as other factors associated with accessibility during plant operations (e.g., temperature, atmosphere, location, etc.), and the recommendations of Regulatory Guides 8.8 and 8.10. The addition or deletion of any hydraulic or mechanical snubber shall be made in accordance with 10 CFR 50.59. [ -The-+4sual-inspect-ion-frequency-is-based-upon-maintainteg-a-constant level of-snubbe r-p rotec ti on-to-e ac h - s a f e ty= re l a t ed- sy stem-dur-i ng-an-ea r thqua ke-o r--

         -seve re- t ransientr--Therefo re, -the-requi red- i ns pecti on- i nterva l- va ri e s-i nve rs e ly-with-the observad -snubber-f ailures on a given system- and-is determined-by-the--
         -number-of -inoperable-snubbers-found-during-an-inspect 4;n of each-system. In -

order-to - e s tabl i c h -the-i ns pect i on - f requency- f o r- eac h- type- o f -s nubber- on- a-safety-related-system,--it .;;s assumed-that-the-frequency of snubber fe4+ures-and initiating-events is constant-with-time and- that the f ailure-of-any-snubbee len- t ha t- s ys t em- c oul d-c aus e- the- s y s t em- t o- be- unp rot ec ted-a nd -to- result-in

          -                                                               Inspections performef) nefore-that--

b_ f*Hure-during-en-essumed-initiating-event. b M!w e h de md de CPCNBILITY ls b f k reeme nts Ju g rud- W .

U s @e nce ;c.s pegw m nu c4 tbe COMANCHE PEAK - UNIT 1 B 3/4 7-5

IXX-88512 !

      ,           AriACHMENT10 PAGE 52 0F 54 n            PLANT SYSTEMS g,

(V) h Uh

                                                                                                                              '5. 2 l I

l BASES i SNUBBERS (Continued) intervabhas-elapsed-may-be-used--as-4-new--reference point to deteH n the next-

                -inspection         However, the-result: Of sush-early in:pestion: perfor ed before
               -the-or-iginal-required-time-leterval h:: elapsed (n0minal tim: 10;; 25%) : y n:tr
               -beu s ed- t o-l engt hen- t he- requi-red- i ns pec t i on - in t e rvel .      Any inipectica i hcie-
               -casults- require-a- shorter -inspection interval wi'l override the previcu;
              --sc heduler 3
                        -The-acceptance-criteria -are-to- be- used-fft-the .isual inspection to deter-
            -mine-OPERAMITY-of-the-snubber . For example, if : fluid p0rt of a hydrault:
            --S nubbe r-i s- f o u nd -to- be- unc o v e redr-the-- snubber-s h a l l be deciared inopenable and
              -M1 not- be determined OPERABLE- via- functiona! testing.
                      --Ttr provide-assttrence-of-snubbeefunct40n:1 reli:bi'ity, one of three
              -f unc t to na 1 - t e s t i ng - met hod s-- i s- u s ed- w4 h-the- stated-ac c e ptanc e-cr i.te ni a :
                       -b---Functional?y-test-10% of a type-of-snubber with an additional .

405-tested-for ::ch functional testing f ailure, ne .s 2 . - Functionally-test a- sample size-and-determfee-samp4e-acceptanae-on-

                              -rejection using Figure d               ,, er
    \     (             -3 . Functionally-test a-representative sa ple si:e and deten'=4ne s==nle                      ~

merantanea ne Palartinn incinn the t r a t ati antiat ion . I l

                         "igure 4.7-1 w:s-developed u:ing "Y:ld': Sequential Probabi'ity Datie-
             - Man" as described in "Quality Centro! :nd Indu: trial Stati tic:" by                                      g-
             -AcnesonA Duncan Fermanent or other exemptions from tne surveillance program for individual snubbers may be granted by the Commission if a justifiable basis for exemption

! is presented and, if applicable, snubber life destructive testing was performed to qualify the snubbers for the applicable design conditions at either the com-pletion of their fabrication or at a svbsequent date. Snubbers so exempted l shall be listed in the 1lst of individual snubbers indicating the extent of the exemptions. The service life of a snubber is established via manufacturer input and information through consideration of the snubber service conditions and asaociated installation and maintenance ricords (newly installed snubbers, seal replaced, spring replaced, in high radiation area, in high temperature area, etc.). The requirement to monitor the snubber service life is included to-ensure that the snubbers periodically undergo a performance evaluation in view l of their age and operating conditions. These records will provide statistical bases for future consideration of snubber service life. O COMANCHE PEAK - UNIT 1 B 3/4 7-6 ,

t IXX-88512 ATTACICIENT10 PAGE530F54 . , PLANT SYSTEMS b BASES 3/4.7.10 SEALED SOURCE CONTAMINATION The limitations on removable contamination for sources requir:ng leak testing, including alpha emitters, is based on 10 CFR 70.39(a)(3) limits for plutonium. This limitation will ensure that leakage from Byproduct, Source, and Special Nuclear Material sources will not exceed allowable intake values. Sealed sources are classified into three groups according to their use, with Surveillance Requirements commensurate with the probability of damage to a source in that group. Those sources which are frequently handled are required to be tested more often than those which are not. Sealed sources which are continuously enclosed within a shielded mechanism (i.e., sealed sources within radiation monitoring or boron measuring devices) are considered to be stored and need not be tested unless they are removed from the shielded mechanism. 3/4.7.11 AREA TEMPERATURE MONITORING The limitations en nominal area temperatures ensure that safety-related equipment will not be subjected to temperatures'that would impact their envi'ron-mental qual;fication temperatures. Exposure to temperatures in excess of the maximum temperature for normal conditions for extended periods of time could reduce the qualified life or design life of that equipment. Exposure to tem-O peratures in excess of the maximum abnormal temperature could degrade the ( operability of that equipment. 3/4.7.16 UPS HVAC SYSTEM The OPERABILITY of the UPS HVAC System ensures that the uninterruptible power supply and distribution rooms ambient It1:0340 air temperatures do not exceed the allowable temperature per 3/4.7.13 for continuous-duty rating for the equipment and instrumentation cooled by this system. O COMANCHE PEAK - UNIT 1 B 3/4 7-7 .

TXX-08512 ATTACHMENT 10 PAGE $4 0F 54 PLANT SYSTEMS : n () BASES 3/4.7.15 SAFETY CHILLED WATER SYSTEM The OPERABILITY of the Safety Chilled Water System ensures that sufficient I cooling capacity is available for continued operation of equipment h4*g a;;id;nt ;;nditi;n;. The redundant cooling capacity of this system, assuming a single failure, in consistent with the assumptions used in the safety analyses. U.n k $ TheSafetyChilled[WaterSystem(SCWS),inconjunctionwithrespective l emergency fan coilY, is required in accordance with Specification Definition 1.20 (OPERABILITY) to provide heat removal in maintaining the various Inl0341 Engineered Safety Features (ESFs) room space design temperatures below th,e associated equipment qualification limits for the' range of Design Basis Accid 't conditions. Action Requirements are provided to ensure OPERABILITY of the vital bus inverters and emergency battery chargers, by verifying within O Q one hour that the normal HVAC system is providing spa.ce cooling to the vital power distribution rooms. The Action Requirement is provided to establish i within 8 hours OPERABILITY of the Emergency Core Cooling Systems (ECCS) which do not depend on the inoperable SCWS. The 8 hour period provides a reasonable time in which to establish OPERABILITY of this complement of key safety systems. This requirement ensures that c functional train of ECCS equipment is available to put the plant in a safe, stable condition for t M most probably abnormal operational occurrences. An Action Requirement )f 24 hours is provided to establish OPERABILITY of the remaining required safety systems which do not depend on the mcperable SCWS. A seven day Action Requirement is for a single SCWS out of service, based en the high reliability of offsite power. The term "verify" is used in this context to determine if certain components are out-of-service for maintenance or other reasons, it does not mean to perform the Surveillance Requirements needed to demonstrate OPERABILITY of the component. l l l A b 1 1

TU.-88512 AITACHMENT11 PAGE 1 0F 52 O COMANCHE PEAX STEAM ELECTRIC STATION TECHNICAL SPECIFICATI,0]! .. . 3/4.8 1 , f O

fxx 88512 ( ATTACHnENT11 CPSES Technical Specifications

  . PAGE 2 Of 52                 NRC Draft 2 Markup Section 3/4.8 Change 10#        Justification For Change

(] v 0347 Added Action (C) to address the combination of one offsite circuit and one diesel generator inoperable simultaneously. Revision 4 of the Standard Technical Specifications covered this situation; Revision 5 does not.- It would appear to be an inadvertent oversight. Lacking this action, this condition defaults to Specification 3.0.3 requiring action within one hour. This is not warranted in this situation because both trains are not inoperable (will still meet design basis barring a single failure). This is also not enough time to take corrective action to avoid a forced shutdown. This change is similar to that licensed at Vogtle, South Texas and most other plants licensed to Revision 4 and previous Standard Technical Specifications. This Action Statement was also issued as part of Generic Letter 84-15. 0352 Change this Action Statement that references "Action a or b" to reference "Action b or c". Due to new "Action c." Present Action Statement c is additional action to be taken in the event of Diesel Generator inoperability to preserve Train OPERABILITY and to verify availability of O V the steam driven AFW pump. Action a does not involve inoperability of the Diesel Generator but Action b and (proposed) Action c do. This appears to be another oversight in the change from Revision 4 to Revision 5 of the Standard. It is essentially the same as all other licensed plants. O

IIX 88512 ATTA N T H CPSES Technical Specifications M 3 5 52 NRC Draft 2 Markup

                . .                Section 3/4.8 m

lj Change 10# Justification For Change 0358 This change implements measures to reduce to a minimum 0360 the number of cold, fast starts performed on the 0361 Emergency Diesel Generators. Concerns over premature diesel engine degradation from cold, fast starts go back a number of years and are officially documented in Generic Letter 84-15. Subsequent to the issuance of this generic letter, plants began to receive Technical Specifications that relaxed the requirements for diesel starts from "ambient" conditions and the rapid loading that was required on a monthly or more frequent basis. This was incorporated in the Standard Technical Specifications as a note to 4.8.1.1.2a5) and 4.8.1.1.2a6); however, there have been several different wordings incorporated into plcnt specific Technical Specifications and placement in different sections of Specification 4.8.1.1.2. The Technical Specifications require starts from "ambient conditions" only in 4.8.1.1.2a (nominal monthly frequency)' but all other required diesel tests require start verification without addressing an "ambient" o- other required starting condition. Lack of tois detail leaves an ambiguous requirement with respect , what measures to

  ,                 reduce premature diesel degradation ma?      e used. To clarify the intent of Generic Letter 84-15, a slightly N-revised footnote has been attached to 4.8.1.1.2 rather than 4.8.1.1.2a4. It clearly states that prelube periods are allowed for all starts and that other warmup measures are permissible for all starts except for the 184 day requirement from "ambient conditions". Thus, measures to reduce wear and stress on the diesel engine are clearly allowed for 18 months testing in addition to the monthly tests and are consistent with the intent of Generic Letter 84-15. These precautions are also in accordance with NUREG 1216 on TDI diesel generators and the CPSES response (TXX-6236).

The 184 day test requirement has also been moved from the footnote to a stand alone surveillance requirement. This is consistent with Standard Technical Specification format and clearly segregates the two different requirements. This facilitates understanding and implementation of the Specification including tracking and scheduling of surveillance requirements. Finally, the note requiring the diesel to be operated with a load in accordance with the manufacturer's recommendations has been combined in the common note of 4.8.1.1.2. This makes the requirement apply to all diesel surveillance testing, not just the 18 month testing. These changes are similar to that Licensed at Vogtle, Millstone, Shearon Harris and South Texas.

    ,      trx-ess12 ArrAcer !!             CPSES Technical Specifications PAGE 4 0F 52.                 NRC Draft 2 Markup Section 3/4.8

(.)

 . f3 Change 10#        Justification For Change Change maximum time to load diesel generator during testing from 60 seconds to 80.5 seconds. Under accident or transient conditions, the CPSES emergency diesels are designed to load via the Solid State Safeguards Sequencer (SSSS). All automatic loading times are greater than or equal to 80.5 seconds. For the purposes of surveillance testing, the diesel generator should not> be subjected to any more adverse conditions than it will be subjected to during a design base condition. Although the loading under design base conditions is in "blocks" it on be approximated by uniform loading over the same period of time es the SSSS time. The actual loading by the sequencer is tested every 18 months in loss of offsite power and Safety Injection testing. Requiring the diesel to load more rapidly than the design condition subjects the diesel to abnormal stresses and wear and is contrary to Generic Letter 84-15. The plant specific time for this specification is 80.5 seconds.

0359 Delete loading time requirement since this is manually performed by the operators and does not demonstrate any increased teliability of the diesels and increases the o chance for operator error. Regulatory Guide 1.108 defines t V the intent of this surveillance to verify load handling capabilities of the diesel generators by loading it to the maximum extent practicable and maintaining that load for 1 hour. The Regulatory Guide does not mention any requirement for a specified time to reach that load. 0360 See ID# 0358 0361 See IDJ 0358 0368 Deleted the requirement for surveillance testing of safety features. CPSES is d3 signed such that all start signals are blocked by engaging the barring device or placing the diesel in the maintenance lock out mode. It is neither customary, nor desirable, to intentionally challenge industrial safety devices. This change is similar to that l licensed at Callaway, Vogtle, South Texas, Wolf Creek and Shearon Harris. I l Q l

IIX 88512 AliAC M NT 11 nrE 5 or 52 CPSES Technical Spscifications ' NRC Draft 2 Markup

                 . .                Section 3/4.8 g

() -Change IDA Justification For Change 0372 Change the test schedule to reflect industry practice. 0373 This change adds the note to table 4.8-1 that allows for the provision to reduce the failure count -to zero. Additionally, this change deletes the reliability action in table 4.8-2. This table is deleted since the TDI Diesels have had to undergo a very rigorous reliability program. Under this reliability / inspection program information was presented to the NRC from the Owners group and a subsequent Safety Evaluation Report was issued as NUREG 1216. These new reliability requirements which have been added to the desk top working copy of Standard Technical Specifications are from Generic Letter 84-15 but this letter specifically states that this program is out for comment and comparison. This is similar to what was performed at-Shearon Harris who participated in the TDI Owners Group. The reliability issue, in general is being addressed in conjunction with the resolution of the Blackout Issues. It is anticipated that the outcome of this resolution will require that all utilities have a reliability program but that it wiTT not be part of the Technical Specifications. This change is similar to that Licensed at South Texas, Vogtle, Shearon Harris and n Seabrook. 0374 The ACTION Statement is revised to refer to Specification 0388 3.4.9.3 for Reactor Coolant System pressure relief methods. The revised ACTION Statement will, for example, now allow utilization of residual heat removal (RHR) suction relief valves to provide overpressurization relief for the Reactor Coolant System. These valves are unaffected by a transient, since utilization of a Reactor Coolant System vent requires cooldown of the primary, going "solid", and opening a vent to the containment atmosphere. Also, this would allow the Reactor Coolant System to remain closed unless the reliefs were required to mitigate an overpressurization event. This revision is similar to the specification approved at Vogtle. O

in 88512

   '"'" I I N 6 & 52             CPSES Technical Specifications NRC Oraft 2 Markup Section 3/4.8 Change-10#      Justification For Change 0390       Delete reference to Table 3.8-1. This change is based 0392       on the relocation of Table 3.8-1 to the TS!P which is 0393       justified below.

0395 The' description of the function test is revised to reflect the previously agreed upon surveillance from CPSES Final Draft Technical Specifications. The specific values for testing acceptance criteria is more appropriately described in plant surveillance procedures where the specific manufacturer's recommendation for this acceptance criteria can be maintained and kept up to date. This change is similar to that Licensed at Wolf Creek, Callaway, Vogtle, San Onofre, Byron, Catawba and McGuire. 0396 This TS is being relocated to the CPSES Technical Specification Improvement Program since precedence for complete removal of the TS has not yet been set. TV Electric believes the inclusion of this TS is unnecessary and the requirements would,be more appropriately addressed; in the CPSES Technical Specification improvement Program, however, only the table is being proposed for relocation , at this time. Relocation of this specification is consistent with the guidance provided in the NRC's Interim Policy Statement (52FR3788), February 6,1987, and the recommendations of the Westinghouse Owners Group MERITS Program. Priority is given to the relocation of this table since the detailed nature of this table would require formal , license amendments for such things as a nomenclature change, a change in vendor supplier, a design improvement by a vendor, or plant modifications. Changes or modifications to equipment are already under the control of 10CFR50.59. The information currently in this table is more appropriately maintained in a document subject to TV Electric administrative control and 10CFR50.59 review under the CPSES Technical Specification improvement Program. The specific information supplied in Table 3.8-1 has not completed the validation process and is expected to change. As soon as this information is available it will be transmitted to the NRC under a separate letter. This change is similar to that Licensed at Seabrook, Vogtle, Shearon Harris and South Texas.

TXX-88512 ArrAC E NT 11 CPSES Technical Specifications PAGE 7 0F 52 NRC Draft 2 Markup Section 3/4.8 V Change 10f Justification For Change 0563 Added note to 24 hour DG Load Run to clarify that variations of voltage and frequency outside of the specified bands during the test caused by grid disturbances do not invalidate the test. Performance of the 24 hour load test of the Diesel Generator requires that the diesel be running in parallel with and supplying power to the grid, in order to achieve the required load. As a result, any fluctuation or disturbance on the grid will be seen on the Diesel Generator output as well. If the disturbance is severe enough in magnitude or duration, protective relaying will divorce the diesel from offsite power. Lesser transients will not cause protective action and from the standpoint of a difference of load on the diesel, it is insignificant. The intent of this surveillance is to test the endurance of the diesel including demonstration of-proper voltage and frequency. regulation under full load . conditions. Grid induced fluctuations are not indicative of a diesel or generator problem and ther7 ore should not cause rejection of test results. This test is a major plant evolution, usually on the critical path of an j outage. To prevent the possibility of a non-diesel generator related problem from invalidating or placing into question the acceptability of the test, this footnote is added to clarify the issue in advance. This change is similar to that licensed at Vogtle. 1 0576 Add the option to reject the CCW pump or a load of > 783 kw since in either case the intent of Regulatory Guide 1.108 is being met. The 783 kw is derived from the name plate data for the CCW pump which may or may not be the actual load of the CCW pump. But, the CCW pump is going to be the largest load on the bus at any given time and by using the CCW pump regardless of whether or not its actual power requirements are > 783 kw will ensure the Diesel Generator can handle thTs load rejection. The Regulatory l Guide requirement is that the largest load be rejected and the generator maintain voltage and not reach the overspeed t rip. Using either the CCW pump or a load of > 783 kw meets the guidance of Regulatory Guide 1.108. l O l

E 11X 88512 AliACH e t 11 CPSES Technical Specifications PAGE 8 of 52- NRC Draft 2 Markup Section 3/4.8 g Change 104 Justification For Change 0927 Changed Hot "Shutdown" to Hot "Standby" in Action Statements a.,b., and c. to be consistent with the rest of the Action Statements and to ensure that their is no confusion as to how long the operator has to get to Hot Standby between Action Statements. 0928 Changed frequency from 31 days to 92 days for check / removal of water in the fuel oil storage tanks. The applicable requirement from the Standard Technical Specifications for CPSES is 92 days because the groundwater level is below the level of the fuel oil storage tanks. This significantly reduces the probability of water incursion into the storage tanks and 92 day verification is satisfactory. Performance of this surveillance requires physical access to the fuel oil storage tank by removal of the missile shield covering the manways. This evolution requires significant effort and lowers the tank integrity while the shield is removed. This is not warranted due to the low probability of water entry to the tank. Note that normal fuel sampling from the tank is accomplished by pumping fuel oil with the fuel oil transfer pump and doesn't require physical tank access. In' addition, if water was to buildup sufficiently (7 in the fuel oil storage tank sludge pit to be pumped to L/ the day fuel tank, the water would be evident in the 31 day check required by Specification 4.8.1.1.2b or the fuel sample required by Specification 4.8.1.1.2e. 0929 Change 440 rpm to 441 r'pm (typo). Add (58.8 Hz) after the 441 rpm requirement to allow use of the frequency indication to perform surveillance. 120 (freq) = RPM 120 (f) = 441 (# of poles) 16 f = 58.8 Hz 0930 Changed wording in Action Statements a.1 and a.2. The change is made to allow removal of fuses to act as the redundancy for penetration isolation of an inoperable protective device. Additionally, the breaker wording has been removed since a breaker is really a subset of the protective devices and the word "backup" has been replaced with "associated" to prevent confusion as to what the table refers to as backup. O

      ,s
       , TXXP.512 Affecm p i !!                                   CPSES. Technical Specifications PAGE 9 0F 52                                            NRC. Draft 2 Markup Section 3/4.8' Change 10#                                  Justification For Change 0931                            This Specification is being deleted since all penetrations are being modified to meet the requirements of primary and backup overcurrent protection. This is described in more y                                                 detail in FSAR section 8.3.

O l 1 lO 4 i i

                                     ~     ^

J DWH2 : ATTACMENT 11 PAGE 10 0F $2

                    .      .                  I 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1      A.C. SOURCES OPERATING LIMITING CONDITION FOR OPERATION

3. 6.1.1 As a minimum, the following A.C. electrical power sources shall be OPERABLE:

a. Two physically independent circuits between the offsite transmission network and the onsite Class 1E Distribution System, and

b. Two separate and independent diesel generators, each with:

1) A separate day fuel tank containing a minimum volume of 1440 gallons of fuel,

2) A separate Fuel Storage System containing a minimum volume of 88,175 gallons of fuel, and

3) A separate fuel transfer pump.

APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: '- -

a. With one offsite circuit of the above-required A.C. electrical power sources inoperable, demonstrate the OPERA 8ILITY of the remaining A.C.

sources by performing Surveillance Requirement 4.8.1.1.1.a within O 1 hour and at least once per 8 hours thereafter. If either diesel generator has not been successfully tested within the past 24 hours, demonstrate its OPERABILITY by performing Surveillance Requirements i 4.8.1.1.2.a.4 and 4.8.1.1.2.a.5 for each such diesel generator. 108 0927 srww separately, within 24 hours.g Restore the offsiteleircuit to OPERABLE status within 72 hours er be in at least HOT 4WT40W within the next 6 44-hours and in COLD SHUTOOWN within the following 4+ hours. 30

b. With either diesel generator inoperable, demonstrate the OPERABILITY of the above required A.C. offsite sources by performing Surveillance Requirement 4.8.1.1.1.a within 1 hour and at least once per 8 hours thereafter. If the diesel generator became inoperable due to any cause other than preplanned 9teventive maintenance or testing, demon-strate the OPERABILITY of the remaining OPERABLE diesel generator by performing Surveillance Requirements 4.8.1.1.2.a.4 and 4.8.1.1.2.a.5 within 24 hours unless the diesel is already operating and loaded.#

Restore the inoperable diesel generator to OPERABLE status within t within the next 4e hours and 4 72 hours or be in atsrenew least HOT in COLD SHUTOOWN within the following 30 5""TC~f44 hours. Id M I A -* 108:0347

       *This test is required to be completed regardless of when the inoperable diesel generator is restored to OPERABILITY.

0uring performance of surveillance activities as a requirement for ACTION O statements, the air-roll test shall not be performed. COMANCHE PEAK - UNIT 1 3/4 8-1 -

TXX-88512 INSERT A-3/4.8'.1

c. With one offsite circuit and one diesel generator of the above required [

A.C. electrical power sources inoperable, demonstrate the OPERABILITY of eg the remaining A.C. offsite source by performing Surveillance Requirement ( 4.8.1.1.la. within 1 hour and at least once per 8 hours thereaf ter, and, if the diesel generator became inoperable due to any cause other than ' preplanned preventative maintenance or testing, demonstrate the OPERABILITY of the remaining OPERABLE diesel generator by performing Surveillance Requirements 4.8.1.1.2a.i) and 4.8.1.1.2a.5) within 8 hours *, unless the OPERABLE diesel generator is already operating #, Restore at least one of the inoperable sources to OPERABLE status within 12 hours or - be in at least HOT STANOBY within the next 6 hours and in COLD SHUTOOWN within the following 30 hours. Restore the other A.C. power source (offsite circuit or diesel generator) to OPERABLE status in accordance with the provisions of 3.8.1.1, Action Statement a. or b., as appropriate, with the time requirement of the Action Statement based on the time of initial loss of the remaining inoperable A.C. power source. A successful test of diesel generator OPERABILITY per Surveillance Requirements 4.8.1.1.2a.4) and 4.8.1.1.2a.5) performed under the Action Statement for an OPERABLE diesel generator or a restored to OPERABLE diesel generator statisfies the diesel generator test requirement of Actlcn Statement a. or b.

                                                                         . .            ~          .

o O l 3 l l I , J !O

4 IIX-88512 ATTACRMENT11 E EC RICA POWER SYSTEMS LIMITING CONDITION FOR OPERATION ACTION (Continued) n. c. dg With one diesel generator inoperable, in addition to ACTIONg! or)(. l above, verify that: ID1:0352

1. All required systems, subsystems, trains, components, and devices that depend on the remaining OPERABLE diesel generator as a source of emergency power are also OPERABLE, and

2. When in MODE 1, 2, or 3, the steam-driven auxiliary feedwater pump is OPERABLE.

If these conditions are not satisfied within 2 hours be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. dg. With two of the above required offsite A.C. circuits inoperable, demonstratetheOPERABILITYoftwodieselgeneratorsseparatelyby[ performing Surveillance Requirements 4.8.1.1.2a.4 and 4.8.1.1.2.a 5 within 8 hours unless the diesel gener& tors are already operatin'g#;

restore at least one of the inoperable offsite sources to OPERABLE status within 24 hours or be in at least HOT STANDBY within the next 6 hours. With only one offsite source restored, restore at least two O offsite circuits to OPERABLE status within 72 hours from time of initial loss or be in at least HOT STANOBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.

9 g. With two of the above required diesel generators inoperable, demonstrate the OPERABILITY of two offsite A.C. circuits by performing Survail-lance Requirement 4.8.1.1.la. within 1 hour and at least onca per 8 hours thereaf ter; restore at lea *+ one of the inoperable diesel generators to OPERABLE status w', /,n 2 hours or be in at least HOT STANOBY within the next 6 hours od in COLD SHUTDOWN within the following 30 hours. Restore at east two diesel generators to OPERABLE status within 72 hours from time of initial loss or be in d least HOT STANOBY witnin the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.8.1.1.1 Each of the above required independent circuits between the offsite transmission network and the Onsite Class 1E Distribution System shall be:

a. Determined OPERABLE at least once per 7 days by verifying correct breaker alignments, indicated power availability, and

b. Demonstrated OPERABLE at least once per 18 months during shutdown by transferring (manually and automatically) the 6.9 kV safeguards bus power supply from the preferred offsite source to the alternate offsite sourcs.

      #0uring performance of surveillance activities as a requirement for ACTION statements, the air-roll test shall not be performed.

COMANCHE PEAK - UNIT 1 3/4 8-2 . l

   -  - ' ' ' ~

ItX88512 . AliACHMEHf 11 PACE 13 0F 52 ELECTRICAL POWER SYSTEMS i SURVEILLANCE REQUIREMENTS (Continued) 4.8.1.1.2 Each diesel generator shall be demonstrated OPERA 8LE:

10 I: 0 58

a. In accordance with the frequency specified in Table 4,8-1 on a STAGGERED TEST BASIS by:

1) Verifying the fuel level in the day and engint-=ted fuel tank,

2) Verifying the fuel level in the fuel storage tank,

3) Verifying the fuel transfer pump starts and transfers fuel from the storage system to the day fuel tank, cad lfw C

4) Verifying the diesel starts '--- ""-*cceedition-and-acceler- l ates-te-at-least-441-rpe-in-less-than-ee-equal-te-10-seconds? gu The generator voltage and frequency th:P W(6900

69011bTts and 601 1.2 HzUithin 10 :: cent
ef ter the eteet-+ignal".

diesel generator shall be started for this test by using one ofThel IDI 0358 the following signals: a) Manual, or b) Start-up transformer secondary winding undervoli. age, or c) Simulated loss of preferred offsite power by itself, or M d) Simulated safeguards bus undervoltage, or e) Safety Injection Actuation test signal in conjunction with loss of preferred offsite power, or ( f) 3afety Injection Actuation test signal by itself.

5) Verifying the generalor is synchronized, loaded to between .

[6,800) and 7,000 kr in 1::: th:n er :qu:1-to-60-seco and operates at this load condition for at least 60 minute and 10 1: 0359

6) Verifying the diesel generator is aligned to provide standby power to the associated emergency busses.

1xssRT 6 3 l

                 *f          itsel generator starts from ambient conditions shall be performed only once per                  in these surveillance tests and all other engine starts for the purpose of this su                 ce testing shall include the acceleration to i                   rated speed in less than or equa                   econds and be preceded by an engine l                   pre-lube period and/or other warmup procedure                  s gradual loading           IDI 0358

(>80 sec) recommended by the manufacturer so that the me stress and wear on the diesel engine is minimized, i **0uring performance of surveillance activities as a requirement for ACTION statements, the air-roll test shall not be performed. l This band is meant as guidance to avoid routine overloading of diesel i generator. Loads in excess of the band or momentary variations due to chang-ing bus loads snall not invalidate the test. I COMANCHE PEAK - UNIT 1 3/4 8-3 l

                    ;,--g-=~     ' ~ ~         __ ,                 .-

AllACHMENT11 PAst 14 of 52 INSERT B - 4.8.1.1.2 All diesel generatcr starts for the purpose of surveillance testing as required by Specification 4.8.1.1.2 may be preceded by an engine prelube period. Further, all surveillance tests, with the exception of the 184 O day test, may also be preceded by warmup procedures (e.g gradual acceleration and/or gradual loading) as recommended by the manufacturer so that the mechanical stress and wear on the diesel engine is minimized. In additinn, for all starts, the diesel niust be operated with a load in accordance with the manufacturer's recommendations. l O O

i

TIX 88512 AllACHMENT !! PAGE 15 0F 52

ELECTRICAL POUER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) (d { b. At least once per 31 days and after each operation of the diesel where the period of operation was greater than or equal to I hour by checking for and removing accumulated water from the day fuel tank /; g 91 1000928

c. At least once per X days by checking for and removing accumulated water from the fuel oil storage tanks; l

d. By sampling new fuel oil in accordance with ASTM-04057-1981 prior to addition to storage tanks and:

1) By verifying in accordance with the tests specified in ASTM-0975-1981 prior to addition to the storage tanks that the sample has:

a) An API Gravity of within 0.3 degrees at 60*F, or a speci-fic gravity of within 0.0016 at 60/60 F, when compared to the supplier's certificate, or an absolute specific gra-vity at 60/60*F of greater than or equal to 0.83 but less than or equal to 0.89, or an API gravity of greater than or equal to [26] degrees but less than or equal to [_38] degrees; ' b) A kinematic viscosity at 40*F of greater than or equal to 1.9 centistokes, but less than or equal to 4.1 centistokes f( (

(alternatively, Saybolt viscosity, SUS at 100*F of greater than or equal to 32.6, but less than or equal to 40.1), if gravity was not determined by comparison with the sup-plier's certification; c) A flash point equal to or greater than 125'F; i d) A clear and bright appearance with proper color when tested in accordance with ASTM-04176-1982;

2) By verifying within 30 days of obtaining the ssmple that the other properties specified in Table 1 of ASTM-0975-1981 are met when tested in accordance with ASTM-0975-1981 except that the analysis for sulfur may be performed in accordance with ASTM-01552-1979 or ASTM-02622-1982.

e. At least once every 31 days by obtaining a sample of fuel oil in accordance with ASTM-D2276-1978, and verifying that total particu-late contamination is less than 10 mg/ liter when checked in accor-dante with ASTM-02276-1978, Method A; I

b bd bY 10 I: 0361 O( C0 4NCHE PEAK - UNIT 1 3/4 8-4 , l l

ixr 88512 l

                        -ATTACMENT11                                       .

PAE 16 0F 52 INSERT C - 4.8.1.1.2

f. At least once per 184 days by:

Verifying the diesel starts from ambient conditions and the O _ 1) generator voltage-and frequency are 6900 1 [690] volts and 60 f l.2 Hz within 10 seconds after the start signal. The diesel l generator shall be started for this test by using one of the signals listed in Surveillance Requirement 4.8.1.1.2a.4).

2) Verifying the generator is synchronized, loaded to between [6,800]

and 7,000 kw in less than or equal to 80.5 seconds and operates at this load condition for at least 60 minutes. ~

                           #     This band is meant as guidance to avoid routine overloading of diesel generator. Loads in excess of the band or momentary variations due to changing bus loads shall not invalidate the test.
                                                                             ..                       ~     .

O 9 ) f 4 4 iO

   - - - - - -   .w   e ve-

   ,.                                              _. .  . _ . . . . . . ._                             _.. i.

TU-88512 ATTACHMENT 11 PAGE 17 Of 52- - a sqh a ' s ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) cp,f' At least once per 18 months %, during shutdown, by: h 0576

1) Subjecting the diesel to an inspection in accordance with procedures prepared in conjunction with its manufacturer's recommendations for this class of standby service;

      % e. sikt. \a c&       2)     Verifying the generator capability to reject#,a load of grea'ter e**%g3 6                                                                                                                          l (CW pup) or                 [.thaiorequalto[783]kWwhilemai;1tainingvoltageat6900t 690, volts and frequency at 60 t[1.2, Hz;

3) Verifying the generator capability to reject a load of 7000 kW without tripping. The generator voltage shall not exceed 7590 volts during and following the load rejection;

4) Simulating a loss-of-offsite power by itself, and:

a) Verifying deenergization of the emergency busses and load, shedding from the emergency busses, and b) Verifying the diesel starts'on the auto-start signal,- ' energizes the emergency busses with pemanently connected loads within 10 seconds, energizes the auto-connected shutdown loads through the load sequencer and operates for O greater than or equal to 5 minutes while its generator is loaded with the shutdown loads. After energization, the steady-state voltage and frequenc_y of the emergency busses i I shall be maintained at 6900 t[690J volts and 60 + 1.2 Hz during this test.

5) Verifying that on a Safety Injection Actuation test signal, without loss-of-offsite power, the diesel generator starts on the auto-start signal and operates on standby for greater than or equal to 5 minutes. The generator voltage and frequency shall be 6900 1 690 volts and 60 i 1.2 Hz within 10 seconds
after the auto-start signal; the steady-state generak r voltage and frequency shall be maintained within these limits during this test;: 6) Simulating a loss-of-offsite power in conjunction with a Safety Injection Actuation test signal, and:

a) Verifying deenergization of the emergency busses and load shedding from the emergency busses; 4 i 4esel- mus t- be- ope ra t ed - w i t be-l oad - i n - In '0358 lO "f;r a ,y st

              .-< - - - * - *-                   <e'           - <          e '4   -                                                i COMANCHE PEAK - UNIT 1                             3/4 8-5         '

i j IXI88512 l AliACHMENT11 PAGE 18 0F 52 -

l fm ELECTRICAL POWER SYSTEMS

        \
   \

SURVEILLANCE REQUIREMENTS (Continued) b) Verifying the diesel starts on the auto-start signal, energizes the emergency busses with permanently connected loads within 10 seconde, energizes the auto-connected emergency (accident) loads through the load sequencer and operates for greater than or equal to 5 minutes while its generator is loaded with the emergency loads. After ener-gization, the steady-state voltage and frequency of the emergencybussesshallbemaintainedat6900t[690] volts and 60 1 1.2 Hz during this test; and c) Verifying that all automatic diesel generator trips, except engine overspeed and generator differential, are automatically bypassed upon loss of voltage on the emergency bus concurrent with a Safety Injection Actuation signal.

7) Verifying the diesel generator operates for at leisst 24 hours.

During the first 2 hours of this test, the diesel generator shall be loaded to an indicated 7600 - 7700 kW# and during the' remaining 22 hours of this test, the diesel generator shall be loaded to an indicated [680@ - 7000 kW . The generator voltage and frequency shall be 6900 t[690] volts and 6011.2 Hz within O 10 seconds after the start signaT; the steady-state generator voltage and frequency shall be maintained within these limits during this testf#Within 5 minutes after compleping this l 24-hour test, perform Specification 4.8.1.1.2/.f)b);* 10 1: 05 0 76 l

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 capability to: '

! a) Synchronize with the offsite power source while the generator is loaded with its emergency loads upon a simulated restoration of offsite power, b) Transfer its loads to the offsite pwer source, and c) Be restored to its standby status. This band is meant as guidance to avoid routine overloading of the diesel generator. Loads in excess of the band or momentary variations due to changing bus loads shall not invalidate the test.

                  "If Specification 4.8.1.1.2e.6)b) is not satisfactorily completed, it is not necessary to repeat the preceding 24-hour test.            Instead, the diesel generator
            /       may be operated between 6800 - 7000 kW for 1 hour or until operating tempera-ture has stabilized.
            \ F# R h lc , d n.n uit.9 s d 9 p . mf re PN M4 b de 9. 4J d 'ir b i s de r. . ,e r%d o c- M - w r- tut n a               mat.

COMANCHE PEAK - UNIT 1 3/4 8-6 ,

TIX-88512 l ATTACHMENT 11 l PAGE 19 0F 5t . ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

10) Verifying that with the diesel generator operating in a test mode, connected to its bus, a simulated Safety Injectinn 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 fucl 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-th; fellowing diesel-geneeeter-leekeut-features-peevernt- die sel-gene ra to r - s t a r t i ng s a-) Bereing-elevice-engagedr-*e - - -

101:0368 4)--Maintenance-loc kout-Moder-O h* gr At least once per 10 years or after any modifications which could h aff2ct diesel generator interdependence by starting both diesel generators simultaneously, during shutdown, and verifying that both diesel generators accelerate to at least %4Q rpm in less than or equal to 10 seconds; and 441 Its0929 ($s.t da) [)6 At least once per 10 years by:

1) Pumping out each fuel oil storage tank, removing the accumul-ated sediment and cleaning the tank using a sodium hypochlorite solution or equivalent, and

2) Performing a pressure test of those portions of the diesel fuel oil system designed to Section III, subsection ND of the ASME Code, when tested pursuant to Specification 4.0.5.

4.8.1.1.3 Reports - All diesel generator failures, valid or non-valid, shall be reported to the Commission in a Special Report pursuant to Specification 6.9.2 within 30 days. Reports of diesel generator failures shall include the information recommended in Regulatory Position C.3.b of Regulatory Guide 1.108, Revision 1, August 1977. If the number of failures in the last 100 valid tests on a per nuclear unit basis is greater than or equal to 7, the report shall be supplemented to include the additional information recommended in Regulatory Position C.3.b of Regulatory Guide 1.108, Revision 1, August 1977. O COMANCHE PEAK - UNIT 1 3/4 8-7 .

IXX 88512 ATTACHMENT!) e TABLE 4.8-1 (\) v O!ESEL GENERATOR TEST SCHEDULE NUMBER OF FAILURES IN NUMBER OF FAILURES IN LAST 20 VALID TESTS

LAST 100 VALID TESTS
TEST FREQUENCY I1 54 Once per 31 days 1 2*" 15 Once per 7 days TABLE 4.8-2 ID I: 0373 ADDITIONAL RELIABILITY ACTIONS 0F FAILURES IN NO. OF FAILURES IN LA RO VALID TESTS LAST 100 VALID TESTS ACTION 6 ' Within 14 days prepare and maintain a report for NRC audit describing the diesel g'

i generator reliability improvement program l d implemented at the site. Minimum requirements for the report are indicated in Attachment.1 to this table, j 5 11 eclare the diesel generator in rable. Perform a re-quali qtion test pr'ogram for the ai'f(cted diesel generator. R alification test program requ ments j . are indicated in Atta .ent 1 2 to this table. l JAlsidTS ID 1: 0372 l

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.

          **The associated test frequency shall be maintained until seven consecutive l

) p( ( failure free demands have been performed af the number of failures in the last 20 valid demands has been reduced to one. 1 l l COMANCHE PEAK - UNIT 1 3/4 8 8 ,

lxx 88512 AirAtmENT 11 PHI 21 0F 52 INSERT D - Table 4.8-1 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 O. manufacturer and if acceptable reliability has been demonstrated. The reliability criterion shall be the successful completion of 14 consecutive tests in a single series. Ten of these tests shall be in accordance with the routine Surveillance Requirements 4.8.1.1.2.2a.4 and 4.8.1.1.2a.5 and four tests in accordance with the 184-day testing requirement of Surveillance Requirement 4.8.1.1.2f. If this criterion is not satisfied during the first series of tests, any alternate criterion to be used to transvalue the f ailure count to zero requires NRC approval. { i O 4 I I I l l i P O l

IXX 8%I2 AliACHMENT 11 PAGE 22 0F 52 ATfACHMENT 1 TO TABLE 4.8-2 REPORTING REQUIREMENT ( \ IM 0373 As minimum the Reliability Improvement Program report for NRC audit shall incl : a) as ary of all tests (valid and invalid) that occurred within the time perio over which the last 20/100 valid tests were performed b) analysis f failures and determination of root causes of failures c) evaluation each of the recommendations of NUREG/CR-0660, "Enhancement of Onsite Eme ency Diesel Generator Reliability in Operat.ing Reactors," with respect to their application to the plant d) identification of 1 actions taken or to be taken to 1) correct the root causes of failures fined in b) above and (2) achieve a general improvement of diesel enerator reliability e the schedule for implemen tion of each action from d) above

         ,)    an assessaient of the existin reliability of electric power to engineered-safety-feature equi ent Once 4. licensee has prepared and maint n an initial report detailing the diesel generator reliability improvement rogram at his site, as defined OC     above, the licensee need prepare only a su lemental report within 14 days after each failure during a valid demand fo so long as the affected diesel genernor unit continues to violate the crite a (3/20 or 6/100) for the reliability improvement r gram remedial action. The supplemental report need only update the failure /cemand history for the a ected diesel generator unit since the last rept rt for thet diesel generator.                e supplemental report l

shall also present an analysis of the failure (s) wit a root cause determination, if p.issible, and shall delineate any fu ther procedural,. ', hardware or operat ~ .al changes to be incorporated into he site diesei generator improvem~ ; program and the schedule for imple tation of those changes. l In addition to the above, submit a yearly data report on the dt sel generator reliability. N l \ x l \

l COMANCHE PEAK - UNIT 1 3/4 8-9 . l

< IXX88512 AllACHMENT!! PAGE 23 0F 52 IDI 0373 ATTACHMENT 2 TO TABLF 4.8-2 OIESEL ENERATOR REQUALIFICATION ROGRAM 1. rform seven consecutive successful demands without a failure within 30 da s of diesel generator being restored to operable status and fourteen con cutive successful demands without a failure within 75 days of diesel gener tor of being restored to operable status.

2. If a fai ure occurs during the first seven tests in the requalification test prog m, perform seven successful demands without an additional failure wit in 30 days of diesel generator of being restored to operable status and f eteen consecutive successful demands without a failure within 75 days f being restored to operable states.

3. If a failure occu during the second seven tests (tests 8 through 14) of (1) above, perform ourteen consecutive successful demands without an additional failure w thin 75 days of the failure which occurred during the requalification t ting.

4. Following tue second fat re during the requalification test program, be in at least HUT ST.'.NOBY wi in the next 6 hours and COLD SHUTDOWN within

           ,                  the following 30 hours.

5. During requalification testing he diesel generator should not be tested more frequently than at 24-hour tervals.

l p {~ After a diesel generator has been succes ully requalifisd, subsequent i \ repeated requalification tests will not be required for that diesel generator i under the following conditions: (a) The number of failures in the last 20 val demands is less than 5. (b) The number of failures in the last 100 valid emands is less than 11. (c) In the event that following successful requalif1 ation of a diesel generator, the number of failures is still in exc s of the remedial I action criteria (a and/or b above) the following e eptien will be allowed until the diesel generator is no longer in y lation of the remedial action criteria (a and/or b above). l Requalification testing will not be required provided that aft r each valid demand the number of failures in the last 20 ana/or 100 valid d ands has not increased. Once the diesel generator is no longer in vio ation of the remedial action criteria above the provisions of those criteric al e will prevail.

O' COMANCHE PEAX - UNIT 1 3/4 8-10 ,

i i l TU 88512 l ATTACHMENT 11 PAGE240FSf O ELECTRICAL POWER SYSTEMS b A.C. SOURCES SHUTDOWN LIMITING CONDITION FOR OPERATION

3. 8.1. 2 As a minimum, the following A.C. electrical power sources shall be OPERABLE:

a. One circuit between the offsite tran:, mission network and the Onsite Class 1E Distribution System, and

b. One diesel generator with.

1) Day fuel tank containing a minimum volume of 1440 gallons of fuel,

2) A fuel storage system containing a minimum volume of [88,175]

gallons of fuel and

3) A fuel transfer pump.

        , APPLICABILITY:         MODES S and 6.

k/ ACTION: With less than the above minimum required A.C. electrical power sources OPERABLE, immediately suspend all operations involving CORE ALTERATIONS, i positive reactivity changes, movement of irradiated fuel, or crane operation ID 1: 0374l 6,ith loads over the fuel storage pool, and within 8 hours,4epressuei-ze-and- h vent-the-Reactoe-Coolant Syst= through-a-greater-than-on-equal-to-2.9e. souare s beh-ve st. In addition, when in MODE 5 with the reactor coolant loops.not filled, or in MODE 6 with the water level less than 23 feet above the reactor vessel flange, immediately initiate corrective action to restore the required sources to OPERABLE status as soon as possible. pmv do. ce.\;e.5 u.p & \;by Sor hL E"-' be '~ C.^U ' system la acw.-J-c e. ~; n 9 eu ; f.w bt 3.01.3. SURVEILLANCE REQUIREMENTS 4.8.1.2 The above required A.C. electrical power sources shall be demonstrated OPERABLE by the performance of each of the requirements of Specifications

4. 8.1.1.1, 4. 8.1.1. 2 (except for Speci fication 4. 8.1.1. 2a. 5)), and 4. 8.1.1. 3.

l t O i l COMANCHE PEAK - UNIT 1 3/4 8-11

l

TrX-88512 AliACHMENT 11 PAGE 25 0F 52 l

3/4.8.2 D.C. SOURCES OPERATING LIMITING CONDITION FOR OPERATION 3.8.2.1 As a minimum, the following 0.C. electrical sources shall be OPERABLE:

a. Train A - 125 volt D.C. Station Batteries BTIE01 and BTIED3 cad at least one full capacity charger associated with each battery;and

b. Train 8 - 125 vol; D.C. Station Batteries BT1ED2 and BTIED4 and at least one full capacity charger associated with aach battery.

APPLICABILITY: _ MODES 1, 2, 3, and 4. ACTION: With one of the required battery trains and/or required full-capacity chargers inoperable, restore the inoperable battery train and/or required full-capacity charger to OPERABLE status within 2 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTD0hh within the-following 30 hours. '- - SURVEILLANCE REQUIREMENTS 4.8.2.1 Each 125 V D.C. station battery and charger shall be demonstrated OPERABLE:

a. At least once per 7 days by verifying that:

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

2) The total battery terminal voltage is greater than or equal to 128 volts on float charge.

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

         'l  in-88512 l  AllACHMEH1 11 PAGE 26 0F 52      -

I B r 0. C. SOURCES ( ( 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 Categor 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 8 chm, 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) Tne 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

( less than or equal to 150 x 10 8 ohm, and

4) The battery charger will supply at least 300 amperes at 125 volts for at least 12 hours.; d. At least once per 18 months, during shutdown, by verifying that the battery capacity is adequate to supply and maintain in OPERABLE l status all of the actual or simulated emergency loads for the'derign duty cycle when the battery is subjected to a battery service test; i; e. At least once per 60 months, during shutdown, by verifying that the l battery capacity is at least 80% of the manufacturer's rating when I 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 l 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 manufacturer's rating.

l Oc l COMANCHE PEAK - UNIT 1 3/4 8-13 , i

AliACH"rHi11

        ' AGE 21 vs 52   .

TABLE 4.8-3 w itril 3 BATTERY SURVEILLANCE REQUIREMENTS k CATEGORY A(1) CATEGORY B(2) CARAMETER LIMITS FOR EACH LIMITS FOR EACH ALLOWABLE (3) DESIGNATED PILOT CONNECTED CELL VALVE FOR EACH , CELL CONNECTED CELL Electrolyte > Minimum level > Minimum level Above top of Level indication mar indication mark, plates, . and < 4" above and < " above and not ! maximum level maximum level overtsowing indication mark indication mark Float Voltage > 2.13 volts ), 2.13 volts (6) > 2.07 volts Not more than 0.020 below the average of all Specific > 1.200(5) > 1.195 connected cells

                                                                  '   '                                 ~       ~

Gravity (4) Average of all Average of all connected cells connected cells

                                                              > 1.205
                                                                                                -> 1.195(5)

C TABLE NOTATIONS (1) For any Category A parameter (s) outside the limit (s) shown, the battery may be considered OPERABLE provided that within 24 hours all the Category 8 measurements are taken and found to be within their allowable values, and provided all Category A and B parameter (s) are restored to within limits within the next 6 days. (2) For any Category 8 parameter (s) outside the limit (s) shown, the battery may be considered OPERABLE provided that the Category 8 parameter; are within their allowable values and provided the Category B parameter (s) are

 =            restored to within limits within 7 days.

(3) Any Category 8 parameter not within its allowable value indicates an inoperable battery. (4) Corrected for electrolyte temperature (reference temperature of 77*F) and level. (5) Or battery charging current is less than 2 arps when on charge. (6) Corrected for average electrolyte temperature. O< COMANCHE PEAK - UNIT 1 3/4 8-14 ,

TXX-88512 ATTACHMENT 11 PAGE 28 0F 52 . D.C. SOURCES 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: MODES 5 and 6. ACTION: Withtherequiredbatterytrain!and/orrequiredfull-capacitychargers inoperable, immediately suspend all operations involving CORE ALTERATIONS, positive reactivity changes, or movement of irradiated fuel; initiate correc-tive action to restore the required battery train and "ull-capacity chargersto OPERABLE status as soon as possible, and within 8 hours, de;;rc::urize-and-vent-the- Reac to r - C e c l a nt-System-t h rough-a-2r98-squa re-inc h-ven t' ID 3: 0374 pruids rthe V capcA; \ hy (ce hk ilac.

w C os ta.s+ Sys k in ev.cccaa sce m ik Spet s St ur F.w 3 ,.

SURVEILLANCE REQUIREMENTS 4.8.2.2 The above required 125V D.C. station batteries and full-cenacity l charge',' shall be demonstrated OPERABLE in accordance with Spacifit nfon l 4.8.2.1. t l l t l O COMANCHE PEAK - UNIT 1 3/4 8-15 , I

I

IXX 88512 l AITACHMENT 11

                              ~                                                                  l PAGE 29 0F 57                                                                         l

( 3/4.8.3 ONSITE POWER DISTRIBUTION OPERATING LIMITING CONDITION FOR OPERATION 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 1EA1,

2) 480-Volt Emergency Bus 1EB1 from transformer T1EB1, 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 TIEB2, and

3) 480-Volt Emergency Bus 1EB4 from transformer T1EB4.

c. 118-Volt A.C. Instrument Bus IPC1, IPC3, and 1EC1 energized from its associated inverter connected to 0.C., Bus 1E01*; ._ ,

d. 118-Volt A.C. Instrument Bus IPC2, IPC4, and 1EC2 energized from its associated inv)rter connected to 0.C. Bus 1E02*;

I O e. 118-Volt A.C. Instrment Bus 1ECS energized from its associated inverter connected to 0.C. Bus 1E03*;

f. 118-Volt A.C. Instrument Bus IEC6 energized from its associated inverter connected to 0.C. Bus 1E04*;

g. Train A 125-Volt 0.C. Busses 1E01 and 1E03 energized from Station Batteries BTIE01 and BTIE03, respectively; and l h. Train B 125-Volt 0.C. Busses 1E02 and IE04 energized from Station Batteries BTIE02 and BT1E04, respectively.

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

a. With one of the required trains of A.C. emergency busses not fully energized, reenergize the trains within 8 hc,urs or be in at least HOT STANDBY within the next 6 hours and in COLD SHUT 00WN within the folicsing 30 hours.

1 l l *The inverters may be disconnected from their 0.C. bus for up to 24 hours 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 i energized from their associated inverters and connected to their associated O.C. bus. COMANCHE PEAK - UNIT 1 3/4 8-16 - 1 l

f IXX-88512 AliACMENT11 PAGE 30 0F 52 ONSITE POWER DISTRIBUTION g.RA1,,- 7- . V 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 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.

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 O.C. source (s), or operating with the inverter not supplying the A.C. instrument bus (but with the instrument bus energized from its associated bypass distribution source), energize the A.C. instrument bus (ses) from its associated 0.C. bus within 24 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUT 00VN within the following 30 hours.

d. With one O.C. bus not energized from its associated station battery, reenergize the D.C. bus from its associated station battery within i 2 hours or be in at least HOT STAND 8Y within the next 6 hours and in COLD SHUT 00WN within the following 30 r.aurs.

l SURVEILLANCE REQUIREMENTS l l 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 l indicated voltage on the busses. l l l l l l l G \ COMANCHE PEAK - UNIT 1 3/4 8-17 ,

Txx-885ft AITACHMENT 11 PAGE 31 0F 52 , ONSITE POWER DISTRIBUTION SHUT 00WN

        'c     LIMITING CONDITION FOR OPERATION 3.8.3.2 As a minimum, the following electrical busses shall be energized in the specified manner:

a. One train of A.C. emergency busses consisting of one 6900-volt and two 480 volt A.C. emergency bus;

b. Two 118-volt A.C. instrument busses (channel-oriented) energized from their associated inverters connected to their respective D.C.

busses;

c. One train of A.C. instrument busses consisting of two 118-volt A.C.

instrument busses energized from their associated inverters connected to their respective D.C. busses. Busses shall be of the same train as Specifications 3.8.3.2a. and d.; and

d. One train of D.C. busses consisting of.two 125-volt D.C. busses.

energized from their associated battery banks. Busses shall be of the same train as Specifications 3.8.3.2a. and c. y APPLICABILITY MODES 5 and 6. (~ ACTION: With any of the above required electrical busses not energized in the required manner, immediately suspend all operations involving CORE ALTERATIONS, positive i reactivity changes, or movement of irradiated fuel, initiate corrective action to energize the required electrical busses in the specified manner as soon as possible, and within 8 hours, d:;;r:::urh: :nd v:nt the RCS thr:;;h st-least-a 2.08 squar; inch vent. y w Je nl ~ e i caphW4 fer Mc deader Goluf , sysh 2,s ce c co .-da.. - e - 4k S,m k. hm 7. 4, 'I. 3 . 10 h 0388 SURVEILLANCE REQUIREMENTS i 4.8.3.2 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. l O< COMANCHE PEAK - UNIT 1 3/4 8-18 ,

l

.)

TXX 88512' ATTACHMENT 11 ' fAGE 32 0F $2_ .

  ~g        h/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES                                                i A.      IRCUITS INSIDE PRIMARY CONTAINMENT                            7     , ,  , ,08 0931
                        \

LIMITING CDMOITION FOR OPERATION J . f, ! I 3.8.4.1 The f ing A.C. feeding the Fuel Transfer System circuits inside primary containee all be deenergized:

a. Circuitnumbeh4BLinpanelMCCIEB2-3. ,

1

b. Circuit number 50 panel MCC IEB1-2.

APPLICABILITY: MODES 1, 2, and 3. ACTION: With any of the above required circuits ene ized, trip the associated circuit breaker (s) in the specified panel (s) within 1 hqur. SURVEILLANCE REQUIREMENTS 4.8.4.1 Each of the above required A.C. circuits shall be det quined to be deenergized at least once per 31 days by verifying that the associated ()N ( circuit breakers are locked in the open position. Nx 0 COMANCHE PEAX - UNIT 1 3/4 8-19

IXX-08512 AIIACHnENT11 PAGE 33 0F $2 -

ELECTRICAL EQUIPMENT PROTECTIVE DEVICES v CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES LIMITING CONDITION FOR OPERATION 3.8.4.2 All containment penetration conductor overcurrent protective devices gi;:n ir T:b!: 3.8-1-shall be OPERABLE. APPLICABILITY: MODES 1, 2,' 3, and 4. D h OMO ACTION: With one or more of the containment penetration conductor overcurrent protective device (s) gi;;r ir Table 3.S-1 inoperable: (0 8: 0H2

a. Restors the protective device er feeder breakee to OPERABLE status or:

1. Deanergize the circuit (s) by racking out, locking open, or removing the Inoperable c4eeutt-beeakeFoe protective device and 7

p~ ..trippingjthe-en ist:d backup circuit breaker within 72 hour's7 N9 i declare the affected system or component inoperable, and verify scwed PNd" the inoperable circuit bre:koe-se protective device racked out, locked open, or removed at least once per 31 days thereafter; the um

                              '46 provisions of Specification 3.0.4 are not applicable to over-
                                       'cTJrrent* devices ++ der--beeakee+ in circuits which have their backup circuit breakees- trippe4&nd their inoperable cireilit prAWu e

[ br::ker: racked out, locked open, or remo'ved' or Jmd Deenergize the circuit (s) by trippin/renwny; d6 a I { *t**""\j/ 2. gathe :::0:iit:d b: kup. circuit br::k:r or racking out, locking open, or removing the inoperable.p4rcuTI bre:kee within 72 hours, declare the affected IDI 0930 js du " \7 ~ system or component in e

                        - j ' Nbeeaker to be trippedg               o 'yage, and verify the backup circu48 the inope i
       \

pic h" kM ) i - the provisions of Specification 3.0.4 out, locked open, or removed'at leasfonce per 7 days thereafter; are not applicable to

         \                                overcurrent devices in circuits which have their te:Eu? Cir: Lit
                                 --w breaker; tripped or their inoperable circuit bre:koes racked out, j                                           locked open, or removed; or              P" t'd A d *v * *-

l b. Be in at least HOT STANDBY within the next 6 hours and in COLD SHUT 00WN within thd following 30 hours. SURVEILLANCE REQUIREMENTS 4.8.4.2 N containment penetration conductor overcurrent protective devices gi r '- T:bl: 3.S-1 shall be demonstrated OPERABLE: D h OM3

a. At least once per 18 months:

l

1) By verifying that the medium voltage 6.9 kV and low voltage 480V l switchgear circuit breakers are OPERAELE by selecting, on a rotating basis, at least one or 10% of the circuit breakers whichever is greater of each current rating and performing the following:

a) A CHANNEL CALIBRATION of the associated protective relays, b) An integrated system functional test which includes simulated

 /'                                              automatic actuation of the system and verifying that each l \                                               relay and associated circuit breakers and control circuits function as designed, and l               COMANCHE PEAK - UNIT 1                             3/4 8-20       ,

IXX-88512 - ATTACCENT 11 PAGE 34 0F 52 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES w il" O( SURVEILLANCE REQUIREMENTS (Continued) c) For each circuit breaker found inoptrable 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 functior. ally 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 selected for functional testing shall be selected on a rotating basis. Testing of these circuit breakers shall consist of '-

injecting a current M th valu: ; quel to 300% of the pickup--of-4.5 l o ng- time-dehy-t eip-element-and-1500e f-the- pickun-o f-the short-ti:: delay-trip-element,- and v:ei-fying-that-the-circuit- 10 : 0395 break e r- ope ra tes- wi t h in- t he- t ime-delay-band-w i d t h- f o r- t ha t-c u r ren t-spec i f-f e d- by-t he- ma n u f ac tu re r . The-instantaneous-

                              +1 eme nt- s ha l4-be-te s t e d- by-i nj ec ting-+-c ue rent-eq u a b to-t 20%- o f '
                          -I the-pi c k up-v a l ue- o f-t he-e l eme n t' a n~d-ve ef-fy i ng- t ha t - t he- c i rc u i t-b rea ke e-t eips-ins ta nta ne ous ly-with-no-inten t iona b t i me- de l ay ,

Mc!ded case-circuit-breaker-testing-shal4-also-follow-thi5-- Spee edure-e xcept-tha t- ge ne rally-no-more- than-two-tr-i p- e l eme n t s , i ti:: del y :nd instentaneou:, will be i vcived. i Circuit l L (~ t.reakers found inoperable during functional testing shall be restored to OPERABLE status prior to resuming operation. For each circuit breaker found inoperable during tridse functional

            ,                 tests, an additional representative sample of at least 10% of I                   all the circuit breakers of the inoperable type shall also be j                     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.
              \

p _ , a . a o g a .., _ ., 2k m c.1se e z;,%. ne ,,,m u r.e p ,, n,, , a u C.c.wp M ha c hg, ,% yg ,\ b bg EhA A b 15 e 'as ILt g e ,- e l g

                                                                                                  )
                            ' by bkl        r',w w im b -et, Oc COMANCHE PEAK - UNIT 1                                3/4 8-21                ,

IXX 88512 ATTACHMENT 11 PAGE 35 Of 52

                     ~       '

TABLE 3.8-1 Mignl I O CONTAINMENT PENETRATION CONDUCTOR V( ' OVERCURRENT PROTECTIVE DEVICES

                                                                                      }     D 1: 0396 OEVICE NUMBER                                                     SYSTEM AND C0 CATION                                                     POWERED
                 \

1. 6. KVAC from Switchgears

a. witchgear Bus 1Al RCP #11

1) Primary Breaker IPCPX1 g) Relay 50M1-51

                             &     Relay 26 c)    Relay 86M

2) Back Breakers IAl-1 or 1Al-2 a) Re ay 51M2 b) Re1 51 for 1Al-1 c) Rela 51 for 1Al-2 d) Relay 6/1A1 ..

b. Switchgear Bus 1A2 RCP #12

{ 1) Primary Breaker PCPX2 t a) Relay 50H1-51 l b) Relay 26 l c) Relay 86M

2) Backup Breakers 1A2-1 o 1A2-2

ai Relay 51M2 b) Relay 51 for 1A2-1 c) Relay 51 for 1A2-2 d) Relay 86/1A2

c. Switchgear Bus 1A3 RCP #13

1) Primary Breaker IPCPX3 a) Relay 50H1-51 b) Relay 26 c) Relay 86M l 2) Backup Breaker 1A3-1 or 1A3-2 a) Relay 51M2 b) Relay 51 for 1A3-1 c) Relay 51 for 1A3-2

( d) Relay 86/1A3 COMANCHE PEAK - UNIT 1 3/4 8-22 .

                                      -           - - -             ,-            -                   /

IXX-88512 FA 6 . . . f 0396 TABLE 3.8-1 (Continued) V CONTAINMENT PENETRATION CONDUCTOR

                                        ~0VERCURRENT PROTECTIVE DEVICES 0 ICE NUMBER                                                                    SYSTEM AND OCATION                                                                     POWERED

1. 6. KVAC from Switchgears (Continued)

a. Switchgear Bus 1A4 RCP #14

1) Primary Breaker IPCPX4 Relay 50M1-51 b Relay 26 c) Relay 86M

2) Backup Breakers 1A4-1 or 1A4-2 a) Re1 51M2 b) Relay 1 for 1A4-1 c) Relay , for 1A4-2 - ~ -

d) Relay 86 4

2. 480 VAC from Switchgears 2.1 Device Location - Containment Recirc.

( 480V Switchgears 1EB1, 1E 1EB3 and IEB4 Fans and CRDH Vent Fans

a. Pr* mary Breakers - IFNA'I IFNAV2, IFNAV3, IFNAV4, l'

1 1FNCB1 and IFNCB2

b. Backup Brr ikers - IEB1-1, 1EB2-1, IEB3-1 and 1EB4-1

1) long Time & Instantaneous Relays

50/51 50/51 I TAV1 (IEB1-1) 77gXV2 (IEB2-1) 3 (IEB3-1) h4(1EB4-1) 50/51 50/51 l g i (1EB3-1) M2 1EB4-1) i
Associated circuit breaker shown in parentheses; e.g. ,1EB3-1, is bac up to

     . 1FNAV3 and IFNCBl.
                                                                                                              '\

COMANCME PEAX - lJNIT 1 3/4 8-23

IXX-88512 7 ' () IM 0396 ATTACHMENT 11 i.!.Uiq"lt. Jn PAGE 37 0F 52

                     ,                                                TABLE 3.8-1 (Continued) b

( CONTAINMENT PENETRATION CONOUCTOR OVERCURRENT PROTECTIVE DEVICES DEV E NUMBER SYSTEM AND L CATION POWERED

2. 480 VAC from Switchgears (Continued)

Time Delay Relays 6 1 1 v1 (1EB1-1) 1%;12 (1EB2-1) 1 NA (1EB3-1) )fg{f4(IEB4-1) 1 1 81 ( B3-1) 1 C82 (1EB4-1) 2.2 Device Location - 480V Containment Polar Switchgear IE84 - Crane

                                                                                                                          ~  -

a. Primary Breaker 15CCP1
b. Backup Breaker - 1E -1: 1) 51 ISCCP1; 2) 62 13CCP1; 3. 480 VAC from Motor Control Centers .

! 3.1 Device Location - MCC 1EB1-2 Containment bers listed below. l Primary and Backup - Bot primary and backup l Breakers breargrs have identical trip ! rating and are in the same i MCCCom(p(. These breakers are General Electric type THED or. THPK with thermal-magnetic tri'p elements. l MCC 1EB1-2 G.E. l COMPT. NO. BKR. TYPE SYSTEM POWERED 4G THED Motor Operated Valve 1-TV 4691 l 4M THED Motor Operated Valve 1-TV-4'693 1 COMANCHE PEAK - UNIT 1 3/4 8-24 , l

IXX-88512 ATTACH;;ENT 11 PAGE 38 0F 52

TABLE 3.8-1 (Continued) . CONTAINMENT PENETRATION CONOUCTOR ( OVERCURRENT PROTECTIVE DEVICM DEVI NUMBER AND LO TION R: f ME '" =

3. 480 V from Motor Control Centers (Continued)

MCC 1EB1-2 G.E. COMPT. NO. BKR. TYPE SYSTEH POWERED 3F THED Containment Drain Tank Pump-03

   .             9H                    THED                Reactor Cavity Sump Pump-01 9M                    THED                Reactor Cavity Sump Pump-02 7H                    T E0                Containment Sump #1 Pump-01 7M                    TH                  Containment Sump #1 Pump-02 6H                    THE0                RIP #11 Motor Space Heater-01 6M                    THED                RCP #13 Motor Space Heater-03 8B                    THED                Incore Detector Drive "/,"

80 THED Incore Detector Drive "B" 7B THE0 Incore Detector Drive "F" 50 THE0 el Transfer System Reactor Side Co t. PNL FDR-01 3B THED Stud ensioner Hoist Outlet-01 70 THED Hydraull Deck Lift-01 4B THE0 Reactor Coo eit Pump Motor Hoist Receptacle-4 8H THE0 RC Pipe Penetra ion Cooling Unit-01 8M THED RC Pipe Penetration ooling Unit-02 5H THED RCP #11 Oil Lift Pump-0 SM THED RCP #13 Oil Lift Pump-03 - ( 10B THED Preaccess Filter Train Packages Receptacle - 17

COMANCHE PEAK - UNIT 1 3/4 8-25

IXX-88512 ATTACHhTNT11-1 PAGE 39 0F 52 TABLE 3.8-1 (Continued) ( CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES I DEVJCE NUMBER AND10 CATION

3. 48 VAC from Motor Control Centers (Continued) .

kIDI0396 MCC 1EB1- G.E. COMPT. NO. BKR. TYPE SYSTEM POWERED SB THED Containment L7g XFMR-14 (PHL-C3) 10F THED S.G. Wet Layup Circ. Pump 01 (CPI-CFAPRP-01) 12M THED S.G. Wet Layup Cire. Pump 03 (CPI-CFAPRP-03) 12H T K Cont. Ltg. Transf. CPI-ELTRNT-28 (AULC-11AMC-12) ._ , 60 THED Refueling Machine (Manipulator Crane-01) 2M THED RC Drain Tank Pump No. 1 l 2F THED Containment Ltg XFMR-16 l (PNL C7 & C9) IM THED Containment Ltg XFMR-12 PNL C1 & C5) 3M THED Pr ccess Fan No. 11

3. 2 Device Location -

C 1EB2-2 Containment N bers listed below. Primary and Backup - Both rimary and backup Breakers breake have identical trip ratings nd are located in the same C compt. These breakers a General Elec-tric type TH or THFK with thermal magnet'c trip elements. 1 l MCC 1EB2-2 G.E. l COMPT. NO. BKR. TYPE SYSTEM POWERED 4G THED Motor Operated Valve 1-TV-4'6 2 ( 4M THED Motor Operated Valve 1-TV-4694

                                                                                            \ s COMANCHE PEAK - UNIT 1                  3/4 8-26 l

IXX-88512 ATTACHMENT 11 .

                    @ W 52 TABLE 3.8-1 (Continued)

O

 \J        \

( CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES gG gg D CE NUMBER AND OCATION h gg g%

3. 48 VAC from Motor Control Centers (Continued)

MCC IEB2 ( G. E. COMPT. NO. BKR. TYPE SYSTEr. POWERED 3F THED Containment Orcin Tank Pump-04 7H THED Containment Sump No. 2 Pump-03 7H THED Containment Sump No. 2 Pump-04 6H THED RCP No.12 Motor Space Heater-02 6M T D RCP No.14 Motor Space Heater-04 53 THED Inc[re Detector Drive "C" 2B THE's Incore Detector Drive "D" { 7B THED Incore Detector Drive "E" 50 THED Containment Fuel Storage Crare-01 3B THED Stud Tensioner Hoist Outlet-02 48 THED Containeent Solid Rad Waste C mpactor-01 10B THED RCC hange Fixture Hoist Drive-01 10F THED Refuel Cavity Skimmer Pump-01 12B THED Power Race tacles (Cont. El. 841') IM THED S.G. Wet Layu Circ. Pump 02 (CPI-CFAPRP-02 12M THED S.G. Wet Layup Ci . Pump 04 (CPI-CFAPRP-04) 8H THED RC Pipe Penetration Fas 93

8M THED RC Pipe Penetration Fan-0 SH THED RCP #12 011 Lift Pump-02 d( \ COMANCHE PEAK - UNIT 1 3/4 8-27

                                                                                                  's A

IXX-88512 AITACMENT 11 - PAGC 410F 52 TABLE 3.8-1 (Continued) CONTAINMENT PENETRATION CONOUCTOR (, OVERCURRENT PROTECTIVE DEVICES DEVI NUMBER , AND LO TION , [ IDI J396

3. 480 V frce Motor Control Centors (Continued)

MCC 1EB2-2 G.E. COMPT. NO. BKR. TYPE SYSTEM POWERED SM THED RCP #14 Oil Lift Pump-04 12H THED Preaccess Filter Train Package Receptables - 18 60 THED Containment Auxiliary Upper Crane-01 2F TH Containment Ltg. XFMR-13 (PN,L C-2)., 70 THED Containment Elevator-01 l 2D THED Containment Access Rotating Platform-01 l - 2M THED Reactor Coolant Drain Tank Pump-02 9F THED Containment Ltg. XFMER-17 I (PNL C8 & C10) l l 9H THED ontainment Ltg. XFMR-15 l ( L C4 & C6) 3M THED Prea ess Fan-12 1G THFK Contain nt Welding Machine Power Supply U t 3.3 Device Locatien - MCC IEB3-2 ontainment numbers ! listed below. Primary and Backup Breakers - Unless noted oth'epise, both primary and backuA breakers have identical trip ratiqgs and are located in the same MCC compt. l These breakers are General l Electric type THED or THFK with thermal-magnetic trip elk ents. Oc \ COMANCHE PEAK - UNIT 1 3/4 8-28 ,

TXX-88512 ATTACHMENT 11 PAGE 42 0F 52' TABLE 3.8-1 (Continued) CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVI E NUMBER ANO LO'(ATION

3. 480 C from Motor Control Centers (Continued)

                                                                                  $l k 108 0396 MCC 1EB3-2                 G.E.

COMPT. NO. BKR. TYPE SYSTEM POWERED 8RF THED JB-1S-1005 for Altern. Feed to Motor Operated Valve 1-8702A 1G THED Motor Operated Valve 1-8112 9G THED Motor Operated Valve 1-8701A 9M T D Motor Operated Valve 1-8701B

SM THED Soto'r Operated Valve 1-8000A' SG THED Motor Operated Valve 1-HV-6074 1 C\ V 4G THED Motor Operated Valve 1-HV 6076 4M THED* Hotor Operated Valve 1-HV-6078 2G THED Motor Operated Valve 1-HV-4696 2M THED Motor Operated Valve 1-HV-4701 3G THED Mo r Operated Valve 1-HV-5541 l 3M THED Motor perated Valve 1-HV-5543 1M THED Motor Op ated Valve 1-HV-6083 6F THED Motor Opera d Valve 1-HV-8808A 6M THED Motor Operated alve 1-HV-8808C 7M THED Containment Ltg. MR-18 (PNL SC1 & SC3) l I

      "Primary protection is provided by G>uld Tronic TR5 fusible switch with h2A N

O fuse. COMANCHE PEAK - UNIT 1 3/4 8-29 -

9 TXX-88512 AliACHMENT 11 PAGE 43 0F 52 TABLE 3.8-1 (Continued) CONTAINMENT PENETRATION CONDUCTOR ( OVERCURRENT PROTECTIVE DEVICES DEVICE AND LOCAt{0N MBER I?fl0 C/ TE m m

3. 480 VA from Motor Control Centers (Continued)

MCC 1EB3-2 G.E. COMPT. NO. BKR. TYPE SYSTEM POWERED 8M THED Neutron Detector Well Fan-09 7F THFK Electric H Recombiner Power 2 Supply PNL-01 8RM ED Fire Protection Containment Isolation MOV1-HV-4075C 3.4 Device Location - MCC 1EB4-2 Containment numbers listed below. - - Primary and Backup - Unless noted otherwise, both Br takers primary and backup breakers have identical trip ratings and are O {- located in the same MCC compt. These breakers are General Electric type THED or THFK with thermal-magnetic trip elements. MCC IEB4-2 G.L COMPT. NO. BKR. TYPE SY EM POWERED IM THED JB-15 230G, Altern. Power' Supply Feed t Hov 1-8701B 8G THED Motor Ope ted Valve 1-8702A BM THED Motor Operat Valve 1-87028 4M THED Motor Operated lve 1-8000B 4G THED Motor Operated Valve 1-HV-6075 3G THED Motor Operated Valve -HV-6077 3M THED* MotorOperatedValve1-b 9

Primary protection is provided by Gould Tronic TRS fusible switch with 3.'2 fuse.

( COMANCHE PEAK - UNIT 1 3/4 8-30

IXX-88512 AllACHMENT11 PAGE 44 0F 52 . (~' s TABLE 3.8-1 (Continued) l CONTAINMENT PENETRATION CONDUCTOR yInni I OVERCURRENT PROTECTIVE DEVICES DEVI E NUMBER r AND LOCATION IDI 03%

3. 480\ACfromMotorControlCenters(Continued)

MCC IEBd-2 G.E. COMPT. NO. BKR. TYPE SYSTEM POWERED 2G THED Motor Opreated Valve 1-HV-5562 2M THED Motor Operated Valve 1-HV-5563 SF THED Accumulator Iso. VLV, Mov-1-8808B SM ED Accumulator Iso. VLV, Mov-1-88080 6M TH Containment Ltg. XFMR-19 ~ ~

                                                             '(PNI. SC2 & SC4) 7M                    THED                    Neutron Detector Well Fan 10 m

6F THFK Elect. H Recombiner 2 Power Supply PNL-02 DEVICE NUMBER SYSTEM AND LOCATION POWERED

4. 480 VAC From Panelboards For Pressurizer Pressurizer Heaters Heaters

a. Primary Breakers - General Electric Typ TJJ Thermal Magnetic breaker.

Breaker No. & Location - Ckt. Nos. 2 thru 4 of Panelbcards 1EB1-1, 1EB1-2, 1EB2-2, 1EBit2, 1EB4-1, 1EB4-2 and

.                                                 Ckt. Nos. 2 thru 5 of\ anelboards 1EB2-1 and IEB3-1.

b. Backup Breakers - General Electric Type THJS with ongtime and insts solid state trip device withr 400 Ah . sensor.

BreakerNo.& Location-Ckt.No.1ofPane1 boards 1EBhl,1EB1-2, 1EB2-1, 1EB2-2, 1EB3-1, 1EB3-2,\1EB4-1 and 1EB4-2.

x COMANCHE PEAK - UNIT 1 3/4 8-31 ,

T IXX-88512 AllACHMEHi11 PAGE 45 0F 52 TABLE 3.8-1 (Continued)

CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEV! E NUMBER SYSTEM RE'"DC/h ".': IDI 039 AND LO. CATION POWERED

5. 120 Space Heater Circuits Containment Recirc. Fan from OV Switchgears and CRDM Vent. Fan Motor Space Heaters

a. Pri ry Breakers BKR. L ATION WESTINGHOUSE

 .                   & NUMBE                         BKR. TYPE _

Swgr. 1EB1, EB1010 Cubicle 3A CP1-VAFNAV-01 Space Heater Bk Swgr. 1EB2, EB1010- - '- - Cubicle 3A CP1-VAFNAV-02 Space Heater Bkr. Swgr. 1EB3, EB1010 Cubicle 9A s CP1-VAFNAV-03 Space Heatar Bkr. Swgr. 1EB4, EB10 Cubicle 9A CP1-VAFNAV-04 Space Heater Bkr. Swgr. 1EB3, EB1010 Cubicle 8A, CP1-VAFNCB-01 Space Heater Bkr. Swgr. 1EB4, EB1010 Cubicle dA CP1-VAFNAV-02 Space Heater Bkr. \

b. Backup Breakers

                                                                               '\ \

BKR. LOCATION GENERAL ELECTRIC \s

                    & NUMBER                         BKR. TYPE                         'n Panel 1EC3-2                     TED Ckt. No. 3 Panel IEC3-2                     TED Ckt. No. 4 COMANCHE PEAK - UNIT 1                    3/4 8-32

IXX 88512 AllACHMEHi 11 PAGE 46 0F 52 _

TABLE 3.8-1 (Continued) CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES

r OEVICE N ER SYSTEM AND LOCATI PDWitTD

5. 120V Space eater Circuits from 480V Switchgears (Continued) 1 BKR. LOC ION GENERAL ELECTRIC

                    & NUMBER                     BKR. TYPE                                    l Panel 1EC4-2                 TED                                          l Ckt. No. 3 Panel IEC4-2                 TED                                          l Ckt. No. 4 l

I 1 1 !O COMANCHE PEAK - UNIT 1 3/4 8-33 , l

IXX-88512 " ' 1 " ATTACHMEHi11 7ggg PAGE 47 0F 52 ,

                                                                  -  'Jr/u  j.

TABLE 3.8-1(Continued CONTAINMENT PENETRATION CONDUCTOR N OVERCURRENT PROTECTIVE DEVICES DEVICE UMBER SYSTEM AND LOCAM ON PWElilD

6. 125V Lighting Emergency DC Lighting

a. Pri ry Breaker BREAKE LOCATION G. E. BKR.

AND NUMB R TYPE DC Panelbo j TFJ 102-1, Ckt #

b. Backup Device - /A (Fuse)

7. 125V DC Control Power Various

                                                        "   ~

a. Primary Devices - N/A uses

b. Backup Breakers GENERAL ELECTRIC l CAB. NO. PANELBOARD NO. CKT. NO. BREAKER TYPE 1

l 01 XED1-1 6,7,8,9,10 TED 02 XE02-1 3,6,7,8,9,10 TED 03 X02-3 8, ,12,14,17 TED 04 XE01-1 1,6, ,8,9,10 TED 05 1E02-1 7,10, 2,15,16,17 TED 06 XD2-3 8,9,12, 4,17 TED 07 1E01-1 7,10,14, 7 TED 08 XED2-1 1,3,6,7,8 9,10 TED 09 102-3 7,10,11,14, 7 TED 10 1ED1-1 7,10,14,17 TED 11 1E02-1 7,10,12,15,16, 7 TED 13 1E01-1 7,10,14,17 TED 39A X02-1 11 TED l !O COMAhCHE PEAK - UNIT 1 3/4 8-34 ,

   . . ..           .        .=-                               -

IXX-88512 ATTACH)iENT 11_ . PAGE 48 0F 52 TABLE 3.8-1 (Continued) CONTAINMENT PENETRATION CONDUCTOR

                      ,                  OVERCURRENT PROTECTIVE DEVICES 0     CE NUMBER AND OCATION                                                       ret 00M. E = = >

8. V AC Instrument Distribution Panel Board 1C3-3

a. rimary Device

b. ba up Breaker - GE Tyi.e TED located in instrument Distribution Panel Board 1C3-CK #11 l 9. 120V AC Power r Personnel and Emergency Airlocks

a. Primary Devic s

b. Backup Breakers -

GENERAL ELEC' .<lc PANELBOARD No. i(T. NO. BREAKER Ti.. XEC2 34 TED XECl-2 2 TED

10. 118V AC Control Power

a. Primary Devices

b. Backup Breakers ENERAL ELECTRIC PANELBOARD NO. CKT. NO. SREAKER TYPE XEC12-1 3,5,7,9,10,12 TED XEC2-1 3,5,7,9,10,12 TED 1C2 12,22 TED 1C3 12,14 TED l IPC1 10,13 TED 1PC2 10 TED 1PC4 6,10 TED 1EC1 3,4,8,9 TED 1EC2 3,4,7,9 TED 1EC5 3,8 TED 1EC6 3,8 TED O

COMANCHE PEAK - UNIT 1 3/4 8-35 .

                                                                               ' i IXX-88512 ATTACHMENT 11                                                                                 l 1

PAGE 49 0F 52- - l T TABLE 3.8-1(Continuedj CONTAINMENT PENETRATION CONDUCTOR W II ! OVERCURRENT PROTECTIVE DEVICES ! I DEVICE ER AND LOCAT'10N '1""'0CA"':ici0396> '

11. Emerge Evacuation System i'arning Lights Power

a. Prima Devices

b. Backup Breakers SQUARE D SINGLS POLE PANELBOARD NO. CKT. NO. BREAKER TYPE XEC3 1 FAL-12020 XEC4 3 FAL-12020 IL M. ORPI Data Cabinet Power 5 plies

a. Primary Devices

b. Backup Breakers GENERAL ELECTRIC PANELBOARD NO. CKT NO. BREAKER TYPE 1C14 1,2 TED 1

l 1 l l l lO l l COMANCHE PEAK - UNIT 1 3/4 8-36 , 1

1

            .            TXX-88512 ATTACH S T ll PAGE50Of$1 g

3/4.8 ELECTRICAL POWER SYSTEMS v....

 - Q)
                ,       BASES 3/4.8.1, 3/4.8.2, and 3/4.8.3                                                A.C. SOURCES, D.C. SOURCES, and ONSITE POWER DISTRIBUTION The OPERABILITY of the A.C. and 0.C power sources and associated distribu-tion system Curing operation ensures that sufficient power will be available to supply the safety-related equipment required for: (1) the safe shutdown of the facility, and (2) the mitigation and control of accident conditions within the facility.                        The minimum specified independent and redundant A.C. and 0.C.

power sources and distribution systems satisfy the reouirements of General Des 'n Criterion 17 of 10 CFR 50 Appendix A. The ACTION requirements specified for the levels of degradation of the power sources pr./ide restriction upon continued fat.lity operation commensurate with the level of degradation. The OPERABILITY of the power sources are consistent woh the initial condition assumetions of the safety analyses and are based upoi, maintaining at least one redundant set if onsite A.C. and 0.C. power sources and associated dist.ribution systems.0PERABLE during accident . cond tions coincident with an assumed loss of-offsite power and single failure of the other onsite A.C. source. The A.C. and D.C. source allowable out-of-service times are based on Regulatory Gui'e 1.93, "Availability of Electrical p) Power Source ," Deceaber 1974 and Generic Letter 84-15, "Proposed Staff Position te Improve and Maintain Diesel Generator Reliability." When one diesel generator (( is ' nope ^able, there is an additional ACTION requirement to verify that all

                       .aouired systems, subsystems, trains, components and devices, that depend on the remaining OPERABLE diesel generator as a source of emergency power, are also OPERABLE, and that the steam-driven auxiliary feedwater pump is OPERABLE.

This requirement is intended to provide assurance that a loss-of-offsite power event will not result in a complete loss of safety function of critical systems during the period one of the diesel generators is inoperable. The term,

                    . verify, as used in this context means to administrativcly check by examiriing logs or other information to determine if certain corponents are out-of-service for aintenance or other reasons. It does not mean to perform the Surveillance hequirements needed to demonstrate the OPERABILITY of the comr                                                                                   it.

The OPcRABILITY of the minimum specified A.C. and 0.C. oower sources and assoc h ted distribution systems curing shutdown and refueling ensures that: (1) the freility can be maintained in the shutdown or refueling condition for extenced t' v perkds, and (2) sufficient instrumentation and control capa-bility is a lr. for mod toring and maintaining the unit status. The Sur:- w:e Reg , v% for demonstrating the ' OPERABILITY uf the diesel Mne . Et,e with the recommendations of Regulatory Guides 1.9, i. - S Generator Set C ea':ity for Standby Power Supplies," 'a t ,) ,

                                                                                                          "Periodic Testing of Diesel Generator Units Used as Onsite                          . ,
                                                                                           >       .S tems               at Nuclear Power Plants," Revision 1, August 1977; i                          A.,                              -

Oil Systems for Staridby Diesel Gererators,' Reviti- 1, Octcber e - ic Letter 84-15, and Generic Letter 83-26, (y "L'i ;n on of Survoi- . ace Requirements for Diesel Fuel Impurity Level Tew COMAhoiE PEAK - UNIT 1 B 3. 8-1

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

IXK-88512 ATTACHMENT 11 PAGE $1 0F 52 , p ELECTRICAL DOWER SYSTEMS BASES L C. SOURCES, D.C. SOURCES, and ONSITE POWER DISTRIBUTION (Continued) The Diesel Generator Test schedu!e, Table 4.8-1, is based on the recommendt-tions of Regulatory Guide 1.108, "Periodic Testing of Diesel Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants," Revision 1, August 1977, and NRC Technical Report A-3230, "Evaluation of Diesel Unavailability and Risk Effective Surveillance Test Intervals," May 1986, and Generic Letter 84-15 "Proposed Staff Position to Improve and Maintain Diesel Generator Reliability." The Surveillance Requirement for demonstrt ing the OPERABILITY of the station batteries are based on the recommendati.ns of Regulatory Guide 1.129, "Maintenance Testing and Replacement of Large Lead Storage Batteries for Nuclear Power Hants," February 1978, Regulatory Guide 1.32 "Criteria for Safety Related Electric Power Systems for Nuclear Power Plants," Revision 2, and IEEE Std 450-1980, "IEEE Recommended Practice for Maintenance, Testing, and Replace-ment of Large Lead Storage Batteries for Generating Stations and Substations." The operational requirement to energize the instrumenc busses from their associated inverters connected to its associated D.C. bus is satisfied only when the inverter's output is from the regulated portion of the inverter and not from the unregulated bypass source via the internal static switch. ,, Verifying average electrolyte temperature above the minimum for which the, battery was sized, total battery terminal voltage on float charge, connection resistance values, and the performanca of battery service and discharge tests r3 ensures the. effectiveness of the charging system, the ability to handle high V discharge rates, and compares the battery capacity at that time with the rated capacity. Table 4.8-2 specifies the normal limits fer each designated pilot cell and each connected cell for electro,yte level, float voltage, and specific gravity. The limits for the designated pilot cells float voltage and specific gravity, greater than 2.13 volts and 0.015 below the manufacturer's full charge specific gravity or a battery charger current tnat had stabilized at a low value, is characteristic of a charged cell with adequate capacity. The normal limits for each connected cell for float voltage and specific gravity, greater than 2.13 volts and not more than 0.020 below the manufa:turer's full charge specific gravity with an average specific gravity of all the connected celb not more than 0.010 below the manuf:ccurer's full charge specific gravity, ensures the OPERABILITY and capability of the battery. Operation with a battery cell's parameter outside the normal limit but witisin the allowable value specified in Table 4.8-2 is permitted for up to 7 : day During this 7-day period: (1) the allowable values for electrolyte l level ensurts no phytical damage tn the plates with an adequate electron transfer capability; (2) the allowable value for the average specific gravity of all the cells, .iot more than 0.020 below the manufacturer's recommended full charge t.pecific gravity, ensures that the decrease in rating will be less than the safety margin provided in sizing; (3) the allowable value for an individual cell's specific gravity, ensures that c.r individual cell's specific gravity wil' not be more than 0.040 below the mtnufacturer's full charge specific gravity and that the overall capability of t),e battery will be m31nt ined p d within an acceptable limit; and (4) the aliowable value for an iniividual cell's float voltage, greater than 2.07 volts, ensures the battery's capability to perform its design function. COMANCHE FEAK - UNIT 1 B 3/4 8-2 ,

TXX88512 q Af f ACHMENT 11_ PAGE $2 of $2 h.h

          , ELECTRICAL POWER SYSTEM _S BASES 3/4.8.4    ELECTRICAL EQUIPMENT PROTECTIVE DEVICES o Containment electrical penetrations and            i       actor operation conduc penetration tected by either deenergizing circuits not required dur ng re      current protec-or by demonstrating the OPERABILITY of primary and backup overT tion circuit breakers during periodic surveillance. Electric Penetration    Power recommendations of regulatory guide 1.63 Revisio Plants."                                                                    i breakers The Surveillance Requirements                applicable i        tto   lower least     Q% ofvoltage    c and fuses provide assurance of breaker              reliability Each              by test manuf ac .urer's i

ng mo!oed ples a. case which each manufacturer's and metal brand case circuit of circuit breakers are breaker. grouped iinto k are tested. If ve sa representat are then tested on a rotating basis to ensJre that allbreakers, it breatreaters it each a wide variety exists within any manufacturer,'s_ group as a separate type of breaker for surveillance purposes. d with therma.1 p I All Class IE motor-operated valves motor starters are provide

 '             overload protection which is permanently bypassed and provTherefore,                          t h    will not only at Comanche Peak Steam Electric Station.

OPERABILITY or Surveillance Requirements i for (refer thesa to i Motors on devices prevent safaty-related valves from performing th Motor Operated Valves," Revision 1, March 1977). O 8 3/4 8-3 , CO MNCHE PEAK - UNIT 1 1" - - --

IXI-88512 l ATTACMENT 12 PAGE 1 0F 20 O i l COMANCHE PEAK STEAM ELECTRIC STATION TtCHNICAL SPECIFICATI0N , ._ , 3/4.9 l lO l O I

r, ,

       . T!X 88512 ATi&CHMDli 12 PAGE 2 of 20 CPSES Technical Specificat. ions NRC Draft 2 Markup (g                                                                                        Section 3/4.9 Change 10#                                 Justification For Change 0945                                 This Technical Specification is being deleted since.CPSES' does not take credit for this system in the accident analysis for the fuel building area to meet the limits 10CFR100. This is consistent with what was approved for
                                                    -CPSES in the Final Draft Technical Specifications of 1984.

, For a more detailed discussion, see FSAR section 9.4.2. O l l l 1 ' l lO - 1 L- .._.. .__ .,.._._._._ _ __._ ____ _.. _ _. _ __. _ _ _ . -... _ _ ._ _ . .._. . _ -

IXX-885I2 , i-----

AllACH S T 12 I

PAGE J Of 20 l 3/4.9 REFUELING OPERATIONS l 3/4.9.1 80RON CONCENTRATION LIMITING CONDITION FOR OPERATION 3.9.1 The boron concentration of all filled portions of the Reactor Coolant System and the refueling canal shall be maintained uniform and sufficient to I ensure that the more restrictive of the following reactivity Onditions is met, i either: l

a. A K,ff of 0.95 or less, or !

b. A boron concentration of greater than or equal to 2000 ppm.* (

Additionally, either valve 1C5-8455 or valves 1C5-8560, FCV-1118, 1C5-8439, 1C5-8441 and 1C5-8453 shall be closed and secured in position. APPLI RBILITY.: MODE 6. ACTION:

a. With the requirements a or b of the above not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reacti-~

vity changes and initiate and contirfue'boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm boron or its equivalent until K,7f is reduced to less than or equal to 0.95 or the boron concentration is restored to greater O than or equal to 2000 ppm, whichever is the more restrictive. (' b. If either valve 1C5-8455 or valves 1C5-8560, FCV-1118, ICS-8a39, 105-8441 and 1C5-8453 are not closed and secured in position, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes and take action to isolate the dilution paths. Within 1 hour, verify the more restrictive of 3.9.1.a or 3.9.1.b or carry out Action a. above. SURVEILLANCE sEQUIREMENTS

4. 9.1.1 The more restrictive of the above two reactivity conditions shall be determined prior to:

l a. Removing or unbolting the reactor vessel head, and ( . b. Withdrawal of any full-length control rod in excess of 3 feet from j its fully inserted position within the reactor vessel.

4. 9.1. 2 The boron concentration of the Reactor Coolant System and the refueling l canal shall be determined by chemical analysis at least once per .2 hours.

4. 9.1. 3 Either valve 1C5-8455 or valves 1C5-8560, FCV-1118,1C5-8439,1C5-8441

! and 1C5-8453 shall be verified closed and secured in position by mechanical l stops or by removal of air or electrical power at least once per 31 days to l verify that dilution paths are isolated.

          "During initial fuel load, the boron concentration limitation for the refueling l     (     canal is not applie:able provided the refueling canal level is verified to be below the reactor flange elevation at least once per 12 hours.

! COMANCHE PEAK - UNIT 1 3/4 9-1 .

TXX-88512 ! ATTACHMfNi12 l PAGE 4 0F 20 l REFUELING OPERATIONS 3/4.9.2 INSTRUMENTATION (. LIMITING CONDITION FOR OPERATION 3.9.2 As a minimum, two source range neutron flux monitors shall be OPERABLE, each with continuous visual indication in the control room and one with audible indication in the containment and control room. APPLICABILITY: MODE 6. ACTION:

a. With one of the above requireo monitors inoperable or not operating, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

b. With both of the above required monitors inoperable or not operating, determine the boron concentration of,the Reactor Coolant System.at least once per 12 hours.

O . I SURVEILLANCE REQUIREMENTS 4.9.2 Each source range neutron flux monitor shall be demonstrated OPERABLE by performar.ce of:

a. A CriANREL CHECK at'least once per 12 hours,: b. An ANALOG CHANNEL OPERATIONAL TEST within 8 hours prior to the initial j start of CORE ALTERATIONS, and; c. An ANALOG CHANNEL OPERATIONAL YEST at least once per 7 days.

Oc . l C0HANCHE PEAK - UNIT 1 3/4 9-2 , l

in-88512 ATTACHl!ENT 12 PAGE 5 0F 20 REFUELING OPERATIONS 3/4.9.3 DECAY TIME LIMITING CONDITION FOR OPERATION 3.9.3 The reactor shall be suberitical for at least 100 hours. APPLICABILITY: During movement of irradiated fuel in the reactor vessel. ACTION: With the reactor subcritical for less than 100 hours, suspend all operations involving movement of irradiated fuel in the reactor vessel. O SURVEILLANCE REQUIREMENTS 4.9.3 The reactor shall be determined to have been suberf tical for at least 100 hours by verification of the date and time of suberiticality prior to movement of irradiated fuel in the reactor vessel. 4 ( COMANCHE PEAK - UNIT 1 3/4 9-3 ,

 . _   Tu-88512                                   _ . . . .                     .

ATTACHMENT 12 PAGE 6 0F 20 REFUELING OPERATIONS 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS LIMITING CONDITION FOR OPERATION 3.9.4 lne containment building penetrations shall be in the following status:

a. The equipment hatch' closed and held in place by a minimum of four bolts,

b. A minimum of one door in each airlock is closed, and

c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere shall be either:

1) Closed by an isolation valve, blind flange, or manual valve, or

2) Be capable of being closed by an OPERABLE automatic containment ventilation isolation valve.

APPLICABILITY: During CORE ALTERATIONS or moVemint of irradiated fuel within" the containment. ACTION: With the requirements of the above specification not satisfied, immediately I suspend all operations involving CORE ALTERATIONS or movement of irradiated ! fuel in the containment building. SURVEILLANCE REQUIREMENTS 4.9.4 Each of the above required containment building penetrations shall be determined to be either in its closed / isolated condition or capable of being closed by an OPERABLE automatic containment ventilation isolation valve within 100 hours prior to the start of and at least once per 7 days during CORE ALTERATIONS or movement of irradiated fuel in the containment building by:

a. Verifying that,

1. Containment ventilation isolation occurs on a high radiation test signa' from a containment atmosphere gaseous monitoring instrumen-tation channel and the containment ventilation isolation valve (s) can be closed remotely from the control room, or

2. the containment ventilation isolation valve (s) are closed /

isolated. l b. Verifying the remaining penetrations of 3.9.4 not covered by a. above, are in their closed / isolated condition. COMANCHE PEAK - UNIT 1 3/4 S-4 ,

TXX 88512 - .

   .                       AliACHMENT!?

PAGE 7 0F 20 REFUELING OPERATIONS 3/4.9.5 COMUNICATIONS LIMITING CONDITION FOR OPERATION 3.9.5 Direct communications shall be maintained between the control room and personnel at the refueling station. APPLICABILITY: During CORE ALTERATIONS. ACTION: When direct communications between the control room and personnel at the refueling station cannot he maintained, suspend all CORE ALTERATIONS. I SURVEILLANCE REQUIREMENTS { 4.9.5 Direct communications between the control room and personrel at the refueling station shall be demonstrated within 1 hour prior to the start of and at least once per 12 hours during CORE ALTERATIONS. 1 1 l l ( COMANCHE PEAK - UNIT 1 3/4 9-5 ,

IXX-88512 - ' '

   . AfiACHilENT12 PAGE 8 0F 20 REFUELING OPERATIONS 3/4.9.6 REFUELING MACHINE LIMITING CON 0! TION FOR OPERA.' ION 3.9.6 The refueling machine main hoist and auxiliary monorail hoist shall be used for movement of drive rods or fuel assemblies and shall be OPERABLE with:

a. The refueling machine main hoist used for movement of fuel assemblies having:

1) A minimum capacity of 2850 pounds, a.M

2) An overload cutoff limit less than or equal to 2800 pounds,

b. The auxiliary monorail hoist used for latching, unlatching and movement of control rod drive shafts having:

1) A minimum capacity of 610 pounos, and , ._ ,

2) A load indicator which shall be used to prevent lifting loads in excess of 600 pounds, y APPLICABILITY: During movement of fuel assemblies and/or latching, unlatching

( or movement of control rod drive shaf ts within the reactor vessel. ACTION: With the requirements for refueling machine main hoist and/or auxiliary monorail hoist OPERABILITY not satisfied, suspend use of any inoperable refueling machine main hoist and/or auxiliary monorail hoist from operations involving the movement of fuel assemblies and/or latching, unlatching, and movement of control rod drive shafts within the reactor vessel. SURVEILLANCE REQUIREMENTS 4.9.6.1 The refueling machine main hoist used for movement of fuel assemblies within the reactor vessel shall be demonstrated OPERABLE within 100 hours prior to the start of such operations by performing a load test of at least 2850 pounds and demonstrating an automatic load cutoff when the 4r4Ae. s load exceeds 2800 pounds. % Md . 4.9.6.2 The auxiliary monorail hoist and associated load indicator useo for i latching, unlatching, movement of control rod drive shafts within the reactor

vessel shall be demonstrated OPERABLE within 100 hours prior to the start of such operations by performing a load test of at least 610 pounds.

O' ( COMANCHE PEAK - UNIT 1 3/4 9-6 _-..-,..-y .

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

TXX 88512

   ,      Af fACHENT 12 PAGE 9 0F 20 REFUELING OPERATIONS 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE AREAS LIMITING CON 0! TION FOR OPERATION 3.9.7 Loads in excess of 2150 pounds shall be prohibited from travel over fuel asserblies in a storage pool.

APPLICABILITY: With fuel assemblies in a storags pool. ACTIg:

a. With the requirements of the lhove specification not satisfied, place the crane load in a safe condition,

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

O SURVEILLANCE REQUIREMENTS l l l 4.9.7 Each hoist load indicator used for loads over spent fuel storage pools shall be demonstrated OPERABLE within 7 days prior to the start of such opera-t tions and at least once per 7 days thereafter during operation by performing a j load test of at least 2200 pounds. l l f l O i COPANCHE PEAK - UNIT 1 3/4 9-7 l

l

IIX8?512 AllACHMENT12 , PAGE 10 0F 20 l O REFUELING OPERATIONS V $l 3/4.9.8 RESIOUAL HEAT REMOVAL AND COOLANT CIRCULATION ( HIGH WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.8.1 At least one residual heat removal (RHR) loop shall be OPERABLE and in operation." APPLICABILITY: MODE 6, when the water level above the top of the reactor vessel flange is greater than or equal to 23 feet. ACTION: With no RHR loop OPERA 8LE and in operation, suspend all operations involving an increr.se in the reactor decay heat load or a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to OPERA 8LE and operating status as soon as._ possible. Close all containment penetrationsproviding direct access fror: , the containment atmosphere to the outside atmosphere within 4 hours. Oc ' SURVEILLANCE REQUIREMENTS l 4.9.8.1 At least one RHR loop shall be verified in operation and circulating reactor coolant at a flow rate of greater than or equal to 3800 gpm at least once per 12 hours. - l l l

                *The RHR loop may be recoved from operation for up to I hour per 8-hour period during the performance of CORE ALTERATIONS in the vicinity of the reactor vessei hot legs.

COMANCHE PEAK - UNIT 1 3/4 9-8

IXX-88512 AIIACHMENT12 PAGE 11 0F 20 RfEfUELINGOPERATIONS LOW WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.8.2 Two independent residual heat removal (RHR) loops shall be OPERABLE, and at least one RHR loop shal.1 be in operation." APPLICABILITY: MODE 6, when the water level above the top of the reactor vessel flange is less than 23 feet. ACTION: - a. With less than the required RHR loops OPERABLE, immediately initiate corrective action to return the required RHR loops to OPERABLE status, or to establish greater than or equal to 23 feet of water above the reactor vessel flange, as soon as possible.

b. With no RHR loop in operation, suspend all operations involving a reduction in boron concentration of th,e. Reactor Coolant System and ,

immediately initiate corrective action to return the required RHR loop to operation. Close all containment penetrations providing direct access from the containment atmosphere to the outside 6tmosphere within 4 hours. ( { SURVEILLANCE REQUIREMENTS 4.9.8.2 At least one RHR loop shall be verified in operation and circulating reactor coolant at a flow rate of greater than or equal to 3800 gpm at least once per 12 hours. -

           =
                       "Prior to initial criticality, the RHR loop may be removed from operation for up to I hour per 8-hour period &lring the performance of CORE ALTERATIONS in the vicinity of the reactor vessel hot legs.

( COMANCHE PEAK - UNIT 1 3/4 9-9 w.- . - _ _ _ .

IXX 88512 _ ATTACHMENT!? PAGE 12 0F 20 Q REFUELING OPERATIONS 3/4.9.9 WATER LEVEL - REACTOR VESSEL FUEL ASSEMBLIES LIMITING CONDITION FOR OPERATION 3.9.9.1 At least 23 feet of water shall be maintained over the top of the reactor vessel flange. APPLICABILITY: During movement of fuel assemblies within the containment when either the fuel assemblies being moved or the fuel assemblies seated within the reactor vessel are irradiated. ACTION: With the requirements of the above specification not satisfied, suspend all operations involving movement of fuel assemblies within the containment. O SURVEILLANCE REQUIREMENTS 4.9.9.1 The water level shall be determined to be at least its minimum l required depth within 2 hours prior to the start of and at least once per 24 hours thereafter during movement of fuel assemblies within the containment. l [ l l 1O COMANCHE Pr..'sK - UNIT 1 3/4 9-10 .

 .. TIX-88512                               - - - - -

1 ATTACHMENT 12 PAGE 13 0F 20 3 WA - REACTOR VESSEL l CONTROL RODS LIMITING C1 0ITION FOR OPERATION _ 3.9.9.2 At least 23 feet of water shall be maintained over the top cf the irradiated fuel assemblies within the reactor vessel. APPLICABILITY: During movement of control rods within the reactor vessel while in MODE 6. ACTION: With the requirements of the above specification not satisfied, suspend all operations involving movement of control rods within the reactor vessel. O SURVEILLANCE REQUIREMENTS 4.9.9.2 The water level shall be de'.ewined to be at least its minimum required depth within 2 hours prior to the start of and at least once per 24 hours thereafter during movement of control rods. O COMANCHE PEAK - UNIT 1 3/4 9-11 .

i 1XX-68512 ~ ATTACHMENT 12 PAGE 14 0F ?0_ , REFUELING OPERATIONS 3/4.9.10 WATER LEVEL - IRRADIATED FUEL STORAGF Unn)l LIMITING CONDITION FOR OPERATION 3.9.10 At least 23 feet of we.ter shall be maintained over the top of irradiated fuel assemblies seated in the storage racks. APPLICABILITY: Whenever irradiated fuel assemblies are in the storage racks. ACTION:

a. With the requirements of the above specification not satisfied, suspend all novament of fuel assemblies and crane operations with loads in the fuel storage areas and restore the water level to within its 1 lait within 4 hours.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

!O SURVEILLANCE REQUIREMENTS 4.9.10 The water level above the storage racks shall be determined to be at least its minimum required depth at least once per 7 days when irradiated fuel l assemblies are in the fuel storage racks.

l l

l l l i l l COMANCHE PEAK - UNIT 1 3/4 9-12 ,

i IXX-88512 - .. ._ . ATTACHMENT 12 PAGE 15 Of 20 , 10809t$ C REFUELING OPERATION 5 fy fU 3/4.9.11 FUEL STORAGE POOL AIR CLEANUP SYSTEM y ,t N LIMITING CONDITION FOR OPERATION /

3.9.11 ndependent Fuel Storage Pool Air Cleanup Systems 11 be OPERABLE. APPLICABILITh Whenever irradiated fuel is in the storage pool. ACTION:

a. With one huel Storage Pool Air Cleanup Systes inoperable, fuel movement, within the storage pool or crape operation with loads over the Storage p'ool may proceed provided the OPERABLE Fuel Storage Pool Air Cleanup Sy' stem is capable of bei g powered from an OPERABLE emergency power' source and is in o ration and discharging through at least one train of HEPA filter and charcoal adsorbers. .. ,

b. WithnoFuelStorage\PoolAir leanup System OPERABLE, suspend all operations involving eqvemen of fuel within the storage pool or crane operation with lo d ver the storage pool until at least one Fuel Storage Pool Air C1 up System is restored to OPERABLE status.

c. The provisions of Spec tica ions 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.11 TheaboverequirdFuelStoragePoo1\ir01eanupSystemsshallbe demonstrated OPERABLE: \

a. At least ceper31daysonaSTAGGERE4TESTBASISbyinitiating, from tho/ control room, flow through the H PA filters and charcoal adsorbe'rs and verifying that the system o(p(rates for at least 10 e tinuous hours with the heaters operatt g;

 .          b. A least once per 18 months or (1) after any s uctural maintenance n the HEPA filter or charcoal adsorber housing         or (2) following
                / painting, fire,orchemical release in any ventil tion zone
              / communicatin] with the system by:                       *
            /
         /

p' O COMANCHE PEAK - UNIT 1 3/4 9-13

Txx-88512 . -. '

ATIACHMENT12 FAGE 16 Of 20 '

o V

         \REFUELING OPERATIONS                                                             ggp3l       W Om N

SURVE1 HANCE REQUIREMENTS (Continued)

1) Verifying that the cleanup system satisfies thef n place tatice criteria

                          \ penetration gof less than and

(*]% bypass and us3sleakage testing accep/ guidance in the test procedur Regulatory Positions C.5.a. C.5.c and C.S.4 of Regulatory Guide 1.52, Revision 2, March 1978, and t system flow rate is\ cfm i 10%;

2) Verify,ing, within 31 days after remov/l, that a laboratory analysis, of a representative carbon diample obtained in accor-dance with Regulatory Position C.6f d of Regulatory Guide 1.52, Revision 2g March 1978, meets the/ laboratory testing criteria of Regulatory Position C.6.a ofjAegulatory Guide 1.52, Revision 2, March 1978, for a methyl iodide' penetration of less than [**]%;

and \

3) Verifying a systeg flow ra e of- - cfm i 10% during system -

operation when tested in 6ccordance with ANSI N510-1975.

c. After every 720 hours of ch rcoal adsorber operation by verifying, within 31 days after remoy'al, that a laboratory analysis of a representative carbon semple 'obtained in accordance with Regulatory Position C.6.b of Regu)'atory Gbide 1.52, Revision 2, March 1978,

{ meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guid .52, Revistort 2, March 1978, for a methyl iodide penetrati f less than [*A)%. l d. At least once per 18 months by:

1) Verifyip that the pressure drop across the combined HEPA filte $ and charcoal adsorber banks fs less than [6] inches Wate Gauge while op6 rating the systed t a flow rate of cfm i 10%,

2) rifying that on a High Radiation test si al, the system automatically starts (unless already operati p) and directs its l

exhaust flow through the HEPA filters and cha oal adsorber I banks, l l i o< l l COMANCHE PEAK - UNIT 1 3/4 9-14

TXX-88512 ~ ~ ' AliACMENT 12 N GE 17 of 20 FUELING OPERATIONS (

SURVEft, LANCE REQUIREMENTS (Continued) N / Verifying that the system maintains the spent fuel /storage pool 3( s area at a negative pressure of greater than or endal to [1/4]

                           \ inch Water Gauge relative to the outside atmosp 'ere during system
                             ' operation,
                                \

4) Verifying that the filter cooiing bypass ve ves can be manually opened, and

5) Verifying that the heaters dissipate i kW when tested to accordance with ANSI N510 C

e. Af ter each complete or partial replacement of a HEPA filter bank, by verifying that the. cleanup system sapi'sfies the in-place penetration and bypass leakage' testing acceptance criteria of less than [*]% in accordance with ANSIsN510-1975 for/a 00P test aerosol while operating thesystemataflowKateof / sfs.2 10%. .. .

f. N / After each complete or pertia31 replacement of a charcoal adsorber bank, by verifying that t e cleanup system satisfies the in place penetration and bypass le ge testing acceptance criteria of less O( L than (*]% in accordance witft ANSI g M510-1975 for a halogenated hydrocarbon refrigerant/test gas while operating the system at a flow rate of cf 2 10%. l l l

         *0.05% value applip ble when a HEPA fiiter or cha p al adsorber efficiency of 99% is assumed, or 1% when a HEPA filter or chatcoal adsorber efficiency of 95% of less As assumed in the NRC staff's safety \ evaluation. (Use the value assumedpfor the charcoal adsorber efficiency ifsthe value for the HEPA filter /l s different from the charcoal adsorber of icienty in the NRC staff's        tety evaluation).
        **Value       plicable will be determined by the following equat n:

P= , when P equals the value to be used in the test r uirement ( E is efficiency assumed in the SER for methyl iodide re al (%), nd SF is the safety factor to account for charcoal degradatio between tests (5 for systems with heaters and 7 for systems without heat s). lO< 1 l COMANCHE PEAK - UNIT 1 3/4 9-15

IXX-88512 AITACHMENT12 PAGE 18 0F 20 3/4.9 REFUELING OPERATIONS BASES 3/4.9.1 BORON CONCENTRATION The limitations on reactivity conditions during REFUELING ensure that:

(1) the reactor will remain subcritical during CORE ALTERATIONS, and (2) a uniform boron concentration is maintaissd for reactivity control in the sater volume having direct access to the reactor vessel. These limitations are consistent with the initial conditions assumed for the boron dilutiar, incident in the safety analyses. The value of 0.95 or less for K gf incluves a 1% ok/k conservative allowance for uncertainties. Similarly, the boron concentration value of 2000 ppm or greater includes a conservative uncertainty allowance of 50 ppm boron. The locking closed of the required valvas during refueling operations precludes the possibility of uncontrolled boron dilution of the filled portion of the RCS. This action prevents flow to the RCS of unborated water by closing flow paths from sources of unborated water.

                                                                                  ~.                              -   .

344.9.2 INSTRUMENTATION The OPERABILITY of the source range neutron flux monitors ensures that i redundant monitoring capability is available to detect changes in the reactivity condition of the core. l ( 3/4.9.3 OECAY TIME l l The minimum requirement for reactor subcriticality prior to movement of ( irradiated fuel assemblies in the reactor vessel ensures that sufficient time has elapsed to allow the radioactive decay ;f the short-lived fission products. This decay time is consistent with the assumptions used in the . safety analyses. 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS The requirements on containment building penetration closure and OPERABILITY ensure that a release of radioactive material within containment will be restricted from leakage to the environment. The OPERABILITY and closure restrictions are sufficient to restrict radioactive material release from a fuel element rupture based upon the lack of containment pressurization potential while in the REFUELING MODE. 3/4.9.5 COMMUNIC /TIONS The requirement for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the O facility status.or core reactivity conditions during CORE ALTERATIONS. , COMANCHE PEAK - UNIT 1 B 3/4 9-1 , l [ l

IXX-88512 i AllACHMENT 12 . PAGE 19 0F 20 REFUELING OPERATIONS BASES 3/4.9.6 REFUELING MACHINE The OPERABILITY requirements for the refueling machine main hoist and auxiliary monorail hoist ensure that: (1) the main hoist will be used for movement of fuel assemblies, (2) the auxiliary monorail hoist will be used for latching, unlatching and movement of control rod drive shafts, (3) the main hoist has sufficient load capacity to lift a fuel assembly (with control rods), (4) the auxiliary monorail hoist has sufficient capacity to latch, unlatch and move the control rod drive shafts, and (5) the core internals and reactor vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations. 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE AREAS The restriction on movement of loads in excess of the nominal weight of a fuel and control rod assembly and associated handling tool over other fuel assemblies in a storage pool ensures that in thr event this load is dropped: ' (1) the activity release will be limited to that contained in a single fuel assembly, and (2) any possible distortion of fuel in the storage racks will not result in a critical array. This assumption is consistent with the activity release assumed in the safety analyses. 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION The requirement that.at least one residual heat removal (RHR) loop be in operation ensures that: (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor vessel below 140*F as required during the REFUELING MODE, and (2) sufficient coolant circulation is maintained through the core to minimize the effect of a boron dilution incident and prevent boron stratification. The requirement to have two RHR loops OPERABLE when there is less than 23 feet of water above the reactor vessel flange ensures that a single failure of the operating RHR loop will not result in a complete loss of residual heat removal capability. With the reactor vessel head removed and at least 23 feet of water above the reactor pressure vessel flange, a large heat sink is avail-able for corts cooling. Thus, in the event of a failure of the operating RHR loop, adequate time is provided to initiate emergency procedures to cool the core. O COMANCHE PEAK - UNIT 1 8 3/4 9-2 ,

1 IXX-88512 ;

         -                                 "IACHMENT 12 RAGE 20 0F 20 REFUELING OPERATIONS BASES 3/4.9.9 and 3/4.9.10 WATER LEVEL - REACTOR VESSEL and IRRADIATED FUEL STORAGE The restrictions on minimum water level ensure that sufficient water depth is available to remove 99% of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly. The minimum water depth is consistent with the assumptions of the safety analysis.

108 0945

                                    -4/AL11_-STORAGE-P00t-VENTitATION SYSTEM The-Heitations-en-the-6terage-Poel-Venti 4ation-Systes-enture that all ris eioac tive-ma te rial-re lea sed-f+ce: On irradfottd-fvel asse21y will be filtered threvgh the-HEPA-fHters-and-charecel-edsorber reier te discherg to the atcespherer-Operation-of-the-system-with-the-heaters- eperating for et least 19-continuous hours in-a-31-day-period-ts-stifficient te reduce the buildup of meisture-en-the-edsorbers-and-HEPA-fMters.-The-OPERA 81ti-TV ei Uii> ayateh               -

and-the-resulting-ledino-remove 4-capec4ty :r; cont 4 stent with the-assumptions of-the-safety-analysescr-ANH-M&lO4975 will be used-es e precedurel gwide for-surveH4ance-test-iner

i COMANCHE PEAK - UNIT 1 B 3/4 9-3

IXX-88M2 ATTACH"2NT l3 PAGE 1 Uf 8 O COMANCHE PEAX STEAM ELECTRIC STATION TECHNICAL SPECIFICATION - - 3/4.10 O l O l

TXX-88512 AllACH7Di 13 PAGE 2 Of 8 CPSES Technical Specification O NRC Draft 2 Markup Section 3/4.10 l Change ID! Justification For Change j No Changes i

l I l l i l . t O 1 1 1 l l 1 l l l l O l

TXX-88512 - ATTACMENT 13 FAGE 3 0F 8 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTOOWN MARGIN ( LIMITING CONDITION FOR OPERATION 3.10.1 The SHUTOOWN MARGIN requirement of Specification 3.1.1.1 may be suspended for measurement of control rod worth and SHUTDOWN MARGIN provided reactivity equivalent to at least the highest estimated control rod worth is Svailable for trip insartion from OPERABLE control rod (s). APPLICABILITY: MODE 2. ACTION:

a. With any control rod not fully inserted and with less than the above reactivity equivalent available for trip insertion, immediately ini-tiate and continue boration at greater than or equal to 30 gpm of a solution containing greater th,an or equal to 7000 ppm boron or its equivals.nt until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored. - - ~ '; b. With all control rods fully inserted and the reactor subcritical by less than the above reactivity equivalent, immediately initiate and
continue boration at greater than or equal to 30 gpm of a solution s containing greater than or equal to 7000 ppm boron or its equivalent: ( until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.

SURVEILLANCE REQUIREMENTS l 4.10.1.1 The position of each control rod either partially or fully wiEhdrawn shall be determined at least once per 2 hours. l 4.10.1.2 Each control rod not fully inserted shall be demonstrated capable of full insertion when tripped from at least the 50% withdrawn position within i 24 hours prior to reducing the SHUTDOWN MARGIN to less than the limits of ( Specification 3.1.1.1. l l l ( COMANCHE PEAK - UNIT 1 3/4 10-1 .

TXX-88512 '~~

   . ATTACHMENT 13 PAGE 4 0F 8 SPECIAL TEST EXCEPTIONS 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS                ' "

k LIMITING CONDITION FOR OPERATION t

3.10.2 The group height, insertion, and power distribution limits of Specifications 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.1, and 3.2.4 may be suspended during the performance of PHYSICS TESTS provided:

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

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

APPLICABILITY: MODE 1. ACTION: - With any of the limits of Specification 3.2.2 or 3.2.3 being exceeded while the requirements of Specifications 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.1, and 3.2.4 are suspended, either:

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

b. Be in HOT STANDBY within 6 hours.

SURVEILLANCE REQUIREMENTS 4.10.2.1 The THERMAL POWER shall be determined to be less than or equal to 85% of RATED THERMAL POWER at least once per hour during PHYSICS TESTS. 4.10.2.2 The requirements of the below listed specifications shall be performed at least once per 12 hours during PHYSICS TESTS:

a. Specifications 4.2.2.2

b. Specification 4.2.2.3, and

c. Specification 4.2.3.2.

( COMANCHE PEAK - UNIT 1 3/4 10-2 .

IXX-88512 ; Af fACHMEHi 13 PAGE 5 0F 8 SPECIAL TEST EXCEPTIONS 3/4.10.3 PHYSICS TESTS LIMITING CONDITION FOR OPERATION 3.10.3 The limitations of Specifications 3.1.1.3, 3.1.1.4, 3.1.3.1, 3.1.3.5, and 3.1.3.6 may be suspended during the performance of PHYSICS TESTS provided:

a. The THERMAL POWER does not exceed 5% of RATED THERMAL POWER,

b. The Reactor Trip Setpints on the OPERABLE Intermediate and Power Range channels are set in accordance with Table 2.2 1 Functional Units 5 and 2b, and

c. The Reactor Coolant System lowest operating loop temperature (T**9) is greater than or equal to 541*F.

APPLICABILITY: MODE 2.

                                                         ~ '"                      '    '

ACTION:

a. With the THERMAL POWER greater than 5% of RAYED THERMAL POWER, immediately open the reactor trip breakers.

l { b. With a Reactor Coolant System operating loop temperature (Tavg) less than 541*F, restore T,yg to within its limit within 15 minutes or be in at least HOT STANDBY within the next 15 minutes. SURVEILLANCE REQUIREMENTS

l 4.10.3.1 The THERMAL POWER shall be determined to be less than or equal to 5% of RATED THERMAL POWER at least once per hour during PHYSICS TESTS. 4.10.3.2 Each Intermediate and Power Range channel shall be subjected to an ANALOG CHANNEL OPERATIONAL TEST within 12 hours prior to initiating PHYSICS TESTS. 4.10.3.3 The Reactor Coolant System temperature (T,yg) shall be determined to be greater than or equal to 541*F at least once per 30 minutes during PHYSICS TESTS. O< COMANCHE PEAK - UNIT 1 3/4 10-3

l

Txx48512 ATTACHMENT 13 PAGE 6 0F 8 SPECIAL TEST EXCEPTIONS 3/4.10.4 REACTOR COOLANT LOOPS k LIMITING CONDITION FOR OPERATION 3.10.4 The limitations of Specification 3.4.1.2 may be suspended during the performance of hot rod drop time measurements in MODE 3 provided at least two reactor coolant loops as listed in Specification 3.4.1.2 are OPERABLE. APPLICABILITY: During performance of hot rod drop time measurements. ACTION:

        -With less than the above required reactor coolant loops OPERABLE during the performance of hot rod drop time measurements, immediately open the reactor trip breakers and comply with the provision of the action statements of Specification 3.4.1.2.

SURVEILLANCE REQUIREMENTS O( 4.10.4 At least the above required reactor coolant loops shall be determined OPERABLE within 4 hours prior to tne initiation of hot, rod drop time measure-ments by verifying current breaker alignments and indirated power availability and by verifying the indicated secondary side water level to be greater than or equal to 10% narrow range. 4 Oc COMANCHE PEAK - UNIT 1 3/4 10-4 .

IXX-68512 ATTACHMENT 13 FAGE 7 0F 8 SPECIAL TEST EXCEPTIC % 3/4.10.5 POSITION IN0! CATION SYSTEM - SHUTOOWN k LIMITING CONDITION FOR OPERATION 3.10.5 The limitations of Specification 3.1.3.3 may be suspended during the performance of individual shutdown and control rod drop time measurements provided;

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 rod drop time measurements. ACTION: With the required digital rod position indicator (s) inoperable or with moFe than one bank of rods withdrawn, immediately open the Reactor trip breakers. (' SURVEILLANCE REQUIREMENTS 4.10.5 The above required digital rod position indicator (s) shall be determined to be OPERABLE within 24 hours prior to the start of ar.d at least once per 24 hours thereafter during rod drop time measurements by verifying the Demand Position Indication System and the Digital Rod Position Indication System agree:

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

b. Within 24 steps during rod motion.

     *This requirement is not applicable during the initial calibration of the Digital Rod Position Indication System provided:        (1) X   is maintained lessthanorequalto0.95,and(2)onlyonesnutdownoff[ontrolrodbank is withdrawn from the fully inserted position at one time.

Oc COMANCHE PEAK - UNIT 1 3/4 10-5 ,

III-88512 .

.          ATTACMENT 13 PAGE 8 0F 8 3/4.10 SPECIAL TEST EXCEPTIONS v

BASES 3/4.10.1 SHUT 00WN MARGIN This special test exception provides that a minimum amount of control rod worth is immediately available for reactivity control when tests are performed for control rod worth measurement. This special test exception is required to permit the periodic verification of the actual versus predicted core reactivity condition occurring as a result of fuel burnup or fuel cycling operations. 3/4.10.2 GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS This special test exception permits individual control rods to be posi-tioned outside of their normal group heights and insertion limits during the performance of such PHYSICS TESTS as those required to: (1) measure control rod worth, and (2) determine the reactor stability index and damping factor under xenon oscillation conditions. 3/4.10.3 PHYSICS TES,TS This special test exception permits PHYSICS TESTS to be performed at less than or equal to 5% of RATED THERMAL POWER with the RCS T,yg slightly lower U than normally allowed so that the fundamental nuclear characteristics of the core and related instrumentation can be verified. In order for various char-acteristics to be accurately measured, it is at times necessary to operate outside the normal restrictions of these Technical Specifications. For instance, to measure the moderator temperature coefficient at BOL, it is necessary 4e position the various control rods at heights which may not l normally be allowed by Specification 3.1.3.6 and the RCS T,yg may fall slightly below the minimum temperature of Specification 3.1.1.4. 3/4.10.4 REACTOR COOLANT LOOPS This special test exception is required to perform certain STARTUP and PHYSICS TESTS under no flow conditions. 3/4.10.5 POSITION INDICATION SYSTEM - SHUTDOWN This special test exception permits the Digital Rod Dosition Indicator (s) to be inoperable during rod drop time measurements. The exception is required since the data necessary to determine the rod drop time a're derived from the induced voltage in the position indicator coils as the rod is dropped. This induced voltage is small compared to the normal voltage and, therefore, cannot be observed if the Digital Rod Position Indicator (s) remain OPERABLE. The exception to the requirement for the Digital Rod Position Indicator to be OPERABLE during the withdrawal of the rods for the initial calibration of the position indication system is required because the OPERABILITY of the Digital O,' Rod Position Indication System can only be determined by withdrawing the control rod. The limitation on Keff during this evolution provides the necessary assurance that inadvertent critically will be avoided. COMANCHE PEAK - UNIT 1 B 3/4 10-1 .

IXX-88512 AliACMENT 14 PAGE 1 0F 30 O COMANCHE PEAK STEAM ELECTRIC STATION TECHNICAL SPECIFICATION ._ , 3/4.11 O i i

ixx-88512 ArtAcW o i 14 CPSES Technical Specifications PAGE 2 0F 30 NRC Draft 2 Markup Section 3/4.11 IO) U Change 10# Justification For Change 0410 Add the exception to 3.0.3 and 3.0.4. Since there 0419 is no direct correlation between plant power level and liquid effluent concentration, taking exception to 3rovisions of Specifications 3.0.3 and 3.0.4 will probably 1elp to focus attention of shift management on the niore immediate requirement i.e. terminating the release. In fact, by applying Specification 3.0.3 and 3.0.4 may even be detrimental since the shutdown will create more liquid that may need to be released. 0411 This Table is being relocated to the Radioactive Effluent 0412 and Environmental Monitoring Manual. TV Electric believes 0413 the inclusion of this Table is unnecessary and the 0415 information would be more appropriately addressed in the 0420 Radioactive Effluent and Environmental Monitoring Manual. 0421 Relocation of this Table is consistent with the guidance provided in the NRC's Interim Policy Statement (52FR3788), February 6,1987, and the recommendations of the Westinghouse Owners Group MERITS Program. - . This change allows the relocation of tables, while maintaining complicance with 10CFR50.36a. Priority is given to the relocation of this Table since the detailed information is not used by the Licensed Operator. The information currently in this Table is more appropriately maintained in a document subject to TV Electric administrative control and 10CFR50.59 review under the Radioactive Effluent and Environmental Monitoring Manual. This change is similar to that Licensed at Millstone 3. 0417 This Technical Specification is being relocated to the 0418 Radioactive Effluent and Environmental Monitoring Manual. 0428 TV Electric believes the inclusion of this Specification 0431 is unnecessary and the information would be more 0433 appropriately addressed in the Radioactive Effluent and 0434 Environmental Monitoring Manual.

~~~

IXh88512 ATTACWIENT14 CPSES Technical Specifications. PAGE 3 of 30 NRC Oraft 2 Markup Section 3/4.11 Change 10# Justification For Change (cont.) Relocation of this Specification is consistent with the guidance provided in the NRC's Interim Policy Statement (52FR3788), February 6, 1987, and the recommendations of the Westinghouse 0,vners Group MERITS Program. Priority is given to the relocation of this Specification since the detailed information is not used by the Licensed Operator, and requires no immediate action from the Licensed Operator if the Action Statement is applied. The information currently in this Specification is more appropriately maintained in a doucment subject to TV Electric administrative control and 10CFR50.59 review under the Radioactive Effluent and Environmental Monitoring Manual. This change is similar to that Licensed et Millstone 3. 0419 See 10# 0410 0420 See 10# 0411 0421 See 10# 0411

 ,S          0428          See 10# 0417 0430          This change is to remove the word "continuously" from the surveillance requirement. As presently written, this specification is contradicting with Specification 3.3.3.8 and will be a source of confusion for the proper implementation of the Explosive Gas Technical Specifications. With the word "continuously" in Specification 3.11.2.5, "AT ALL TIMES", this Specification alone requires the hydrogen and oxygen monitors on the WASTE GAS H0 LOUP SYSTEM always in service. This however is contradictory to Specification 3.3.3.8 which only requires the hydrogen and oxygen monitors OPERABLE with j

the WASTE GAS HOLOUP SYSTEM in operation. Specification 3.3.3.8 is written correctly. The system l consists of several waste gas holdup tnaks, a l recirculation system containing waste gas compressors and catalytic hydrogen recombiners, and associated piping. l There are several sources of gases that bleed into the l system when in operation, but the most prominent is the bleed off from the Volume Control Tank. (Note: This system services both Units 1 and 2) The explosiva gas monitors are actually part of the recombiner units, located on the inlet or outlet piping. In normal operation, one holdup tank (the "in service" tank) is p recirculated via the compressor and recombiner and gases

 't                        are bled into the suction of the compressor. All other holdup tanks are isolated, either empty or allowing for decay of contained gases.

IIX49512 ATTACMENT14 CPSES Technical Specifications PAGE 4 of 30 NRC Draft 2 Markup Section 3/4.11 () Change !0f Justification For Change

 \.J 0430 (cont.)  Because the monitors are on the recombiner, only the in-service tank is monitored. This is acceptable because once the tank is isolated, there is no source of gas that can enter the pressurized tank to alter the concentrations of either hydrogen or oxygen. Likewise, when there is no tank in service, i.e., the WASTE GAS H0 LOUP SYSTEM is not in operation, the gas concentrations can't change and there is no reason to monitor the concentrations.

Furthermore, it makes no sense to monitor a stagnant portion of the piping, since there is no capability to monitor the tanks directly. Clearly the intent is to monitor (perform surveillance) gas concentrations when gases are being added to the system. Revision to this specification as proposed will clearly require this via the reference to Specification 3.3.3.8.

                       'This change is similar to that Licensed at Farley and Waterford.

0431 See 10# 0417 p 0432 Specification 4.11.2.6 requires that: The quantity of (-) radioactive material contained in each gas storage tank shall be determined to be within the above limits at least once per 24 hours when radioactive materials are being added to the tank. The bases for the limit on total curies contained in each storage tank... provides assurance that in the event of an uncontrolled release of the tank's contents, the resulting whole body exposure to a MEMBER OF THE PUBLIC at the nearest SITE B0UNDARY will not exceed 0.5 rem. This is consistent with Standard Review Plan 11.3, Branch Technical Position ETSB 11-5 " Postulated Radioactive Release Due to a Waste O s System Leak or Failure,: in NUREG-0800, July 1981." Given the above bases, the limiting value for CPSES is 200,000 curies. CPSES FSAR section 11.3.1 gives the design bases of CPSES Waste Gas Treatment System, and states in part that, "The design of the Gaseous Waste Processing System is based on continuous operation of the Nuclear Steam Supply System assuming fission products associated with l's of the core power generation are available for leakage from the fuel into the coolant. This condition is assumed to exist over ( the life of the plant." t

p$# CPSES Technical Specifications NRC Rev. 2 Markup Section 3/4.11 /N Change 10# Justification For Change O 0432 (cont.) CPSES FSAR Table 11.3-3 showt the conservative assumption based on gaseous waste accumulated over the 40 year life of the plant as being 257,855 curies. Section 11.3.2.1.4 further states that "As is seen in the figure, (Figure 11.- " the increase in activity over the plant life is due to .e build up of krypton-85. Other gaseous isotopes reach equilibrium in approximately 30 days." Since krypton-85 accounts for 122,540 curies, this implies that the maximum buildup that could occur, even under the conservative case for any thirty day period would be approximately 134.417 curies (assumes linear accumulation of the krypton-85), or approximately 68% of the DBA based limit in the LC0. Using the more realistic assumptions of FSAR Tablae 11.3-4, the maximum 30 day accumulation would he on the order of 4% of the LC0 limit. This assumes only cne tank is used. Using a 24 hour surveillance requirements for limit that can only be achieved under the worst case assumptions over several months to years is not consistent with the ALARA concept. 0433 See 10# 0417 0434 See 10# 0417 0603 These paragraphs are being relocated to Specification 5.1.3. This change is made to more clearly define the activities within the Exclusion Area Boundary. Since these paragraphs apply to all of 3/4.11.2 this change is the clearest way to dissiminate the information. The alternative is to put these two para each Specification under 3/4.11.2. (graphs in the i.e. all six sections) Bases of O

IXX-88512 ATTACHMENT 14 PAGE 6 Of 30 3/4.11 RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS win G CONCENTRATION LIMITING CONDITION FOR OPERATION 3.11.1.1 Theconcentrationofradioahivematerialreleasedinliquideffluents to UNRESTRICTED AREAS ( ee Figure 5.1-4) shall be limited to the concentrationt specified in 10 CFR 20, Appendix 8, Table II, Column 2 for radi'anuclides other l than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 2 x 10

microcurie /ml total activity.

APPLICABILITY: At all times. ACTION: a.With the concentration of radioactive materh1 released in liquid effluents to UNRESTRICTED AREAS exceeding the above limits, imediately restore the concen-tration to within the above limits. b, Th p :<...; a e V 5p"'f d #"$ 3 c' ' ' * # ' l #' '" # ' 4 # "' ' in't: 0410 SURVEILLANCE REQUIREMENTS O( 4.11.1.1.1 Radioactive liquid wastes shall be sampled 0,1d analyzed according . g tpegampling and analysis program of Table 1.11-1. o.s spei ted in t he. tg : 0411 g 4.11.1.1.2 Theresultsoftheradioactivityanalysesshallbeusedinaccordancel with the methodology and parameters in the 00CM to assure that the concentrations at the point of release are maintained within the limits of Specification 3.11.1.1. 4 ?O< COMANCHE PLAK - UNIT 1 3/4 11-1 ,

                                         - -                ~'

TXX 88512 ATTACHMENT 14 PAGE 7 0F 30 1D h 0612 TABLE 4.11-1 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM LO R LIMIT MINIMUM 0. OETECTION LIQUIO RELEASE SAMPLING ANALYSIS TYPE OF ACTIVITY (LLO)II) TYPE

                         \         FREQUENCY            FREQUENCY                    ANALYSIS             (pCi/ml)

1. Batch Wast' Release P

Each Batch P Each Batch Principal /mma 5x10 7 Tanks I2) Emitters

                                  \                                               I-131[              1x10 8

a. Waste P' s M 01psdivedand 1x10 5 Monitor One' Batch /M etrained Gases Tanks \, /Gama Emitters)

b. Laundry P '

M H-3 Holdup and 1x10 5 Each Batch y Composite (4) - -

                                                    \                            Gross Alpha         1x10 7

c. Waste Water P \ Sr-89, Sr-90 5x10 8

( Holdup Cach Batch Comp ite I4) C Tanks Fe-55 { N 1x10.e

d. Condensate \

P P s Principal Gamma 5x10 7 Polisher Each Batch Each Bateb Backwash Em m ers (3) Recovery

                                                                     \ '
                                                                         \       I-131               1x10 8 Tanks (6)(7)
                                                                           \,

e. Component 'y-3 1x10 5 Cooling Water Drain Tank (7'

2. Continuous W W Princibal Gamma 5x10 7 l Releases (5) Grab Sample Emitters )

I-131

                                                                                               \     1x10 8

a. T bine H-3 1x10 5
edg Sumps: o. 1 & 2 Effluent (6)(7) 1

( l i l COMANCHE PEAR. - UNIT 3 3/4 11-2 * { w

1

.t.

TXX-88512 AllACHMENT-14 IDI 0412 g g PAGE 8 0F 30 TABLE 4.11-1 (Continued) p=p s TABLE NOTATIONS '"I (1 The LLD is defined, for purposes of these specifications, as the smallest conqentration of radioactive material in a sample that will yield /a net count, above system backgrcund, that will be detected with 95% ppobability l with 6nly 5% probability of falsely concluding that a blank observation I represents a "real" signal. / For a particular measurement system, which may include radio hemical separation: /

                                \
                                  \

4.66 s b / 1 LLO = 7 E V 2,22 x 108 Y exp (-Aat)

Where:

                                        \x                              ,/
                                            \'                         /

LLO = the "a priori"\ lower limit of detection (microcurie per unit mass or volume), N\ , j .. , s b = the standard deviatto,n of the background counting rate or of the counting rate of a blank sample as appropriate (counts per minute), ,e ( E = the counting efficiency (counts per disintegration), V = the sample size (units of mass \or volume), 2.22 x 108 = the number of' disintegrations per minute per microcurie, N Y = the fractional rad,ld hemical yield, hn applicable, A = the radioactive decay constant for the particulars radionuclide (sec 1), and ,/ , at le collection and the=time the of elapsed counting/ time (sec).between the midpoint of sa}np\

                                    /                                               \

Typical value's of E, V, Y, and at should be used in the calculation.

ItshouldberebognizedthattheLLDisdefinedasanaprio\((before r thefact)lia/trepresentingthecapabilityofameasurements94temand not as an a costeriori (after the fact) limit for a particular measurement. 7 (2)A batch yelease is the discharge of liquid wastes of a discrete voluge. N Prior t4 sampling for analyses, each batch shall be isolated, and the thorojdhly mixed by : =thod de:cribed " th: 00CM to assure representative sampling. O< / COMMCHE PEAK - UNIT 1 3/4 11-3 -

                             .. . . . . - . . _-         ..---.- .--.       .-           3 IXX 88512 ATTACHMERT 14 -                                                                             3 PAGE 9 0F 30                                                                             ~ ' 0~
                                                                                                     =

j d TABLE 4.11-1 (Continued) W 042 TABLE NOTATIONS (Continued) hk (3)The orincipal gamma emiters for which the LLD specification s s applie/ incibde the following radionuclides: Hn-54, Fe-59, Co-58, Co-60 / Zn-65hMo-99,Cs-134,Cs-137,andCe-141. Ce-144 shall also be/ measured, but with an LLD of 5 x 10.s. Thislistdoesnotmeanthatonipthese nuclides'are to be considered. Other gamma peaks that ere identifiable, together with those of the above nuclides, shall also be analyzed and reported in'the Semiannual Radioactive Effluent Release 8eport pursuant to Specification16.9.1.4 in the format outlined in Regulatory Guide 1.21, Appendix B, Revision 1, June 1974. ' s / (4)Acompositesample\soneinwhichthequantityofIiquidsampledis proportional to the hyantity of liquid waste discriarged and in which the method of sampling employed results in a specimen that is representative oftheliquidsreleasedh /

                                                  's (5)A continuous release is the\dischargeoflfquidwastesofanondiscrete
                                                                          /

volume,e.g.,fromavoiumeoTasystemphathasaninputflowduringthe. continuous release.

d to be sampled and analyzed in (6)Thesewastestreamsshallberequr(r accordance with this table if ei the following conditions exist: (a) Activity is present in th secondary syster -s indicated by either l Steam Generator Blowdown nitors orNsecondo,j sampling and analysis; or (b) Activity was present,in the respective t ks or sumps during the previous four (4) w k If neither of vethe abo exists, situations these tanks and sumI$s need not be performed.

                                                                / ee then s. s     ling and analysis of

( )All ficw from is waste stream shall be diverted to th Wastewater Holdup Tanks when results of sample analyses show radioactivity resent in the wastestreal/atconcentrationsgreaterthanorequaltoth LLO values given in t/is table. Sampling and analysis of the respectiv tanks or l l sumps ar not required when flow is diverted to the Wastewate Holdup Tanks. l l l e0MANCHE PEAK - UNIT 1 3/4 11-4 .

                                   .        --- - ~ -                           .

m-t 512 . ATIACHMENT14 PAGE 10 0F 30 RADIOACTIVE EFFLUENTS ! DOSE LIMITING CONDITION FOR OPERATION 3.11.1.2 The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released, from each unit, to UNRESTRICTED AREAS (see Figure 5.1-4) shall be limited: 1

a. During any calendar quarter to less than or equal to 1.5 mrems to the whole body and to less than or equal to 5 arems to any organ, and

b. During any calendar year to less than or equal to 3 mrems to the whole body and to less than or equal to 10 arems to any organ.

APPLICABILITY: At all times. ACTION:

a. With the calculated dose from the re'le'a's'e of radioactive materiils in liquid effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding O the limit (s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be

(' l takan to assure that subsequent releases will be in compliance with the above limits. This Special Report shall also include: (1) the reruits of radiological analyses of the drinking water source, and (2) the radiological impact on finished drinking water supplies with regard to the requirements of 40 CFR 141, Safe Drinking Water Act.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.1.2 Cumulative dose contributions from liquid effluents for the current calendar quarter and the current calend'" year shall be determined in accordance

 ,      with the methodology and parameters in the ODCM at least once per 31 days.

l l . l i O( \ COMANCHE PEAK - UNIT 1 3/4 11-5 ,

DX-68512 , ATTACHMENT 14 FAGE 11 Of 30 RADI0ACT,IVE EFFLUENTS . Q ID4:0417 .. ( LIQUID RA0 WASTE TREATMENT SYSTEM

         'IMITING CONDITION FOR OPERATION 3.11.1.3 The liquid radwaste treatment system shall be OPERABLE and /appropriate portions of'the system shall be used to reduce releases of radioactivity when the projected doses due to the liqui:! effluent, from each unit, t6 UNRESTRICTED AREAS (see Figure 5.1-4) would exceed 0.06 mrem to the whole body or 0.2 mrem to any organ in'a 31-day period.                                            /                ,
                                    '.                                              /                 \

APPLICABILITY: At,all times. / 7 ACTION: ' f

a. With radioactive liquid waste being discharged without treatment and in excess of the above limits and any portion of the liquid radwaste treatment system not in operation, prepare and submit to the Commis-sfon within 30 days, pursuant to Specification 6.9.2, a Special Report that includes the'following information:

y . .- - .

1. Explanation of why liquid radwaste was being discharged without treatment, identification.of any inoperable equipment or subsystems, and the reason for the inoperability,

( 2. Action (s) taken to est$retheinoperableequipmenttoOPERABLE status, and

3. Summary description of tion (s) caken to prevent a recurrence,

b. The provisions of Specification 3.0.3 and 3.0.4 are not applicable.

MRVEILLANCEREQUIREMENTS' 4.11.1.3.1 Dosesduetoliquidreleasesfrome\chunittoUNRESTRICTEDAREAS shall be projected at least once per 31 days in cordance with the methodology and parameters in.the 00CM when liquid radwaste t atment systems are not being fully utilizad. / 4.11.1.3.2 Th'einstalledliquidradwastetreatmentsgtemshallbe considered OPERA 8LE by meeting Specifications 3.11.1.1 nd 3.11.1.2.

                       /
                  /
                     /

l COMANCHE PEAK - UNIT 1 3/4 11-6 , l

III-88512 AllACHMENT14

PAGE 12 0F W RADI0 ACTIVE EFFLUENTS RMA 'w \ 10 h 008 g{ f LIQUID HOLOUP TANKS *

      '.          x                                                     Ri.0Cb,,4 LIMITING CON 0! TION FOR OPERATION 3.11.1.4 The quantity of radioactive material contained in each unpr ected outdoor tanks.shall be limited to less than or equ61 *.o 10 Curies..e'xcluding tritium and dis olved or entrained noble gases

APPLICABILITY: ktalltimes.

ACTION: N

a. With the qu ritity of radioactive material in any unprotacted outdoor tank ofr/exceedingsthe abovetolimit, adioactive material immediately,s'uspend the tank, all additions within,48 hcurs reduce the tank contents to within the limit, and describe /the events leading to this condition in the next Semiannual Radioactive Effluent Kelease Report, pursuant to Specification 6.9.3/4.

                                                ',               /

b. The provisions of Specifications 3.0/3 and 3.0.4 are nat applicable.-

N /

SURVElliANCE REQUIREMENTS L 4.11.1.4 The quantity of radioactive / material contained in each of-the unprotectedoutdoortanK/shallb/determinedtobewithintheabovelimitby analyzing a representative cample of the tank's ' contents at least once per 7 days when radioactive matericfs are being added to the tank.

                                                                            \
  =
                                                                                    \

7 s

                      /
                 /                                          uv.peelat O
          *Ta'nks included in this specification are thoseVoutdoor tanks that are not
          ,. Surrounded by liners, dikes, or walls capable of holding the tank contents O(        ' and that do not have tank overflows and surrounding area drains connected to the Liquid Radwaste Treatment System.

COMANCHE PEAK - UNIT 1 3/4 11-7 ,

                                     ~                                           '

in-88H2 AliACHMENT 14 PAGE 13 Of 30 - Q RADI0 ACTIVE EFFLUENTS 3/4.11.2 GASEOUS EFFLUENTS DOSE RATE LIMITING CONDITION FOR OPERATION 3.11.2.1 The dose rate due to radioactive materials released in gaseous effluents from the site to areas at and beyond the EXCLUSION AREA BOUNDARY (see figure 5.1-1) shall be limited to the following:

a. For noble gases: Less than or equal to 500 mrems/yr to the whole body and less than or equal to 3000 mrems/yr to the skin, and

b. For Iodine-131, for Iodine-133, for tritium, and for all radio-nuclides in particulate form with half-lives greater than 8 days:

Less than or equal to 1500 arems/yr to any organ. APPLICA81LITY: At all times. ACTION: , , .. ,

a. With the dose rate (s) exceeding the above limits, immediately restore the l release rate to within the above limit (s). 10 1: 0419 b . Tb- protisic o ok S pec'.Cie d~.c % 3.0 3 Q ~5.6 t4 o.re od agl.'ca.de ,

\ { SURVEILLANCE REQUIREMENTS 4.11.2.1.1 The dose rate due to noble gases in gaseous effluents shall be detwrained to be within the above limits in accordance with the methodology and parameters in the 00CM. 4.11.2.1.2 The dose rate due to Iodir.e-131, Iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents shall be determined to be within the above limits in accordance with the methodology and parameters in the 00CM by obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified in T;tle 4.H4. we "TM C M Ni , ID1:0420 1 0< COMANCHE PEAK - UNIT 1 3/4 11-8 ,

v TABLE 4.11-2 8 RADIDA'"IVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM = E; O g A;7

     $             'NN                                 MINIMUM                                          LOWER LIMIT OF             I g, $
     $                              SAMPLING           ANALYSIS                  TYPE OF                DETECTION (LLD)(3)         %5" GASEOUS RELEASE TYPE                                          ACTIVITY ANALYSIS                                       s z                              FREQUENCY          FREQUENCY                                             (pCi/el)'
     ,%      1. Waste Gas Storage P                   P
      .           Tank              Each Tank          Each Tank           Principal Gamma Emitters (2)     1x10'-               .

c Grab Sample / i'i 2. Containment Purge N. P P

or Vent Each'ReleaseII Each Release I3) Principal Gamma Emitters (2),'- 1x10 4

     "                                         le                                                   ,',

Grab Samp \ M H-3 (oxide) ' 1x10 8

3. Plant Vent M I }'I*)'I ) '-

                                                       \ g(3)              Principal Gamma Emitters (2)     1x10 4
                                                            N          H-3 (oxide)     .-               1x10 Y                         l R                              Continuous (6)          y(7)    '

I-131.- 1x10 12 {

Radiofodine -'

g ysorber ,/ Cartinuous(6) y(7) , ' Principal Gaassa Emitters (2) 1x10 81 , Particulate' Sample / Continuous (6) g/ Gross Alpha ' 1x10 81 Com,posite Par-C"3 M iculate Sample Q ~y Continuous (6) Q Sr-89, Sr-90 1x10 18 g* ~ Cos90 site Par- ' - ticulate Sample s

"t3 p

                                                                                                                          &    s
 /                                .                                                   .

TXI-88512 ATTACMENT 14 PAGE !$ OF 30 . O _ TABLE 4.11-2 (Continued) TABLE NOTATIONS RELOCATE M 042) IlkThe LLD is defined, for purposes of these specifications, as the sma'llest concentrationofradioactivematerialinasamplethatwillyielpanet count, above system background, that will be detected with 95% probability with,only 5% probability of falsely concluding that a blank observation represents a "real" signal. / For a pa'rticular measurement system, which may include rad chemical separation:. D LLD = E V 2.22 x 108 Y exp (-Aat) Where: y~ LLD = the "a priori" lower limit of detection (microcurie per unit mass or volume), j/ s b = the standard de iation of thejt'3ckground counting rate 't of , the counting rate of a blank sample as appropriate. (counts per mir.ute) . E = the counting efficiency (counts per disintegration), J{ V = the sample size (units o mass or volume), p 2.22 x 10e = the numberj of dis tegrations per minute per microcurie, Y = the fractional radiochemical feld, when applicable, A = the radioactiv ! decay constant or the particular i'adionuclide (sec 1), and at a the elapsy time between the midp int of sample collection and the time of counting (sec). ' Typics1 valdes of E, V, Y, and at should b usec; in the calculation. It should be pecognized that the LLD is defined an a prior { (before the fact) limit representing the capability of a me surement system and not as an [ posteri,od (after the fact) limit for a rticular measurement.

O(/

 ,/                                                                                     .

COMANCHE PEAK - UNIT 1 3/4 11-10 -

N-68M2 * - -- - - - - - - - - - - - - - - - -- AliACHMEWI14 PAGE 16 OE 30 . TABLE 4.11-2 (Continued) RLOCATE ,/ gg mg

                         \                     TABLE NOTATIONS (Continued)

( )The p ncipal gamma emitters for which the LLD specification applies include the following radionuclides: Kr-87, Kr-88, Xe-133, Xe'133m, Xe-135, and Xe-138 in noble gas releases and Mn-54, Fe-59, Co'-58, Co-60, Zn-65, Mo-99, I-131, Cs-134, Cs-137, Ce-141 and Ce-144 in 16 dine and particulate' releases. This list does not mean that only th'ese nuclides are to be considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and' reported in the Semiannual Radioactive Effluent Release Report pursuvit to Specification 6.9.1.4 in the format outlined in Regulatory Guide l'21, Appendix B, . Revision 1, June 1974. (3) Sampling and analysis shall also be performed ,following shutdown, startup, s or a THERMAL POWER change exceeding 15% of RATED THERMAL POWER within a 1-hour period. This requi,rement does not ap' ply if 1) analysis of primary coolant activity, performed spursuant to Sp4cification A.4.8, shows that the dose equivalent I-131 concentration in the primary coolant has not increased more than a factor of 3, 2) the noble p4s monitor shows that affluent ~

activity has not increased mers than a facEo~r of 3.

(4) Tritium grab samples shall be t e t least once per 24 hours when the s refueling canal is flooded. \ s (5) Tritium grab semples shall be,t ken at\ \ 'least once per 7 days from the ventilation exhaust from the spent fuelyool area, whenover spent fuel is in the spent fuel pool. (6)The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time perfdd covered by each dos'e or dose rate calculation made in accordance with Specifications 3.11.2. 3.11.2.2, and 3.11.2.3. (7) Samples shall be changed at least once per 7 days and analyses shall be completed within,48 hours after changing, or afterhemoval from sampler. Sampling shall also be performed at least once per 24 hours for at least 4 7 days following each shutdown, startup, or THERMAL POWER change exceeding 15% of RATEDjTHERMAL POWER within a 1-hour period and analyses shall be completed within 48 hours of changing. When samples collected for 24 hours are analy This requ, red, does irement the corresponding not apply if: (1)LL0sanalysis may shows be increased by Agfactor that the DOSE of EQUIVA, LINT I-131 concentration in the reactor coolant has notsincreased moref than a factor of 3; and (2) the noble gas moni, tor shows that eff,1uent activity has not increased more than a factor of 3. l /

                 /
               /

O COMANCHE PEAK - UNIT 1 3/4 11-11

r TXX88512 ATTACHttENT14 FAGE 17 Of 37 RADI0 ACTIVE EFFLUENTS DRAiT DOSE - NOBLE GASES LIMITING CON 0! TION FOR OPERATION 3.11.2.2 The air dose due to noble gases released in gaseous effluents, from each unit, to areas at and beyond the EXCLUSION AREA BOUNDARY (see Figure 5.1-1) shall be limited to the following:

a. During any calendar quarter: Less than or equal to 5 mrads for gamma radiation and less than or equal to 10 meads for beta radiation, and

b. During any calendar year: Less than or equal to 10 mrads for gamma radiation and less than or equal to 20 mrads for beta radiation.

APPLICABILITY: At all times. ACTION

a. With the calculated air dose from radioactive noble gases in gaseous--

effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit (s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to O( assure that subsequent releases will be in compliance with the above limits.

b. The provisions of Specifications 3.0.' and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.2 Cumulative dose contributions for the current calendar quarter and current calendar year for noble gases shall be detereined in accordance with the methodology and parameters in the ODCM at least once per 31 days. COMANCHE PEAK - UNIT 1 3/4 11-12 ,

TIX-88512 AllACHf0T14 PAGE 18 0F M , RADI0 ACTIVE EFFLUEdTS DOSE - IODINE-131. IODINE-133, TRITIUM, AND RADI0 ACTIVE MATERIAL IN k PARTICULATE M M LIMITING CON 0! TION FOR OPERATION 3.11.2.3 The cose to a MEMBER 0F THE PUBLIC from Iodine-131, Iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released, from each unit, to areas at and beyond the EXCLUSION AREA BOUNDARY (see Figure 5.1-1) shall be limited to the following:

a. During any calendar quarter: Less than or equal to 7.5 mrems to any organ and,

b. During any calendar year: Lass than or equal to 15 areas to any organ.

APPLICABILITY: At all times.

ACTION: - - *

a. With the calculated dose from the release of Iodine-131 Iodine-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days, in gaseous effluents exceeding any of the above Os limits, preparo and submit to the Commission within 30 days, pursuant

( to Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the 1 Nit (s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits,

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS

  ,       4.11.2.3 Cumulative dose contributions for the current r,a1         Jar quarter and current calendar year for Iodine-131 Iodine-133, tritium and radionuclides in particulate form with half-lives greater than 8 days 55all be determined in accordance with the methodology and parameters in the 00CM at least once per 31 days.

i COMANCHE PEAK - UNIT 1 3/4 11-13 -

IXX 88512 AfiACMENT 14 FASE 19 Of 30 . RAOI0 ACTIVE EFFLUENTS GASEQUS RADWASTE TREATMENT SYSTEM ggg LIMIf!N CON 0! TION FOR OPERATION ,

3.11.2.4 The PRIMARY PLANT VENTILATION SYSTEM and the GASEOUS WASTE PROCESSING SYSTEM shall be OPERABLE and appropriate portions of these systems shall be used to reduce' releases of radicactivity when the projected doses in 31 days due to gaseous effluent releases, from each unit, to areas at and beyond the EXCLUSION AREA 800NOARY (see Figure 5.1-1) would exceed:

a. 0.2 mrad to air from gamma radiation, or

b. 0.4 mrad to air from beta radiation, or

c. 0.3 ares to any o,rgan of a MEMBER OF THE BLIC.

APPLICABILITY: At all times.\ ACTION:

                                                                          . .. .                   -     .       I

a. With radioactive gaseous waste be ng discharged without treatment and in excess of the above111mit's, prepare and submit to the Commission within 30 days, purs,uant to Specification 6.9.2, a SpecialReportthatincludeVthefollowinginformation:

( 1. Identificationofadinoperableequipmentorsubsystems,and the reason for thegi noperability, ;

2. Action (s) taken to restore the inoperable equipment to OPERABLE status, and

3. Summary description of action (s) taken to prevent a recurrence,

b. The provisto of Specifications 3.0.3 and 3.0.4 are not applicable.

l SURVEILtANCE REQUIREMENTS

                                         /                                                  \\

4.11.2.4.1 Doc due to gaseous releases from each unit to a as at and beyond the EXCLtJSION AREA BOUNDARY shall be projected at less pnce per 31 days in accordanWwith the methodology and parameters in the 00CM when Gaseous Radwaste .e'atment Systems are not being fully utilized. 4.11.2.4/2 The installed PRIMARY PLANT VENTILATION SYSTEM and GASEOUS WASTE i PROCESS!NG SYSTEM shall be considered OPERA 8LE by meeting Specificati s 3.11.2'.1 and 3.11.2.2 or 3.11.2.3. O< COMANCHE PEAK - UNIT 1 3/4 11-14 .

TU 88512 - - - - - i ATTACHMENT 14 PAGE20of.30 , O RADI0 ACTIVE EFFLUENTS EXPLOSIVE GAS MIXTURE LIMITING CONDITION FOR V ERATION 3.11.2.5 The concentration of oxygen in the WASTE GAS HOLOUP SYSTEM shall be limited to less than or equal to 3% by volbee whenever the hydrogen concentration exceeds 4% by volume. APPLICABILITY: At all times. ACTION:

a. With too concentration of oxygen in the WASTE GAS HOLOUP SYSTEM greater than 3% by volume but less than or equal to 4% by volume, reduce the oxyg-n concentraticn to the above limits within 48 hours,

b. With the concentration of oxygen in the WASTE GAS HOLDUP !YSTEM greater than 4% by volume and the hydrogen concentration greater th e 4% by volume, ima ?iately suspend all additions of waste gases -

to the system and reduce the concentration of oxygen to less than or equal to 4% by volume, then take ACTION a , above,

c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.5 The concentrations of hydrogen and oxygen in the WASTE GAS HOLOUP SYSTEM shall be determined to be within the above limits by centinueutly I 430 ID monitoring the waste gases in the WASTE GAS HOLOUP SYSTEM with the hydrogen and oxygen monitors required OPERABLE by Table 3.3-13 of Specification 3.3.3.11. , lO l COMANCHE PEAK - UNIT 1 3/4 11-15 ' I

               !!IC512 ATTACHM[Ni14 fAGE21of30        .

RkDI0ACTIVEEFFLUENTS , GAS STORAGE TANKS gg{ [ loi0431' LIMIT!NdsCON0!TIONFOROPERATION / _ 3.11.2.6 The quantity of radioactivity contained in each gas sta ge tank shall be limited to less than or equal to 200,000 Curies of nob)e gases (considered as Xe-133 equivalent). pplICABILITY: 'At all times. ACTION: \

a. With the quantity of radioactive material ip any gas storage tank '

exceeding the'above limit, isusediately suspend all additions of radioactive material to the tank, within contento to within the limit, and descr / De8 hours reduce the events the tank leading to this ' condition in the next Semiannual Radioa)ctive Effluent Release Report, pursuant to Specification 6.9.1.7.

 .                  b. The provisions of Specifications    '.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIPEMENTS \

p 4.11.2.6 The quantity of redioactiv aterial contained in each gas storage 10 40432 l t

( tankshallbedeterminedtobewithj tqe above limit at least once per-24 92 dgs

            % when radicactiva materiais a e bein added to sne tank.

I e l i Oc COMANCHE PEAK - UNIT 1 3/4 11-16 .

l TXX-88512 . . . _ _ _ . - - - _ . - - ATIACHMENT14 i PAGE 22 0F 30, ) a g pu RADIOACTIVE EFFLUENTS U.O .[

           \                                                                                                         /
     ~

3/4.11.3 SOLIO RADI0 ACTIVE WASTES gg{ LIMITING CONDITION FOR OPERATION / W 0433 3.11.3 ' Radioactive wastes shall be solidified or dewatered in accordance~ with the PROCESS CONTROL PROGRAM to meet shipping and transportation requirements during transit, and disposal si.te requirements when received at the d_isposal site. APPLICABILITY: At all times. ACTION:

a. With SOLIDIFICATION or dewatering not meeting disposal site and shipping and transportation requirements, suspend, shipment of the inadequately processed wastes and correct the PROCESS CONTROL PROGRAM, the procedures', and/or the Solid Waste System ,as necessary to prevent recurrence. s

b. With SOLIDIFICATION or dewatering not performed in accordance with the PROCESS CONTROL PROGRAM, test the impr'operly processed waste in each container to ensure that it mes'ts burial ground and shipping .

requirements and take appropriate admij(istrative action to prevent recurrence. /

c. The provisions of Specifications 3 3 and 3.0.4 are not applicatsle.

SURVEILLANCE REQUIREMENTS ( 4.11.3 SOLIDIFICATION of at least one r esentative test specimen ftom at least every tenth batch of each type 9( wet radioactive wastes (e.g., filter ( sludges, spent resins, evaporator bojttoms, boric acid solutions, and sodium sulfate solutions) shall be verified in accordance with the PROCESS CONTROL PROGRAM:

a. If any test specimen ails to verify SOLID

         -                ofthebatchundertstshallbesuspended(FICATION,theSOLIDIFICATION@ til such time av additional test specimens can be obtained, alternative $qLIDIFICATION parameters i                                           CONTROL PROGRAM, can be determine / n accordance with the PROC and a subsequan test verifies SOLIDIFICATION.

the batch mayj en h be resumed using the alternati h SOLIDIFICATION parameters determined by the PROCESS CONTROL PROGRAM

b. If the initial test specimen from a batch of waste fa s to verify SOLIDIFI$dTION, the PROCESS CONTROL PROGRAM shall provide for the l

collect,fon and testing of representative test specimens from each l consep6tive batch of the same type of wet waste until at ihast three l i consecutiveinitialtestspecimensdemonstrateSOLIDIFICATI%asprovided Th/PROCESSCONTROLPROGRAMshallbemodifiedasrequired, i,rt Specification 6.13, to assure SOLIDIFICATION of subsequent b'atches of waste; and

c. With the installed equipment incapable of meeting Specification y i 3.11.3 or declared inoperable, restore the equipment to OPERABLE '

g( j status or provide for contract capability to process wastes as necessary to satisfy all applicable transportation and disposal requirements.

COMANCHE PEAK - UNIT 1 3/4 11-17 .

IXX-88512 AllACHMINI 14 ' PAGE 23 0F 'O ' IDI 0434 s RADIOACTIVE EFFLUENTS g,

       ,       3/4.11.4 TOTAL DOSE ITING CONDITION FOR OPERATION 3.11.        The annual (calendar year) dose or dose commitment to any MEMBER OF THE PU IC due to releases of radioactivity and to radiation from uranium fuel cycle s rces shall be limited to less than or equal to 25 areas to the whole body or a organ, except the thyroid, which shall be limited to less than or                                    ;

equal to 7 areas. ~ APPLICABILITY: At all times. ACTION:

a. With the alculated doses from the release of radioactive materials in liquid r gaseous effluents exceeding twice'the limits of Specifi-cation 3.11. 2a., 3.11.1.2b., 3.11.2.2a., 3c11.2.2b., 3.11.2.3a., or 3.11.2.3b., leulations shall be made including direct radiation contributions *as the units (including outside storage tanks etc.) to determine wheth the above limits of Specification 3.11.4 have been exceeded. If suc is the case, prepare and submit to the Commission '

within 30 days, pu uant to Specific 4fi'o'n 6.9.2, a Special Report that defines the cor active action to be taken to reduce subsequent releases to prevent r urrence of,4xceeding the above limits and includes the schedule r achieving conformance with the above limits. O This Special Report, as fined/in 10 CFR 20.405(c), shall include an analysis that estimates th radiation exposure (dose) to a MEMBER OF THE PUBLIC from uranium fue, cycle sources, including all effluent pathways and direct radiation for the calendar year that includes the release (s) covered by'this toport. It shall also describe levels of radiation and concentrations af radioactive material involved, and the cause of the exposdre levels If the estimated dose (s)exceedstheabovelimits,yconcentrations. and if the release condition result-ing in violation of/40 CFR 190 has n64 already been corrected , the Special Report shall include a request \for a variaace in accordance with the provisions of 40 CFR 390. subhittal of the report is considered a t (nely request, and a varian'ce is granted until staff l action on th request is complete.

b. The provis ns of Specifications 3.0.3 and 3 0.4 are not. applicable.

s l SURVEILLANCE REQVIREMENTS \ 4.11.4.1 C ative dose contributions from liquid and gaseous. effluents s' 01 be de reined in accordance with Specifications 4.11.1.2,'4.11.2.2, and 4.11.2.3, d in accordance with the methodology and parameters in. the ODCM. l 4.11.4. Cumulative dose contributions fror. direct radiation from the units (inclu ng outside storage tanks etc.) shall be determined in accordance with the thodology and parameters in the 00CM. This requirement is applicable on1 under conditions set forth in ACTION a. of Specification 3.11.4. l

         /    COMANCHE PEAK - UNIT 1                      3/4 11-18            .

l

IXX-88512 . ATIACHMENT14 . PAGE 24 0F 30 , su. D) (J 3/4.11 RADIOACTIiE EFFLUENTS BA5E5 , _ _ _ 3/4.11.1 LIQUID EFFLUENTS 3/4.11.1.1 CONCENTRATION This specification is provided to ensure that the concentration of radio-active materials released in liquid waste effluents to UNRESTRICTED AREAS will be less than the concentration levels specified in 10 CFR 20, Appendix B, Table II, Column 2. This limitation provides additional assurance that the levels of radioactive materials in bodies of water in UNRESTRICTED AREAS will result in exposures with'n: (1) the Section II.A design objectives of 10 CFR 50 Appendix I, to o MEMBER OF THE PUBLIC, and (2) the simits of

    . 10 CFR Part 20.106(e) to the population. The concentration limit for dissolved or entrained noble gases is based upon the assumption that Xe-135 is the con-trolling radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2.

This specification applies to the release of radioactive materials in liquid effluents from all units at the site. . - ! 'he required-detection-capabi'itic; fer radie::tiv: aterial; in liqui 4 l

t :: ple: r: t bul:ted in ter ; Of the lower limits of detection (LL0s).

! /3 -Detailed discussion of th: LLO, and Other detecti n 14-its car be found in-- V( Currie, L. ^. , "Lower Limit Of 00tection: 0 fi ition and Elab0rstion of ID I: 0413 n Proposed Pcsit;ca for-Radiological Effluent and Environmental M::weement:," l - N'.l REC /CR-4077 (September 1004), and in the MASL Prc edures Manual, HASL-300 l (revised erne:11y). 3/4.11.1 2 DOSE i l This specification is provided to implement the requirements of Sections II.A, III.A, and IV.A of 10 CFR 50 Appendix I. The Limiting Condition for Operation implements the guides set forth in Section II.A of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure l that the releases of radioactive material in liquid effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." Also, for fresh water sites with drinking water supplies that can be potentially affected by plant l operations, there is reasonable assurance that the operation of the facility will not result in radionuclide concentrations in the finished drinking water l that are in excess of the requirements of 40 CFR 141. The dose calculation i methodology and parameters in the 00CM implement the requirements in Section 1 III. A of Appendix I that conformance with the guides of Appendix I be shown by l calculational procedures based on models and data, such that the actual expo-l sure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. Tt,a equations specified in the 00CM for calcu-l lating the doses due to the actual release rates of radioactive materials in I liquid effluents are consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of COMANCHE PEAK - UNIT 1 B 3/4 11-1 , 1

TXX-88512 AllACHMEHi 14 PAGE 25 0F 30 g RADI0 ACTIVE EFFLUENTS . . . . . V BASES DOSE (Continued) Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.113. "Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementino Appendix I," April 1977. This specification applies to the release of radioactive materials in liquid effluents from each unit at the site. For units with shared Radwaste Systems, the liquid effluents from the shared system are to be proportional among the units sharing that system. _ 3/4.11.1.3 LIQUID RADIOWASTE TREATMENT SYSTEM s\ s The OPERABILITY of the Liquid Radwaste Treatment Systam ensures thjt this syst'em will be available for use whenever liquid effluents require tr,eatment prior to. release to the environment. The requirement that the appyepriate portions'of this system be used when specified provides assurance'that the releases of' radioactive materials in liquid effluents will be,fept "as low as is reasonably' achievable." This specification implements the requirements of 10CFR50.33a,GeneralDesignCriterion60of10CFR50gpendixAandthe design objective given in Section II.0 of 10 CFR'50 endix I. The specified' limits governing the' use of appropriate portions o he Liquid Radwaste Treat-ment System were specifi,ed as a suitable fractio f the dose design objectives set forth in Section II.Axof 10 CFR 50 Append . for liquid effluents. O N Q' This specification appl'i s to the re,le se of radioactive materials in liquid effluents from each un t the p te. For units with shared Radwaste i Systems, the liquid effluents f e' shared system are to be proportioned l among the units sharing that syst q, .. ID 1: 0418 3/4.11.1.4 LIQUID HOLOUP TANX f UNIL rdedCb The tanks listed in tydspecificati include all those outdoor tar. , l both permanent and tempo ary that are not s ounded by liners, dikes, or l walls capable of hold g the tank contents an , hat do not have tank over-flows and surroundi area drains connected to the Liquid Radwaste Treatment System. Restric ng the quaatity of radioactive material tained in the speci-fied tanks rovides assurance that in the event of an un qntrolled release of the tan scontents,theresultingconcentrationswouldbeTessthanthe limi of 10 CFR 20, Appendix B, Table II, Column 2, at the rMarest potable wa,t(r supply and the nearest surface water supply in an UNRESTRLCTED AREA. l

      . f 1

COMANCHE PEAK - UNIT I B 3/4 11-2 ,

   . 1XX-88512 ATTACHMEHf 14 PAGE 26 0F 30      .

Alb[I

 /T      RADIOACTIVE EFFLUENTS                                                                         6 b      BASES 3/4.11.2 GASEOUS EFFLUENTS 3/4.11.2.1 OOSE RATE             p cm This specificatio is provided to ensure that the dose at any time at and beyond the Exclusfon AYea BOUNDARY (EAB) from gaseous effluents from all units on the site will be with'n the annual dose limits of 10 CFR 20 to UNRESTRICTED AREAS. The annual dose limits are the doses associated with the concentrations of 10 CFR 20, Appendix B, Table II, Column I. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of a MEMBER OF THE PUBLIC in an UNRESTRICTED AREA, either within or outside the EAB, to annual average concentrations exceeding the limits specified in Table II of 10 CFR-Peet 20 Appendix B (10 CFR 20.106(b)).

For MEMBERS OF THE PUBLIC who may at times be within the EAB, the occupancy of that MEMBER OF THE PUBLIC will usually be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the EAB. The methodology of calculating doses for such MEMBERS OF THE PUBLIC, shall be given in the 00CM. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to a MEMBER OF THE PUBLIC at or beyond the EAB to less than or equal'to 500 mrems/ year to th'e whole body or to less than or equal to 3000 mrems/ year to the skin. These release rate limits also restrict, at all times, the corresponding thyroid dose rate above background to a child via the inhalation pathway to less than or equal to 1500 mrems/ year. This specification applies to the release of radioactive materials in gaseous effluents from all units at the site. I Activities unrelated to plant operation which may be pe ted within the Exclus1 rea include the exercising of mineral rights the maintenance of pipelines. Applicants will have the necessary c rol to determine these activities and w quire that all persons inv ed in them report to the CPSES Mantger, Plant Op ions or his desi ed representative prior ~to engaging in the activities. 108 0603 l l Publica44ee-recreational a vitie thin the Exclusion Area are limited to Squaw Creek Reservoir an quaw Creek Par . propriate and effective arrangements have been e (in coordination with appropriate agencies) to control access tn, ivities on, and the reN val of pe and property from the reservoir ase of emergency. Arrangements for recrea 1 use and l emergency cedures governing such use have been completed. The icants have t authority to exclude or remove any person from this area at any ti . required detection capabilities for radioactive material in gaseous waste samp abulated in terms of the lower limits of_ detection (LL0s). IDl:0421 Cetailed discussion o and other detection Hmits can be found in Currie, L. A., "Lower Limit of e Proposed Position foryologfifaTEffluent Ar0efinition and Elaboration 7nMuironmental of a ," Measurements NUREG/CR-4,0Z246eptember 1984), and in the HASL ProceduresJianual, HASL-300 (d_) .(revised ~ annually). COMANCHE PEAK - UNIT 1 B 3/4 11-3 ,

l 1 IXX-88512 ' AliACHMENT 14 1 PAGE 27 0F 30

                        ~     '

j RADIOACTIVE EFFLUENTS MkU U K. , 1 BASES DOSE-NOBLE GASES (Continued) 3/4.11.2.2 OOSE - NOBLE GASES This specification is provided to implement the requirements of Sections II.8, III.A and IV.A of 10 CFR.50 Appendix 1. The Limiting Condition for Operatio.1 implements the guides set forth in Section I.8 of Appendix I. The ACTION statements provide the required operating flexibility and at the same l time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in gaseous effluents to UNRESTRICTED AREAS sill be kept "as low as is reasonably achieveble." The Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix ! be shown by calculational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially under-estimated. The dose calculation methodology and parameters established in the 00CM for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents are consistant with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual.. Doses to Man from Routiae . Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR 50, Appendix I, "Revision I, October 1977 and Regulatory Guide 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Efflu-m ents in Routine Releases from Light-Water Cooled Reactors," Revision 1, July 1977. The 00CM equations provided for determining the air aoses at and beyond the EA8 are based upon the historical average att..nspheric conditions. l l This specification applies to the release of radioactive materials in gaseous effluents from each unit at the site. Since both units share the radwaste treatment systems, the gaseous effluents are proportioned among the units. 3/4.11.2.3 OOSE - 100lNE-131, IODINE-133, TRITIUM, AND RADIOACTIVE MATERIAL IN PARTICULATE FORM This specification is provided to iniplement the requirements of Sections II.C, III. A and IV. A of 10 CFR -Pact,- 50 Appendix I. The Limiting Conditions for Operation are the guides set forth in Section II.C of Appendix I. The i ACTION statements provide the required operating flexibility and at the same ! time implement the guides set forth in Section IV.A of Appendix I to assure thdt the releases of radioactive materials in gaseous effluents to UNRESTRICTED l AREAS will be kept "as low as is reasonably achievable." The 00CM calculational methods specified in the Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appe.1 dix I be shown by calculational procedures based on models and data such that the I actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The 00CM calculational methodology i and parameters for calculating the doses due to the actual release rates of the subject materials are consistent with the methodology provided in Pegulatory l q Guide 1.109, "Calculation of Anrual Doses to Man from Routine Releases of i Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine COMANCHE PEAK ' UNIT 1 B 3/4 11-4 , l l

                                                              .                                             l TIX-88512                                                                                          I ATTACHMENT 14                                                                                       j PAGE 28 Of 30 -     -

p'm l N RADI0 ACTIVE EFFLUENTS i J ' BASES 00SE - IODINE-l'1, IODINE-133, TRITIUM, AND RADIOACTIVE MATERIAL IN PARTICULATE FORM (Continued) Releases from Light-Water-Cooled Reactors," Revision 1, July 1977. These equa-tions also provide for determining the actual doses based upon the historical average atmospheric conditions. The release rate specifications for Iodine-131 Iodine-133, tritium, and radionuelidas in particulate form with half-lives greater than 8 days are dependent upon the existing radionuclide pathways to man in the areas at and beyond the EAB. The pathways that were examined in the development of the calculations were: (1) individual inhalation of airborne radionuclides, (2) deposition of racionuclides onto green leafy vegeta- en with subsequent consumption by man, (3) deposition onto grassy areas w' .e milk animals and meat producing animals graze with consumption of the milk and meat by man, and (4) deposition on the ground with subsequent exposure of man. This specification applies to the release of radioactive materials in gaseous effluents from each unit at the site. For units with shared radwaste treatment systems, the gaseous effluents from the shared system are proportioned among the units sharing that system. - - '- - W4.11. 2. 4 GASEOUS RADWASTE TREATMENT S',. M[h ID I: 0428 / Th ERABILITY of the GASEOUS WAST. .NGSYSTEMandthe'IRIMARY

                                                                                       ~
 %      PLANT VENTI          ON SYSTEM ensures that the            .s will be av6Hnble for use i whenever gaseous           fluents require treatmen prior to reJesse to the environ-l       ment. The requireme             that the appropriata portions,or these systems be used, when specified, provides            asonable assurance that'the releases of radioactive materials in gaseous effluen will be keptj" (low as is reasonably achiev-able." This specification imple            ts thpfequirements oi' 10 CFR 50.36a, General Design Criterion 60 of 10 C                   50 Appendix A and the design objec-tives given in Section II.D of J04FR                 50 Appendix 1. The specified limits governing the use of approp,rJete portions o              e systems were specified as a i        suitable fraction of theAfose design objectives                  forth in Sections II.B and l        II.C of 10 CFR +eet- ' Appendix I, for gaseous efflue                 .

This s 1fication applies to the release of radioactive w terials in gaseous j effluents from each unit at the site. For units with shire (radwaste trje teent systems, the gaseous effluents from the shared system are p M or-

       -tToned among the units sharing that system.                                           N   'N l

O l U l COMANCHE PEAK - UNIT 1 8 3/4 11-5 ,

IXX-88512 ATTACHMENT 14 PAGE 29 0F 30 ,

  /^N        RADIOACTIVE EFFLUENTS
       )

BASES N 3/4.11.2.F EXPLOSP!E GAS MIXTURE This specification is provided to ensure that the concentration of poten-tially explosive gas mixtures contained in the WASTE GAS HOLOUP 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 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 hM- 50 Appendix A. N3/4 11.2.6 GAS STORAGE TANKS . ID I: 0431 xThe 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 th.e quantity of radicactiv.ity , contained 'in each gas storage tank provides assurance that in the eveht of an uncontrolled eelease of the tank's contents, the resulting whole body exposure to a MEMBER OF'ItiE PUBLIC at the nearest EAB will not exceed 0.5, rem. This is l consistent with Standard Review Plan 11.3, Branch Technical Posi' tion ETSB 11-5, "Postulated Radicad ve Releases Due to a Waste Gas System Leak or Failure," in NUREG-0800, July 1981. l 3/4.11.3 SOLIO RADI0ACTI E. WASTES E0MT!ioi:o4ss This specification impi s the requirements,of 10 CFR 50.36a and General Design Criterion 60 of 10 CFR 5 A pendix A. The process parameters included in establishing the PROCESS CONTRO PROGRAM may, include, but are not limited

         ,  to, waste type, waste pH, waste / liqui SOLIDIFICATION agent / catalyst ratios, waste oil content, waste principal chemical c'onstituents, and mixing and curing times.

3/4.11.4 TOTAL DOSE . [IDh0434 This specification is provided'to meet the do limitations of 40 CFR 190 1 that have been incorporated into/10 CFR 20 by 46 FR(14525. The specification requires the preparation and sutimittal of a Special Rehcrt whenever the calcu-lated doses due to releases of radioactivity and to radih ion from uranium fuel cycle sources exceed 25 mr, ems to the whole body or any o,rga except the thyroid, which shall be , limited to less than or equal to 75 m ems. For sites l containing up to four reactors, it is highly unlikely that the h sultant dose I tc a MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR 190 if the I individual reactors' remain within twice the dose design objectives of yAppendix I and if direct radf'ation doses from the units (including outside storag tanks, etc. ) are keptjinall. The Special Report will describe a course of acti that i p shouldresult/inthelimitationoftheannualdosetoaMEMBEROFTHEPUBL(to l (') withinthej0CFR190 limits. For the purposes of the Special Report, it maps be assumea that the dose commitment to the MEMBER of the PUBLIC from other \ uraniunvfuel cycle sources is negligible, with the exception that dose contri-butions from other nuclear fuel cycle facilities at the same site or within a s l COMANCHE PEAK - UNIT 1 8 3/4 11-6 ,

IXX-88512 l ATTACHMENT 14 l PAGE 30 0F 30 g3r ! RADIOACTIVE EFFLUENTS y V. . . (d8 BASES N 1 TOTAL DOSE (Continued) radius of 8 km must b k onsidered. If the dose to any MEM estimated to exceed the rl Qrements of 40 CFR 190, thp 4p ,BER ecial Report 4with F THE a PUBLIC i request for a variance (provide the release conditions resulting in violation of 40 CFR 190 have not already bee orrec )f'in accordance with the provi-sions of 40 CFR 190.11 and 10 CFR 20.40)s is considered tc be a timely request and fulfills the requirements of 4 R1 il NRC staff action is completed. The variance only relates to limits of 40 C 90, and does not apply in any way to the other re,q ' ements for dose limitatio 10 CFR 20, as addressed in Specifications f3.1.1 and 3.11.2.1. An individual 1 wot considered a MEMBER OF TH LIC during any period in which he/she is engitJe carrying out an pera ion that is part of the nuclear fuel Cycle. 4 l l l i i COMANCHE PEAK - UNIT 1 8 3/4 11-7 , i [

IXX-88512 1 AffACHiEHi15 I PAGE 1 0F 20 l

O l i COMANCHE PEAK STEAM ELECTRIC STATION TECHNICAL SPECIFICATION - . 3/4.12 lO i l O l \ l l I

TXX-88512 AliACHttENT15 FAGE 2 of 20 CPSES Technical Specificat' ions ("')' U NRC Draft 2 Markup Section 3/4.12 Change ID# Justification For Change 0437 Allows for a timcly verification of sample results, thereby increasing the confidence that positive results are real and reducing submittal of unnecessary results. This change was approved and licensed at Vogtle. 0438 This Technical Specification is being relocated to the 0441 Radioactive Effluent and Environmental Monitoring Manual. 0443 TV Electric believes the inclusion of this Specification is unnecessary and the information would be more appropriately addressed in the Radioactive Effluent and Environmental Monitoring Manual. Relocation of this Specification is consistent with the guidance provided in the NRC's Interim Policy Statement (52FR3788), February 6,1987, and the recommendations of the Westinghouse Owners Group MERITS Program. Priority is given to the relocation of this Specification I p) since the detailed information is not used by the Licensed Operator, and requires no'immediate action from the Licensed Operator if the Action Statement is applied. The information currently in this Specification is more appropriately maintained in a document subject to TV Electric administrative control and 10CFR50.59 review under the Radioactive Effluent and Environmental Monitoring Manual. This change is similar to that Licensed at Millstone 3. l l [ l

l (b'

IXX-88512 ATTACHMENT 15 PAGE 3 Of 20 O u 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM nu - t;; j; ID ?: 0438 LIMITING CONDITION FOR OPERATION

                                                                                                 /
                         \

3.12.1 The\R adiological Environmental Monitoring Program shall be' conducted

as specified'in Table 3.12-1.

APPLICABILITY: 1 all times. ACTION:

a. With the Radiological Environmental Monitori ~g Program not being conducted as specified in Table 3.12-1, pr are and submit to the Commission, in the Annual Radiologic Environmental Operating Report required by Specification 6.9.1. , a description of the reasons for not conducting the program as req red and the plans for preyenting a recurrence. $; m dAt 101:0437

b. With the31evel of raovoa'ctivity a the result of plant effluents in an environmental sampling 'mediu at.a.specified location exceeding -

the reporting levels of Table .12-2 when averaged over any calendar quarter, prepare and submit ,the Commission within 30 days, pursuant to Specification 6.9.2, a ecial Report that identifies the cause(s) l for excecding the limit (s and defines the corrective actions to be ! s taken to reduce radioact< ve efflue'nts so that the potential annual l dose

to a MEMBER OF T PUBLIC is less than the calendar year limits of Specifications 3. .1.2, 3.11.2.2, or 3.11.2.3. When more than l

one of the radionu ides in Table 3.12-2 are detected in the sampling nedium, this repo shall be submitted if: concentr_artion (1) concentration (2) + j reporti. level (1)

reporting level (2) . . 1 1. 0 When radip uclides other than those in Table 3.12-2 are detected and are the t'esult of plant effluents, this report shall be submitted if the po,tential annual dose
to a MEMBER 0/ THE PUBLIC'from all radio-nucijdes is equal to or greater than the calendar year slimits of Specification 3.11.1.2, 3.11.2.2, or 3.11.2.3. This report is not

   .                     r huired if the measured level of radioactivity was not the result f plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental l                       Operating Report required by Specification 6.9.1              .
         *Tfie methodology and parameters used to estimate the potential annual dose to
       / a MEMBER OF THE PUBLIC shall be indicated in this report, kV A c m b e m d o g rc h .% sis 6h Oc e c' g'.M , a byU colc , o r a ~

s qt te uAm % se a.e e v a oc 9a ,7 ge a. w res e s l Q_ o-m - ~ m w e. g.,s 3 aw m a^"1 usaym

a. S c o. s o wA% s e, g, l COMANCHE PEAK - UNIT 1 3/4 12-1 ,

TXX-88512 AllACHMENT15 PAGE 4 0F 20 N RADIOLOGICAL ENVIRONMENTAL MONITORING [ . j N ,

                                                                                                 /    IDI H38 LIMITING CONDITION FOR OPERATION                                                   /

ACTION (Continued)

c. With milk or fresh leafy vegetation samples unava able from one or more of the sample locations required by Table .12-1, identify specific locations for obtaining replacement amples and add them within 30 days to the Radiological Environmental Monitoring Program given in the 00CM. The specific locatiens' from which samples were unavailable may then be deleted from 3h'e monitoring program. Pursuant to Specification 6.14, s submit in thVnext Semiannual Radioactive EffluentReleaseReportdocument on for a change in the ODCM including a revised figure (s) d table for the 00CM reflecting the

     .                   new location (s) with lu por     g information identifying the cause of the unavailability of s       s and justifying the selection of the new location (s) for obt n ng samples,

d. The provisions of ecificati s 3.0.3 and 3.0.4 are not applicable.

                                                                   '                                  ~

SURVEILLANCE REQUIRE. NTS s 4.12.1 The r N logical environmental monitoring sam les shall be collected / pursuant t ble 3.12-1 from the specific locations 'ven in the table and C figure (s the ODCM, and shall be analyzed pursuant to the requirements of I Table F12-1 and the detection capabilities required by Tatle 4.12-1.

              /
         /

O COMANCHE PEAK - UNIT 1 3/4 12-2 ,

O . O O x x

          's N                                          TABLE 3.12-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 5                         NUMBER OF                                                                                539 E                         REPRESENTATIVE                                                                           "@$

g EXPOSURE PATHWAY AMPLES AND g) SAMPLING #D , TYPE AND FREQUENCY

                                                                                                                    %$U g  AN0/0R SAMPLE          S    LE LOCATIONS                      C_0LLECTION FREQUENCY /      OF ANALYSIS          N=

[ 1. Direct Radiation (2) Fortyron\ (ine acnitoring Quarterly. Gamma dose quarterly. H stationsefQerwithtwoor w more dosimete N or with one instrument for sh uring and recording dose rate tinu-ously, placed as foil :

An inner ring of stations, o % n each meteorological sector ,if. thq general area of the EXCijl510M AREk

 ,                        BOUNDARY; 1

An outer ring of tations, one in 5 each meteorol scal sector in O the 6- to a range fros. the site; a m I Thy alance of tt.e static.as to be laced in specia 2nterest areas > such as population centers, nearby residences, schools, and in one Q or two areas to serve as control stations. l

M>

i l . g e Co

O _ O O TABLE 3.12-1 (Continued) . g \ RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM khh N e22 J g NUMBER OF g66 g REPRESENTATIVE g EXPOSURE PA HWAY SAMPLES AND SAMPLING AND T AND FREQUENCY

    ,                                             g)                                                                     '

g AND/OR SAMPLE SAMPLE LOCATIONS COLLECTION FREQUENCY OF ANALYSIS 7 2. Airborne g Radioiodine and acples from five locations Continuous sampler pg r- Radici9 dine Cannister.: q Particulates ation with samplycollec- I-131 analysis weekly. tion weekly, pr'more Three samples (Al--A3) from f r equentlyfif required by close to th three 'ACLUSION c dust loiweng. Particulate Sampler: AREA BOUNDARY cations, in Gross beta radioactivity ) different sectors, of the analysis following highest calculated a al average ground-level D Qi filter change; } and gamma isotopic analysis of composite (by

,   {                         One sample from the      ,,

Im ation) quarterly. vicinityofacomajutty y having the higheyt calcu-A lated annual average ground-level D/Q; n'd One le from a control ' 1peation, as for example 15 to 30 km distant and in the least k prevalent wind direction.(

3. Waterborne Squawk Creek Reservoir (6) Monthly O)

a. Sur ace Gamagisotopic analysis

                  /
                                                                                                                                  ~

monthly Composite for

              /               Lake Granbury                          Monthly composite of        tritium a lysis quarterly.
           '                                                         weekly grab samples
      /-                                                             when Lake Granbury is rec.eiving letdown                                     m fro *m SCR. Otherwise,                 7               rm monthly grab                            Q ms h

C3 saniple. (8) g.,

                                                                                                                        =   ---e
                                                                                                                      .; m E

O O O

                                                                                                                                                                                                                                        ~
                                 \                                                                                                            TABLE 3.12-1 (Continued)                                                                 OW s                                                                                                                                                                              RE7 gx
                                                              ,                                                                    RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM o eo 2ME
                                                                                                                                                                                                                                      -~

I m h NUMBER OF UG REPRESENTATIVE i " EXPOSURE PATHWAY SAMPLES AND SAMPLING AND ' TYPE f.ND FREQUENCY h AND/OR SAMPLE SAMPLE LOCATIONS (I) COLLECTION FREQUENCY OP' ANALYSIS

3. Waterborne (Continued) >

                                                                                                                       ~. Control-Brazos River                                   Monthly y                                                             upstream of Lake Granbury

b. Grou:.d soimip1 s from one or two sources Quarterly. Gamma isotopic (5) and only if likely to be affected.( tritium analysis quarterly.

c. Drinking One sample of each'of one to Gr sample at least I-131 analysis on each three of the nearest water nce per 2-week period grab sample when the dose q water supplies that could b,e when I-131 analysis is calculated for the con-

affected by its discharge. performed; monthly grab sumption of the water s'

g sample otherwise. is greater than 1 arem 4 One sample from a cont ol 1 c W on. per year (10). Composite for gross beta and gamma isotopic analyses (b)

                                                         '                                                                                                                                               monthly. Composite for tritium analysis quarterly.

d. Sediment ne sample from downstream area Semiannually. Gamma isotopic analysis Ib) from with existing or potential semiannually.

, Shoreli recreational value. ll rv, i e

= _r;~~>

h s 2 M

o _ o o TABLE 3.12-1 (Continued) 2 E; C "E2 O RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM f\ E omn z '5" o ', NUMBER OF E E y EXPOSURE PATHWAY REPRESENTATIVE SAMPLES AND gy) SAMPLING AND [ TYPE AND F3EQUENCY i g AND/0R SAMPLE SAMPLE LOCATIGNS COLLECTION FREQUENCY OF ANAEYSIS [ 4. Ingestion

E Camma isotopic (S} and

   -4                                      a. Milk                                  Samples from available milking         Semimonthly when 9                                                                                 animals in three locations             animals are on pasture;7   I-131 analysis semi-within 5 km distance having the        monthly at other times. monthly when animals highest dose potential. If there                                  are on pasture; monthly are none',-then one sample from                                   at other times.

available milking animals in each of three areas bqtween 5 to 8 km distant where dosesare \ calculated to be greater than l em per A yr.( 0) One sample fromInilkings g animals at a control locajjfIQS

to 30 km distant and in'the leas prevalent wind direc/ tion.(3) .

                                                                                                      /                                                Gamma isotopic analysis (S)

b. Fish and One sample of/each commercially Samp' semiannually .

 .                                                 Inverte-                          and recrearionally important                                      on edible portions.

brates speci,esdn vicinity of plant d harge area.

                                                                                  / One sample of same species in
                                                                                /    areas not influenced by plant j                            discharge.

c. One sample of each principal At time of harvest (11). Caquna isotopic analyses (S)

Food

                                                 ~ Products                           class of food products from                                      on    'ble portion.

any area that is irrigated by water in which liquid plant wastes have been i

discharged. . a

                                                                                                                                                                . :'8
                                                                                                                                                                          =  -H rN"1 '

b.

O -

O O TABLE 3.12-1 (Continued) ;==

                                                                                                                                                                            ,,               e--

O 'x I RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM / g m NUMBER OF

                                                                                                                                         /
                                                                                                                                           /                                                 $E" oa REPRESENTATIVE A

s EXPOSURE PATlWAY SAMPLES ANO SAMPLING AND

                                                                                                                                      /                                                   '

g3) TYPE AND FREQUENCY R AN0/OR SAM?tE SAMPLE LOCATIONS COLLECTION FREQUENCY OF ANALYSIS h 4. Ingestion (Continued '

c. Food Samp'l of three different Monthly duri Gamma isotopic (N and I-131 N Products kinds broad leaf vegeta- growing s ason, analysis.

(Continued) tion grown arest each of two different fsite loca-tions of highest predicted annual average groundslevel 0/Q if milk sampling is h t performed. w

               }                                                          One sample of each of he               Mqnthly during        Gamma isotopic (5) and I-131 g                                                          similar broad leaf       geta-         growisig season.      analysis.

y Lion grown 15 t0730 km dis-w tant in the st prevalent wind dir ion ( if milk sampli is not performed. - m - C"3 O C3

Ill5 >

lr:= M N E W

TXX-88512 '

.             ATTACMENT 15 PAGE 10 0F 20     .

O f /

         \                                 TABLE 3.12-1 (Continuad),
           \

i TABLE NOTATIONS (1) Specific parameters of distance and direction sector from the cent'erline of one reactor, and additional description where pertinent, shall be pro-s vided for each and every sample location in Table 3.12-1 in a,t'able and figure (s) in the ODCM. Refer to NUREG-0133, "Preparation of/ Radiological Effluent Technical Specificatians for Nuclear Power Plants / October 1978, and to radiological Assessmert Branch Technical Position / Revision 1, November 1979. Deviations are permitted from the requir'ed sampling schedule if specimens are unobtainable due to circumstances such as hazardous conditions, seasonal unavailability, and p$1 function of auto-matic sampling equipment. If specimens are unobtainable due to samling equipment malfunction, ef fort shall be made to coinplete corrective action prior to the end of the next sampling period. fall deviations from the sampling schedule shall be documented in the mental Operating Report pursuant to Specific'a, tionAnnual Radiological 6.9.1.3. Environ-It is recog-nized that, at times,\it mey not be possible or practicable to continue to obtain samples of the media of choice'at the most desired location or time. In these instances suitable alte'rnative media and locations may be.. chosen fer the particular\ pathway injuestion and appropriate substitutions made within 30 days in the'qadiological Environmental Monitoring Program given in the ODCH. Pursuant A Semiannual Radioactive Efflu\to Specification en Release 6.14, submit infor Repcrt documentation thea next change V in the ODCH including a revis igure(s) and table for the 00CM reflect-ing the new location (s) witty supporting information identifying the cause ( of the unavailability of satnples for the pathway and justifying the selec-tion of the new location (4) for obt'hining samples. (2) One or more instrumen, such as a pr surized ion chamber, for measuring and recording dose fate continuously ma be used in place of, or in addi-tion to, integrat ng dosimeters. For the purposes of this table, a thermoluminesce dosimeter (TLD) is considtred to be one phosphor; two or more phosphp s in a packet are considered \as two or more dosimeters. Filmbadges$hallnotbe'sedasdosimetersfkmeasuringdirectradiation. u (The 40 stat' ions is not an absolute number. Thg number of direct radiation monitoring' stations may be reduced according to geographical limitations; e.g. a)/an ocean site, some sectors will be over kater so that the number of d nfmeters may be reduced accordingly. The freq ncy of analysis or readsut for TLD systems will depend upon the charact gistics of the speci-fi.C' system used and should be selected to obtain optim dose information within minimal fading.) (3) Thepurposeofthissampleistoobtainbackgroundinformat(on. If it is not practical to establish control locations in accordance w(th the

        ,/
          /       distance and wind direction criteria, other sites that providh valid background data may be substituted.

f (4) Airborne particulate sample filters shall be analyzed for gross beta radioactivity 24 hours or more after sampling to allow for radon and s C%) thoron daughter decay. If gross beta activity in air particulate samples is greater than 10 times the yearly mean of control samples, gamma N isotcpic analysis shall be performed on the individual samples. \

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

IXX-88512 ATTACHMENT 15 PAGE 11 0F 20

                      ~    ~

,_ x E00Ali TABLE 3.12-1 (Continued) TABLE NOTATIONS (Continued) ' l IDI0438 (5) G1mma isotopic analysis means the identification and quantification of gaga-emitting radionuclides that may be attributable to the effluents from the facility. (6) Squaw eek Reservoir is a closed cooling water basin which receives plant ef ents at the circulating water discharge. The reservoir shall be sample in an area at or beyond but-m+4 near the mixidg zone. Also the reservo of the disch(a shall

e. be sampled at a distance beyond sig'nificant influence (7) Squaw Creek Rese voir is a closed cooling water b* in which is composited naturally.

(8) Lake Granbury may reteive x letdown from Squaw eek Reservoir to control buildup of solids. Th(s is the only pathway fcr plant effluents to Lake Granbury. The lake shay be sampled near J. e letdown discharge and at a distance beyond significa t influence of he, discharge. ,_ , (9) Groundwater samples shall be g taken whe this source is tapped for drinking or irrigation purposes in are(s wher the hydraulic gradient or recharge properties are suitable for coh ami ation. (10) The dose shall be calculated for ye maximum organ and age group, using the methodology and parameters n the 00CM. N (11) If harvest occurs more than nce a yeaq, sampling shall be performed during each discrete harvep . If harvekt occurs continuously, sampling shall be monthly. Attent,fon shall be pa to including samples of tuberous and root food oducts. 4 O COMANCHE PEAK - UNIT 1 3/4 12-9 ,

O

O - O ' TABLE 3.12-2 ;=g R C *? O REPORTING LEVELS FOR RADI0 ACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMI'LES gob h x - %5U g x REPORTING LEVELS ,- ga h \ , N \ WATER AIRBORNE PARTICULATE FISH MILK F000 PRODUCTS

ANALYSIS (pCi/l) OR GASES (pCi/m3 ) (pCi/kg, wet) (pCi/I) (pCi/kg, wet) ,

E Z H-3 20,000* w Mn-54 1,000 30,000 Fe-59 400 10,000 Co-58 1,000 30,p00

q Co-60 300 / 10,000 s g Zn-65 300

                                                                                                                            /      20,000 Z r-Nb-95                                                                400 I-131                                                                     2                      0.9                                             3              100 Cs-134                                                                    30               10                          1,000                    60            1,000 Cs-137                                                                    50               20                         2,000                     70            2,000

3 300 "

Ba-La-140 200 . cm

              *For drinking water samples.                                                         This is 40 CFR Part 141 value. If no drinking water pathway exists, a value           Q-of 30,000 pCi/l may be used.                                                                                                                                                 y
                                   -                                                                                                                                      N i
                                                                                                                                                                                  .=
                                                                                                                                                                                         =
                                                                                                                                                                                         ?

CD

O - O . O - TABLE 4.12-1 hhh DETECTION CAPA8ILITIES FOR ENVIRONMENTAL SAMPLE ANALYSISII} (2) m h LOWER LIMIT Of DETECTION (LLD)(3) m N WATER \ R80f!NE PARTICULATE FISH MILK F000 PRODUCT 5 SE0lMENT

ANALYSIS (pCi/1) GASES (pCi/m3 ) (pCi/kg, wet) (pci/1) (pci/kg, wet) (pci/kg, dry) <

E - U Gross Beta 4 .01 w H-3 2000* /

Mn-54 15 130 , [ Fe-59 30 260 ! y Co-58,60 15 130 g Zn-65 30 26 Z r-Nb-95 15 I-131 1** 0.07 1 60 Cs-134 15 0.05 130 15 60 150

c3 Cs-137 18 C. 150 18 80 180 $

CD Ba-La-140 15 15 "If no drinking water pathway exists, a value of 3000 pCi/l may be used.

                                                                                       /                                                                              \                                 ~

If no drinking water thway exists, a value of 15 pCi/l n.ay be used. N c:= i g

                                                                                                                                                                                   ?.:n A                                                          .__
                                                                                                                                                                                                                                  ~2 M

n- , i' TXX-88512 - ATTACHMENT 15 PAGE 14 of 20

                                                                                               /
                                                                                             /

TABLE 4.12-1 (Continued) / TABLE NOTATIONS (1)This list de'es not mean that only these nuclides are to be considered. Other peaks that are identifiable, together with those of t6e above nuclides, shall.also be analyzed and reported in the Annual Radiological Environmental Operating Report pursuant to Specificatio /6.9.1.3. (2) Required detection' capabilities for thermoluminescent/ dosimeters used for environmental measurements shall be in accordance with the recommenda-

           . tions of Regulatory Guide 4.13.

(3)The LLD is defined, for, purposes of these specifications, f as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be/ detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a "real" signal. , For a particular measurement system, w6ich may include radiochemical separation: LLD = 4.66 'sb / - - - - E V - 2.22 - Yq - exp(-Aat) Where: '\ Os LLO = the "a priori" lower limit of detection (picoCuries per unit 8 mass or volume) \ s b

                        = the standardyeviationofthebachgrouadcountingrateorofthe counting r te of a blank sample as , appropriate (counts per minute),

E = the cou ing efficiency (counts per d sintegration), V = the mple size (units of mass or volum\ e', 2.22 = t number of disintegrations per minute p r picocurie, Y - the fractional radiochemical yield, when appt cable, A = the radioactive decay constant for the particul r radionuclide (sec 1), and at = the elapsed time between environmental collection, or end of the sample collection period, and time of counting ( c). Typical values of E, V, Y, and at should be used in the calculation.

b v COMANCHE PEAK - UNIT 1 3/4 12-12 .

1 TXX-08512 ATTACHMENT IS

                             "" " !"                                                                                                                                    l RE.0WE        P'M

TABLE 4.12-1 (Continued) O ' TABLE NOTATIONS (Continued) , IDI )438 i It should be recognize s an s priori (before the fact) limit representing (hat the LLO is define tM capability offmeasurement, system and not as an a posteriori (after the fat ) limit f a particular measurement. Analysesshallbeperformedin(sbch manner that the stated LL0s will be achieved under routine condition . Qccasionally background fluctuations, unavoidable small sample si [z, the pestence of interfering nuclides, or other uncontrollable cir pcstances may rbnder these LL0s unachievable. Insucncases,thecettibutingfactorsshaKbeidentifiedanddescribed in tne Annual Rad ogical Environmental Operat- g Report pursuant to Specification . . 1. 3. O, O e O COMANCHE PEAK - UNIT 1 3/4 12-13

                  . - . _       ..,_..,....m._._..         ..,._..,_m_.     . , _ _ _ - - . _ , . , , . . _ ,          . _ . _ , .

TXX-88512

 ,       ATTACHMENT 15                                                                     ,

PAGt 16 0F 20 o O RADIOLOGICAL ENVIRONMENTAL MONITORING DMFT 3 . 12.2 LANC USE CENSUS IDI 0441 LIMNIJNGCONDITIONFOROPERATION '

3.12.2 Land Use Census shall be conducted and shall identify within a distance 8 km (5 miles) the location in each of the 16 meteorological sectorsof(thenearestmilkanimal,thenearestresidence,andthenearest garden" of g ater than 50 m2 (500 ft 2) producing broad leaf vegetation. APPLICABILITY: At all times. ACTION: N

a. WithaLankUseCensusidentifyingalocation(s) that yields a calculated dhpe or dose commitment greater than the values currently being calculatgd in Specification 4.11.2.3, pursuant to Specifica-tion 6.9.1.4, identify the new location (s) in the next Semiannual Radioactive Effl nt Release Report.

b. With a Land Use Cen us identifying a' location (s) that yields a calculated dose or do'Se commitment (via the same exposae pathway) 20% greater than at a Iqcation from which samples are currently p being obtained in accord ce with Specification 3.12.1, add the new

( location (s) within 30 day to the Radiological Environmental Moni-toring Program given in the 00CM. The sampling location (s), exclud-(. ing the control station locat on, having the lowest calculated dose or dose commitment (s), via the same exposure pathway, may be deleted from this monitoring program af r October 31 of the year in which this Land Use Census was conducte . Pursuant to Specification 6.14, submit in the next Semiannual Radi qctive Effluent Release Report documentation for a change in the 00C0 including a revised figure (s) and table (s) for the 00CM reflecting tqe new location (s) with informa-tion supporting the change in sampling Tocations.

c. The provisions of Specifications 3.0.3 an 3.0.4 are not applicable.

Broad leaf vegetation sampling of at least three different inds of vegetation may be performed at the EXCLUSION AREA BOUNDARY in each of th different direction sectors with the highest predicted D/Qs in lieu of the garden census.

Specifications for broad leaf vegetation sampling in Table 3.12a , Part 4.c., shall be followed, including analysis of control samples. O N \ COMANCHE PEAK - UNIT 1 3/4 12-14 ,

TXX-88512

  .         ATTACHMENT 15 PAGE 17 0F 20 O

y/ RADIOLOGICAL M RONMENTAL MONITORING N / 108 0441 SURVEILLANCE REQUIREM'EDTS. f

                                          \                 /                                       !

4.12.2 The Land Use Census shall'qonduped during the growing season at l least once per 12 months using that in mation that will provide the best results, such as by a door-to-doo rvey, rial survey, or by consulting local agriculture authoaities. e results o Land Use Census shall be included in the Annual Ra gical Environmental ating Report pursuant to Specification 6.9.1.# ! (9

O ( i 4 O COMANCHE PEAK - UNIT 1 3/4 12-15 ,

TXX-88512 N AliACHMENT 15 PAGE 18 0F 20 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4112.3 INTERLABORATORY COMPARISON PROGRAM g LIMITfNGCONDITIONFOROPERATION /

3.12.3 Analyses shall be performed on all radioactive materials, supplied as part of an Interlaboratory Comparison Program that has been approved by the Commission, that correspond to samples required by Tcble 3 12-1. APPLICABILITY: 'At all times. ACTION: -

a. With analyses not being performed as rc'uired above, report the corrective actions taken to prevent a/ recurrence to the Commission in the Annual' Radiological Environmental Operating Report pursuant to Specificatio'n,6.9.1.3.

N /

b. The provisions of Specificatio i 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS s / -

4.12.3 The Interlaboratory Compar son Program shall be described in the ODCM.

(' A summary of the results obtiin,ed as'part of the above required Interlaboratory Comparison Program shall be included iq the Annual Radiological Environmental V) ( Operating Report pursuant tcv' Specification 6.9.1.# 6 N N O COMANCHE PEAK - UNIT 1 3/4 12-16 ,

l 1XX-88512 ATTACHMENT 15 PAGE 19 0F 20 --

                             ~    ~

it kcj}ll D h 008 / ! 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONI*CR1,Nf _ y BASES 3/4. 12.1 MONITORING PROGRAM The Radiological Environmental Monitort c Pn m o , recuired by this specification provides representative measuremen' Jf radiation and of radio-active materials in those exposure pathways and 1.,e those radionuclide /that lead to the highest potential radiation exposure of MEMBERS OF THE PUBl.IC resulting from the plaat operation. This monitoring program implemedts 5ection IV.B.2.of Appendix I to 10 CFR Part 50 and thereby supplements the Radiological Ef fluent Monitoring Program by verifying that the measurable concentrations of< radioactive materials ano levels of radiatiory'are not higher than expecteo on the ba:,is of the effluent measurements and tbt modeling of the environmental exposure pathways. Guidance for this monit'oring program is provided by the Radidlogical Assessment Branch Technical Position on Environ-mental Monitoring, Rev'ision 1, November 1979. The initially specified monitoring program will be effecti'e v for at least the first 3 years'of commercial operation. Following this period, program changes may be initiated based on operational experience. \ The required detectioncapabilities \ / for environmental sample analyses-are . tabulated in terms of the lower \ limitsofdetect/on(LLDs). The LLDs required by Table 4.12-1 are considered optimum for routine environmental measurements in industrial laborahries. It s'hould be rec 4gnized that the LLD is defined (] as sn a priori (before the fact) limit representing the capability of a measure-(f ment system and not as an a posteridri (af t'er the fact) limit for a particular i measurement. K Detailed discussion of the LLD, nn Anther detection limits, can be found in Currie, L. A. , "Lower Limit of Deiiection: Definition and Elaboration of a Proposed Position for Radiological / Effluent and Environmental Measurements," NUREG/CR-4007 (September "$L Procedures Manual, HASL-300 (revised annually). 1984)/,andir.the ~~ l 10 h 0441 3/4.12.2 LAND USE CENSUS - This speci/ication i provided to ensure that changes in the us2 of areas at and beyond the EXCLUSf0N AREA BOUNDARY are identi'fied and that modifica-tionstotheRadiologjdalEnvironmentalMor.itoringProgrsmaremadeifrequired l bytheresultsoft)hscensus. The bcst information from the door-to-door survey,fromaeriaJ,surveyorfromconsultingwithlocalNgriculturalauthori-l l ties shall be used. This census satisfies the requirement 4 of Section IV.B.3 of 10 CFR-Pect- 56 Appendix 1. Restricting the census to ga'fdens of greater than :50 m2 prov' ides assurar.athat significant exposure path 4 4 ys via leafy vegetables w M1 be identified ano monitored since a garden of'this size is the i minimum required to produce the quantity (26 kg/yer ) of leafy ' vegetables assumed 'fRegulatory Guide 1.109 for consumptio# by a child. Tdsdetermine this minfrum garden size, the following assumptions were made: (1 20% of the gardenf 4as used for growing broad leaf vegetation (i.e., similar to Jettuce and cabbage), and (2) a vegetation yield of 2 kg/m2 , l l3 l J r COMANCHE PEAK - UNIT 1 B 3/4 12-1

10-88512 , AliACMEhi 15 PAGE 20 Of 20 / sRADIOLOGICAL ENVIRONMENTAL MONITORING . iL 10 8: 0443 / BASklx ,. 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM The requirement for participation in an proved Interlaboratory Comparison Program is provided to ensurg that indep nt checks on the precision and accuracy of the measurements o radiop tve materials in environmental sample matrices are performed as part o e quality assurance program for environmental monitoring in order to demonstr e th the results are valid for the purposes of Section IV.8.2 of 10 CFR 50 Appen I. ( O CCMANCHE /EAK - UNIT 1 B 3/4 12-2

III-88512 Ai!ACHMENT16 PAGE 1 Of 12 j COMANCHE PEAK STEAM ELECTRIC STATION TECHNICAL SPECIFICATION 5.O O { l O l l

Trx 88512 AfiACHMENT 16

      ' PME 2 0F 12 CPSES Technical Specifications

() NRC Draft 2 Markup Section 5.0 Chance 108 Justification For Chance 0603 This change'is made to relocate these paragraphs from the Bases for 3/4.11.2.1 to Specification 5.1.3. The relocation of these paragraphs is made to more clearly define the activities within the Exclusion Area Boundary. Since these paragraphs apply to all Specifications under 3/4.11.2 this change is the clearest way to dissiminate the information. The alternative is to put these paragraphs in the Bases of each specification under 3/4.11.2. O 4 } a i L .

I in-88512 AliACHMENT 16 ' PAGE 3 OF 12 l DRE (' v

5. 0 DESIGN FEATURES 5.1 SITE EXCLUSION AREA b .1.1 The Exclusion Area shall be as shown in Figure 5.1-1.

LOW POPULATION ZONE 5.1. 2 The Low Population Zone shall be as shown in Figure 5.1-2. MAP DEFINING UNRESTRICTED AREAS AND EXCLUSION AREA BOUNDARY FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS 5.1.3 Information regardi~.y radioactive gaseous and liquid effluents, which will allow identification of structures and release points as well as defini-tion of UNRESTRICTED AREAS within the EXCLUSION AREA BOUNDARY that are acces-sibletoMEMBERSOFTHEPUBLIC,shallbeasshowninFigure/5.1- .nu 5.1-4. The definition of UNRESTRICTED AREA used in implementing these Tectinical Specifications has been expanded over that in 10 CFR 20.3(a)(17). The UNRESTRICTED AREA boundary may coincide with thr EXCLUSION AREA BOUNDARY',' as

defined in 10 CFR 100.3(a), but the UNRESTRICTED AREA does not include areas over water bodies. The concept of UNRESTRICTED AREAS, established at or beyond the EXCLUSION AREA BOUNDARY, is utilized in the Limiting Conditions for Operation to keep levels of radioactive materials in liquid and gaseous effluents as low &s is reasonably achievable, pursuant to 10 CFR 50.36a.

      "IA)S6RT 4 -3P

5. 2 CONTAINMENT I 0603 l

CONFIGURATION 5.2.1 The containment building is a steel-lined, reinforced concrete building of cylindrical shape, with a dome roof and having the following design , features:

a. Nnminal inside diameter = 135 feet. g

b. Nominal inside height = 192.5 feet. (Dope 67.5 feet; total =

260 feet)

c. Nominal thickness of concrete walls = 4.5 feet,

d. Nominal thickness of concrete roof = 2.5 feet.

e. Nominal thickness of conc mat = 12.0 feet.

f. Nominal thickness of steel liner wall = 3/8 inch. (Dome = 1/2 inch, Base Mat = 1/4 inch), and

g. Hetfreevolume=[2,985,000)cubicfeet.

DESIGN PRESSURE AND TEMPERATURE 5.2.2 The containment building is designed and shall be maintained for a p maximum internal pressure of 50 psig and a temperature of 280 F. U COMANCHE PEAK - UNIT 1 5-1 .

      -All&CNMENI16 PAGE 4 0F 12 g .)                                                   INSERT A

Activities unrelated to )lant operation which may be permitted within the Exclusion Area include tie exercising of mineral rights and the maintenance of pipelines. The Applicants will have the necessary control to determine these activities and will require that all persons involved in them report to the CPSES Manager, Plant Operations or his designated representative prior to-engaging in thase activities.

Public recreational activities within the Exclusion Area are limited to Squaw Creek Reservoir and Squaw' Creek Park. Appropriate and ef fective arrangements have been made (in coordination with the appropriate agencies) to control access to, activities on, and the removal of persons and property from the reservoir in case of emergency. Arrangements for recreational use and emergency procedures governing such use have been completed. The Applicants have the authority to exclude or remove any person from this area at any time. l l

t III-88512 ATTACHMEHi16 PAGE 5 Of 12 MT ' O I.ta 56GLT h

~~

l i l l FIGURE 5.'.-1 1 EXCLUSION AREA COMANCHE PEAK - UNIT 1 5-2 ,

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   , c,vn --, w, -    . ,..                                  _.                                                 - - - - - - - - - - -                    - - - - -

IXX-08512 i AliACHMENT16 PAGE 10 Of 12 g j' i

                                                                                           . . .sa   o DESIGN FEATURES O            5. 3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The core shall contain 193 fuel assemblies with each fuel assembly containing 264 fuel rods clad with Zircaloy-4 except that limited substitution of fuel rods by filler rods (consisting of Zircalny-4 or stainless steel) or by vacancies may be made if justified by a cycle specific reload analysis. Each fuel rod shall have a nominal active fuel length of 144 inches. The initial core loading shall have a maximum enrichment not to exceed 3.15 weight percent U-235.       Reload fuel shall be similar in physical design t thq initial core loadingandshallhaveamaximumenrichmentnottoexceed3.5jweightpercent U-235.

CONTROL R00 ASSEMBLIES 5.3.2 The core shall contain 53 full-length control rod assemblies. The full-length control rod assemblies shall contain a nominal 142 inches of absorber material. The nominal values of absorber material shall be 95.5% hafnium with the remainder zirconium. All control rods shall be clad with stainless steel tubing. , , , ., ,

5. 4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The Reactor Coolant System is designed and shall be maintained:

I

a. InaccordancewiththeCoderequirementsspecifiedinSection.(5.2/

of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements,

b. For a pressure of 2,485 psig, and

c. For a temperature of 650'F, except for the pressurizer which is 680*F.

VOLUME 5.4.2 The total water and steam volume of the Reactor Coolant System is {12,500];100 cubic feet at a nominal T 3yg of 589.5'F.

5. 5 METEOROLOGICAL TOWER LOCATION 5.5.1 The primary meteorological tower shall be located as shown on Figure 5.1-1.

m COMANCHE PEAK - UNIT 1 5-5 drv

fu 88512 ATTACHMENI 16 PAGE 11 Of 12 W e .nl 3 DESIGN FEATURES 0 V 5.6 FUEL STORAGE CRITICALITY

5. 6.1.1 The spent fuel storage racks are designed and shall be maintained with:

a. A k,ff equivalent to less than or equal to 0.95 when flooded with unborated water, which includes a conservative allowance for uncertainties as described in Section 4.3 of the FSAR, and b.

A nominal 16 inch center-to-center distance between fuel assemblies placed in the storage racks.

5. F .1. 2 The k,ff for new fuel for the first core loading stored dry in the spent fuel storage racks shall not exceed 0.98 when aqueous foam moderation is assumed.

ORAINAGE

5. 6. 2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 854 feet.

CAPACITY 5.6.3 The two spent fuel storage pools are designed and shall be maintained with a storage capacity limited to no more than 1116 fuel assemblies. 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT 5.7.1 The components identified in Table 5.7-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.7-1. O COMANCHE PEAK - UNIT 1 5-6 ,

O .

                             .                      O                                                      O             -

TABLE 5.7-1 a y COMPONENT CYCLIC OR TRANSIENT LINITS z N 3 '*2 2 Gk3 CYCLIC OR DESIGN CYCLE

                                                                                                                    ===
                                                                                                                  , = =

A COMPONENT TRANSIENT LIMIT OR TRANSIENT A

  . Reactor Coolant System   200 heatup cycles at < 100*F/h            Heatup cycle - T     from < 200*F          '

c and 200 cooldown cycles at to > 550*F. ""9 5

 *                           < 100*F/h.
                             -                                         Cooldown cycle - T      from y                                                                       550*F to _< 200*F**9 200 pressurizer cooldown cycles           Pressurizer cooldown cycle at < 200*F/h.                             temperatures from > 650*F to
                                                                      < 200*F.

80 loss of load cycles, without > 15% of RATED THERMAL POWER to immediate Turbine or Reactor trip. 0% of RATED THERMAL POWER. o, 40 cycles of loss-of-offsite Loss-of-offsite A.C. electrical A.C. electrical power. ESF Electrical System. 80 cycles of loss of flow in onej Loss of only one reactor reactor coolant loop. . coolant pump. 400 Reactor trip cycles. 100% to 0% of RATED THERMAL POWER. 10 auxiliary spray Spray water temperature differential actuation cycles. > 320*F, but i 625'F. 200 leak tests. Pressurized to > 2485 psig. 10 hydrostatic pressure tests. Pressurized to > 3107 psig. Secondary Coolant System I steam line break. Break in a > 6-inch steam lire. i M 10 hydrostatic pressure tests. Pressurized to > 1481 psig. ,'.-1l3

2 e

                    - ' ' _            -               ~'    -'   -

TH-88M2 ATIACHMENT17 PAGE 1 Of 41 O COMANCHE PEAK STEAM ELECTRIC STATION TECHNICALSPECIFICA,TI,0]! 6.0 O O

IXX-88512 ATTACHMENT 17

'   PAGE 1 0F 41
                                                                 \

O i l l COMANCHE PEAX STEAM ELECTRIC STATION TECHNICALSPECIFICATI0],4 6.0 O O l l

TXX-88512 AliAC M NT 17 W2#H CPSES Technical Specifications NRC Oraft 2 Markup Section 6 r\ V Change IDf Justification For Change 0518 This change adds the ability to allow for personnel to be task qualified. This is based on the fact that training can be accomplished on a specific task that does not require an overall indepth knowledge of a position as defined in ANSI-N18.1-1971. This allows for a person to be used as a working member of the staff after completing task qunlification and continue to work towards position qualification. This change is similar to that licensed at Vogtle. 0520 This change identifies the 50RC composition for CPSES. This change is made for two basic reasons. The first reason is to eliminate the possibility of having to make a Tech Spec change due to a manageinent reorganization that eliminates, adds, combines or just renames organizational positions. The second reason is to allow the flexibility for the Vice President, Nuclear Operation to designate the most qualified personnel to represent the required areas on SORC. This will result in a higher technical level of competence and maximize consistency. 0522 Change the wording to ensure that there is no confusion as to which administration procedures have to be reviewed by [>} x- SORC. This is based on the SORC will review all administrative procedures recommended by Appendix A of Regulatory Guide 1.33, Revision 2, February 1978. These are procedures which disseminate to all departments within Nuclear Operations how business will be conducted and guidance in writing intradepartmental procedures. O O

in 88512 ATTACHMENT 17 FAE 3 # 'l CPSES Technical Specifications NPC Draft 2 Markup Section 6 (Cont'd) O Change 10# Justification For Change 0523 Add review of changes to the Radioactive Effluent and Environmental Monitoring Manual (REElH) to S0RC responsibilities. This is based on, the requirements of Specification 3/4.11 (in part) and 3/4.12 are being relocated to the REEMM for implementation control. This requirement is an extension of the new Specification 6.15. 0524 Added a new requirement for 50RC to review the Technical Specification Improvement Program (TSIP) and several editorial corrections as a result. This is based on the requirements of several Specifications and parts of Specifications are being relocated to the TSIP for implementation and control. This ensures'that any changes made to the TSIP receive the highest level of review consistent with other documents of the same importance. 0525 Changed the approval authority for changes to the Security Plan and Emergency Plan, as well as changes to the implementing procedures, to the Vice President, Nuclear Operations. This is consistent with Specification 6.5.3.la. OS27 Change so that S0RC makes_ recommendations to the (] 't 0529 designated line manager as defined in 6.5.3. This 0913 is explicitly required since the Vice President, Nuclear Operations has delegated the approval authority for departmental procedures to the responsible line manager. For item 6.5.1.6b, which controls the review of safety evaluations on all procedures, 50RC will be making recommendationsTo the responsible line manager for approval or disapproval. O

IIX-C'512 ATTACMDif 17 PAE 4 of 41 . , CPSES Technical Specifications NRC Oraft 2 Markup Section 6 (Cont'd) Change 10f Justificat_ ion For Change 0531 Added the requirements to perform an audit of the Radioactive Effluent and Environmental Monitoring Manual i (REEM) . The REEM incorporates the Radiological Environmental Monitoring Program, therefore due to the i. increase in.the material to be audited the frequency has been increased to 24 months. This also makes it consistent with tne 00CM. This change is similur to that licensed at Millstone 3. 0534 Added the requirement to perform an audit of the Technical Specification Improvement Program (TSIP). This is to ensure that all requirements located to the TSIP are implemented in a correct manner. 0536 Added the Station Operations Review Committee for 0914 distribution of Licensee Event Reports and made minor editorial changes to more closely agree with the standard technical specifications. O 1 ( I O 1 i

1xxse512 . AtlACM DI11 PAE 6 0F 41 CPSES Technical Sp;cifications

                   . -             NRC Draft 2 Markup Section 6 (Cont'd)

Change 10# Justification For Change 0537 Added to the list of Procedures and Programs are the 0538 Radioactive Effluent and Environmontal Monitoring 0540 Manual, Fire Protection Program ani Technical Specification improvement Program. This will ensure procedures are written to implement these programs and changes to these procedures are reviewed per Specification 6.5. 0541 Replace these sections with_a section which references the 0542 previous Specification 6.5. Specification 6.5 has the same requirement ~as the replaced section with the exception that all the procedures of Specification 6.8.1 are not approved by the Vice President, Nuclear Operations as discussed for Specification 6.5.3.1. 0547 Added the description of the Radioactive Effluent and Environmental Monitoring Manual to provide basic guidance, controls and limitations for sampling and analysis for requireraenta of the Technical Specifications that are being relocated to the Radioactive Effluent and Environmental Monitoring Manual. This change is similar 4 to that licensed at Millstone 3.

O

ixx-C512 AliAC W NI 17 PAGE 6 of 41 CPSES Technical Specifications NRC Draft 2 Markup Section 6 (Cont'd) Change 10# J_ustification For Change 0550 The reporting requirements are being maintained within the Technical Specification for the Annual Radiological Environmental Operating Report and the Semiannual Radioactive Effluent Release Report. The specific details of what information must be included in these reports is being relocated to the Radioactive Effluent and Environmental Monitoring Manual. This change is similar to that licensed at Millstone 3 and maintain the requirements of 10CFR50.36a. 0556 The note requiring the use of specific WCAPs for the analytical methods used to generate the Fsub (xy) limit has been deleted. The sentence is very specific in the fact that whatever method is used, it must be previously reviewed and approved by the NRC. This change may prevent an unnecessary change in the future. 0557 Added the CPSES Technical Specification Improvement Program and Radioactive Effluent and Environmental Monitoring Manual to ensure that the records produced from the implementation of these documents are retained for a finite period of time. V 0559 Change the reference from the snubber Technical Specification to the CPSES Technical Specification Improvement Program. This is based on the snubber requirements being shifted to ASME Section XI and implemented through the CPSES Technical Specification Improvement Program. 0562 Add a new section to set the administrative controls under which the Radioactive Effluent and Environmental Monitoring Manual is initially approved and how all changes initiated by the Licensee are reviewed and approved. O

IIK'.512 ATTACHMENT 17 CPSES Technical Specifications NRC Oraft 2 Markup Section 6 (Cont'd) Change 10# Justification For Change 0907 These changes remove the requirement to maintain an 0916 organizational chart in the Tech Specs. The information that is being removed from the Tech Specs is contained in the FSAR and updated in accordance with 10CFR50.71. This change is made in conformance with Generic Letter 88-06. 0908 NPC initiated these changes during the 4/12/88 meeting 0921 with OSP and NRR. 0909 This note is for, and included on the previous page, therefore should be deleted. 0911 This change deletes the word "may." With the word "may" included it would require that 50RC review all proposed procedures, changes to procedures, equipment, systems, facilities, proposed tests or experiments. This would be so time consuming that it would significantly detract from the real responsibility of 50RC to be cognizant of actual safety concerns. 0912 Deleted the requirement for 50RC to recommend in writing a3 proval or disapproval of 6.5.1.6i through 6.5.2.61. T11s is based on the Standard Technical Specifications only require that responsibilities from 6.5.1.6a through 6.5.1.6e have this specific requirement. Each item (i) throu and (gh (k) have

1) already separate specify whererequirements, to forward the 6.13 through 6.16, required reports.

O

TIX 0 512 ATTACHMENT 17 PAE 8 0F 41

                     . .      CPSES Technical Specifications NRC Draft 2 Markup i-Section6(Cont'd) l Change 10#         Justification For Change 0913         See ID# 0527 0914         See 10# 0536 0916         See 10# 0907 0917'        The change to training requirements is made based on the revision to 10CFR55 and NUREG-1262. The reference to Appendix "A" to 10CFR55 has been deleted since it no longer exists. The reference to the H.R. Denton letter dated March 28, 1980 is removed due to the guidance put out in NUREG-1262. The rule supersedes and includes the requirements of the Harold Denton letter of March 28, 1980 (Q/A number 385 of NVREG-1262). The only requirement to be satisfied is to meet the minimum requirement of the revised rule.

0920 Change the quorum requirement to 5 members and the chairman vice 4 members and the chairman. This is based on having the majority of the 50RC areas represented in every 50RC meeting. (O/ 0921 See ID# 0908

TIX 88512 . AITACMENT !? PAGE90F41 U ...b.

. e O

SECTION 6.0 ADMINISTRATIVE CONTROLS O O . I i i 1

AIIACHMENT 17 PAGE 10 0F 41 AOMINISTRATIVE CONTROLS 6.1 RESPONSI5!LITY 6.1.1 The Vice President, Nuclear Operations shall be responsible for overall operation of the site, while the Manager, Plant Operations shall be responsible for operation of the unit. The Vice President, Nuclear Operations and Manager, Plant Operations shall each delegate in writing the succession to this respon-sibility during their absence. 6.1. 2 The Shift Supervisor (or during his absence from the control roca, a designated individual) shall be responsible for the control room command function. A management directive to this effect, signed by the Vice President, Nuclear Operations shall be reissued to all station personnel on an annual basis. 6.2 ORGANIZATION

      ^" M T!

b ['5.052;_ ii 355- b 3N. 52 ;-es *

                                                                                            ~
                                         . , . . , .   .           = ~ ~ ; ;" '~' 0 ' ' , ,

UNIT STAFF ID 1: 0907 6.2.2 The unit organization shall bes n4 n;7 tt. eg';5.

                                                                       '-     M.c; 5.g: :c. .

a. Each on-duty shif t shall be composed of at least the minimum shif t crew composition shown in Table 6.2-1; N

i

b. At least one licensed Operator shall be in the control room when fusi is in the reactor. In addition, while the unit is in MODE 1, 2, 3, or 4, at least one licensed Senior Operator shall be in the control room;

c. A Radiation Protection Technician

shall be on sits when fuel is in the reactor;

d. All CORE ALTERATIONS shall be observed and directly supervised by either a ifcensed Senior Operator or licensed Senior Operator Limited to Fuel Handling who has no other coricurrent 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 meneers of the minimum shift crew necessary for safe shutdown of the unit and any personnel required for other essential functions during a fire emergency;.4me- )

l

     *The Radiation Protection Technician and Fire Brigade composition may be less                            )

than the minimum requirements for a period of time not to exceed 2 hours, in l order to accosmodate unexpected absence, provided immediate action is taken ' to fill the required positions.

O ' CCMANCHE PEAK - UNIT 1 6-1

I IIX48512 AliACHMENT U .

i PA K 11 0F 41 ! l O INSERT FOR PAGE 6-1 SECTION 6.2.1 H H 6.2.1 Onsite and Offsite Organization An onsite and an offsite organization shall be established for unit operation and corporate management, respectively. The onsite and offsite organization shall include the positions for activities affecting the safety of the nuclear power plant.

a. Lines of authority, responsibility and communication shall be established and defined from the highest management levels through intermediate _ levels to and including; all operating organization positions. Those relationships shall be documented and updated, as appropriate, in the form of organizational T charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in the equivalent forms of documentation. These requirements shall be documented in the FSAR.

b. The Vice President, Nuclear Operations shall

 ;                          be responsible for overall site safe operation and shall have control over those onsite activities necessary for safe operation and maintenance of the plant.

c. The Executive Vice President, Nuclear Engineering and Operations shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety.

d. The individuals who train the operating staff and those who carry out the radiation protection and quality assurance functions may report to the appropriate manager onsite; I

O' however, they shall have sufficient organizational freedom to ensure their independence from operating pressures. l

AllACHMENT17 ' PAGE 12 0F 41 ADMINISTRATIVE CONTROLS W " ' '

O UNIT STAFF (Continued)

f. Administr'ative 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 p l

    '              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. 8212)) ctad,

g. TA *. Shld Opte b s #4 9 e skll hold o. s am;or-P W 4or opera.hr- lictw se .

10 8: 0916 O O COMANCHE PEAK - UNIT 1 6-2

III-88512 , ATTAC W Ni 17 PAGE 13 0F 41 IDI0907 hi L G TE 1 ..

s. .. ..

O O e i i- , i l FIGURE 6.2-1 0FFSITE ORGANIZAf!0N O l COM MCHE PEAE - UNIT 1 6-3 ' t

   . TIr88512                                                            .                            !

AllAC W WT 17 ! PAE 14 W 41 l

[ 1 1000907 l t k i P h I 3

i

                                                          ' ' ^ '

E G T'E I I

I O ! L i i i ( r i L f

I FIGURE 6.2 2

UNIT ORGANIZATION ! i COMANCHE PEAK UNIT 1 6-4 . } .. 's

IXX-88512 ATTACMENT 17 PAGE 15 0F 41 TABLE 6.2-1 MINIMUM SHIFT CREW COMP 0SITION SINGLE UNIT FACII,ITY POSITION NUMBER 07 IN0!VIOUALS REQUIRED TO FILL POSITION MODE 1, 2, 3, or 4 MODE 5 or 6 SS T 1 SRO 1 None RO 2 1 A0 2 1 STA 1* None SS Shift Supervisor with a Senior Operator ifcense on Unit 1 SRO - Individual with a Senior Operator license on Unit 1 R0 - Individual with an Operator license on Unit 1 AO - Auxiliary Operator STA - Shift Technical Advisor l The shift crew composition aay be one less than the minimum requiremertts of; Table 6.2-1 for a period of time not to eice ed 2 hours in order to accommodate unexpected ansance of on-duty shift crew meebers provided immediate action is taken to restore the shift crew ccaposition to within th'a minimus requirements of Table 6.2-1. This provision does not permit any shift crew position to be S uncanned upon shift change due to an oncooing shift .rewman being late or absent. i l During any absence of the Shift Supervisor from the control room while the unit ! is in MODE 1, 2, 3, or 4, an individual with a valid Senior Operator license shall be designated to assume the centrol rc,oe command function. During any absence of the Shift Supervisor from the control room while the unit is in MODE 5 or 6, an individual with a valid Senior Operator license or Operator license shall be designated to ensume the control room command function. l 4 1 "The STA position shall be manned in MODES 1, 2, 3, and 4 unless the Shift Supervisor or the individual with a Seninr Operator license meets the qualifications for the STA as required by the NRC. O COMANCHE PEAK - UNIT 1 6-k 3 . l

l TXX-88512 { ATTACHFEWT 17 PAGE 16 0F 41 i _ . r

i j ADMINISTRATIVE CONTROLS -

4 i 6.2.3 INDEPENDENT SAFETY ENGINEERING GROUP (ISEG) j FUNCTION ! 6.2.3.1 The ISEG shall function to examine unit operating characteristics, j NRC issuances, industry advisories, Licensee Event Reports, and other sources i of unit design and operating experience information, including units of similar design, wMch may indicate areas for improving unit safety. The ISEG shall make detailed recommendations for revised procedures, equipment modifi- . cations, maintenance activities, operations activities, or other mesns of j improving unit safety to the Vice President, Nuclear Operations. l COMPOSITION l 6.2.3.2 The ISEG shall be composed of at least five, dedicated, full-time

engineers located on site. Each shall have a bachelor's degree in engineering

! or related sciones and at least 3 years professional level experience in his i I field. j RESPONSIBILITIES ! 6.2.3.3 The ISEG shall be responsible for maintaining surveillance of unit

activities to provide independent verification." that these activities.are .

j performed correctly and that human errors are reduced as much as practical. RECOR05 1 " 6.2.3.4 Records of activities performed by the ISEG shall be prepared, main-tained, and forwarded each calendar sonth to Vice President, Nuclear Operations. 6.2.4 SHIFT TECHNICAL ADVISOR 6.2.4.1 The Shift Technical Advisor (STA) shall provide advisory technical support to the Shift Supervisce in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe op9 ration 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 minimue quali-fications of ANSI-N18.1-1971 for comparable positions, except for the Radia-tion 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 E qualifications of the supplemental requirements specified in Sect'.ons A and C T. of Enclosure 1 of the March 28, 1980 NRC letter to all licensera.(Pe for to .

Met.G y & goalMted %s .9 Ansi st t-a7f, pers mI my ks perW Hel fo P" h w k 4 t 4 ses.' Ate. AastsWe a;A p t:f h hew W W de%fe.)

     "Not responsible for sign-off function.
   **Until the Radiatica Protection Manager meets all qualification per R.G.1.8, September 1975, an individual who meets all those qualifications shall support the Radiation Protection Manager.

COMANCHE PEAX - UNIT 1 6-D

-.=

TIX-88512 ATTACHMENT 17 . PAGE 17 0F 41 == ll L ADMINISTRATIVE CONTROLS UNIT STAFF QUALIFICATIONS (Continued) 6.4 TRAINING 6.4.1 A retraining and replacement training program for the unit staff shall be maintained under the direction of the Vice President, Nuclear Operations and shall meet or exceed the requirements and recommendations of ANSI-N18.1-1971

. : .".;;.nd'; ^ :' 10 CFR 55 :nd th: ::;; ::;nt;! r:;;ir;::nt: :;::i'i:d '- 3 0;;ti:n; ^, :nd 0 0' in;!;;;r: 1 ;' th " r;h 20, 1^00 '"O 1.tt.r te .li

                                 ;;n;;;;, and shall include familiarization with relevant industry opera-tional experience.

ID 1: 917 6.5 REVIEW AND AUDIT 6.5.1 STATION OPERATIONS REVIEW COMMITTEE (50RC) FUNCTION 6.5.1.1 The 50RC shall function to advise the Vice President, Nuclear Operations on all matters related to nuclear safety. COMPO5ITION

                               .......a
                               , e ,

v.. u.

                                                                               , m a,

_m_,, <_ _______ _,

                                                                                                                                         ,....m             .. m.m..-

ID I: 0520 Ch^ir;;n; "';; "r;;id:nt ";;l;;r 0;;r-tien: ,

                                                         "; ::r;                                                          [C;;r:t'::: !;;;rvi:;r:                                                                              "::t:.- t; b; l    'tgEST " A .,                                        "::::r;                                                          [T:0h-ic:? !;; rv'::r]                                                                               :quiv-1.nt te.
                                                         ";:::r;                                                          [" '-t:::n:: !;;;rv'::r:                                                                             th;;; ;;;i-
                                                         ";;t : r -

[Pler.t In:tr; :.0 :nd 0:ntrel On-in;;r] tion , t.; Neeb r- [FI;rt ";;l;;r Cagineer] e' hivh l

                                                         ".;;er.                                                          [";eith."hy^ici;t]

aenes;;_;_nt y.2....... ALTERNATES

6. 5.1. 3 All alternate members shall be appointed in writing by the 4446- Ulce, fres'. dent kanan to serve on a temporary basis; however, no mare than two alternates l A)u.clearogem% hall participate as voting members in SORC activities at any one time.

l l l

                              ..o_.,,                                                                                 __ __.,__              u______
                                         ......                 m_...,a,.21.._,__
                                                                                                    . .__,. . . .>_s>
                                                                                                                    . . . . . >,. ._,. . . . . x_. . . .-__    . ._

_.... 2.,.

                                                                                                                                                                                                   ,..12.,._._.2.._._.

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                                                                         ,a,,,

c --...... w.,.. %. ,-u w i . ,- . . . . .

                                         ;;;;;rt th; ":di ^ti ^n ."r-t;;ti:n ";n;;;r.

ID 1: 0909 COMANCHE PEAK - UNIT 1 6-\ E~ . l

      -._--s..   ---,. _ ,.__  . _ _ . . . _ . . , - _ ,                   . _ , . _ _ _ , , , . _ , _ _ , . . , _                  _.,__m      _ _ - , - .               - ,. , ,,-,-

IXF88512 AliACHMENT17 PAGE 18 W 41 () l INSERT FOR NEW PAGE 6-5 (OLD 6-7) SECTION 6.5.1.2 , eeAei 6.5.1.2 The SORC shall be composed cf managers or individuals reporting directly to managers from the areas listed below and meet the requirements of ANSI N18.1-1971 Section 4.2.4 for required experience. Operations Maintenance Instrumentation and controls Technical Support Radiation Protection Quality Assurance Emergency Planning Security . . . . - .. . Testing The Manager, Plant operations shall serve as the chairman of

  /~'  SORC.      A senior health physicist is acceptable for the

( Radiation Protection representative on SORC. The SORC members shal). be designated, in writing, by the Vice President, Nuclear Operations. O

TXX 88512 ATTACHMENT 17 , PAGE 19 0F 41 l ADMINISTRATIVE CONTROLS O MEETING FREQUENCY

6. 5.1. 4 The SORC shall meet at least once per calendar month and as convened by the 50RC Chairman or his designated alternate.

QUORUM

6. 5.1. 5 The quorum of the 50RC necessary for the performance of the 50RC responsibility and authority provisions of these Technical Specifications ID 1: 0920 shall consist of the Chairman or his designated alternate and 4+we members including alternates. g ;y, (,)

RESPONSIBILITIES I4 PE'C^ U d5 pru.Jwns tscoa+.ded f N 6.5.1.6 The 50RC shall be responsible for: p Appdr A of 4*Sd *4 * "7 F hA 1.is, ow a, Few

a. Review of :P St:tf r ^.tf- S tr:t h: r:::

a i r::; n ,73 ID 1: 0522

b. Review of the safety evaluations fon (1) procedures, (2) change" to procedures, equipment, systems or facilities, and (3) tests or experiments completed under the provision of 10 CFR 50.59 to verify that such actions did not constitute an unreviewed safety question; O

y/ c. Review of proposed procedures and changes to procedures, equipment, systems or facilities which mey involv: an unreviewed safety ques- { tion as defined in 10 CFR 50.59 or involves a change in Technical Specifications; 10 : 0911

d. Review of proposed test or experiments which eey involve an l unreviewed safety question as defined in 10 CFR 50.59 or requires a change in Technical Specifications;

e. Review of proposed ch4nges to Technical Specifications or the Operating License;

f. Investigation of all violations of the Technical Specifications including the forwarding of reports covering evaluation and recom-mandations to prevent recurrence to the Vice President, Nucitar Operations and to the ORC;

g. Review of reports of operating abnormalities, deviations from ex-pected performance of plant equipment and of unanticipated defici-encies in the design or operation of structures, systems or components that affect nuclear safety;

h. Review of all REPORTABLE EVENTS;

i. Review of the Security Plan and shall submit recommended changes to the ORC; O

COMANCHE PEAK - UNIT 1 6-Q '

i AllACHMEWI17 PACE 20 0F 41 ADMINISTRATIVE CONTROLS RESPONSIBILITIES (Continued)

j. Review of the Emergency Plan and shall submit recommended changes to the ORC;

                                                                               ,,              _ , ,,o % ,g         i k.

Review of changes to the PROCESS CONTROL PROGRAM,gne 0FFSITE DOSE CALCULATION MANUAL, and Radwasta Treatment Systems; l ID : 0523 l

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 operations to detect potential hazards to nuclear safety; e l

n. Investigations or analysis of special subjects as requested b the j Chairman of the ORC or the Vice President, Nuclear Operatio ; l Review of the Fire Protection Program and revisions thereto; a J 6.5.1.7 The 50RC shall: ID h 0913 10 h 0521 dsti g d \;n g a. p. r-4 man (s** Sed k b ' 5' 4 l a. Recommend in writing to the Vf:: "r::ftat, ';;?::r ^;;r:ti::: approval

! or disapproval of itees considered under Specification 6.5.1.6a. through og ?, j, t, : d ' d:::, prior to their implementation; l I O b. Render determinations in writing with regard to whether or not each 10 h 0912 l ! ites considered under Specification 6.5.1.6a. through n. and m. constitutes an unreviewed safety question; and

c. Provide written notificatinn within 24 hours to the Executive Vice President-Nuclear Engineering and Operations and the Operations Review Committee of disagreement between the 50RC and the V+ee-

                        " ::fint, ";:!:n ^^: :ti:::; however, the Vice President, Nuclear
                      ,Opei.. lions shall have responsibility for resolution of such dis-agreements pursuant to Specification 6.1.1.

RECORDS - E' N A mMar t;ne-10 h 0529 i 6.5.1.4 The 50RC shall maintain written minutes of each 50RC meeting that, I at a minious, document the results of all 50RC activities performed under the i responsibility provisions of these Technical Specifications. Copies shall be i provided to the Vice President-Nuclear Operations and the Operations Review Committee.

p. St. Visa o 6 h e. T'se.blu.) Spe.c.;fic,a b y 4 , w g p ,

Aa* d

                                8 VI5.'eu n a ceto 10 h 0524 COMANCHE PEAK - UNIT 1                      6-k7           ,

l 1 M

TXX-88512 - AllACHMElli 17 PAGE 21 Of 41 b ADMINI$TRAT % CONTROLS O ., 6.5.2 OPERATIONS REVIEW COM4!TTEE (ORC) FUNCTION 6.5.2.1 The ORC shall function to provide independent review and audit of designated activities in the areas of:

a. Nuclear power plant operations,

b. Nuclear engineering,

c. Chemistry and radiochemistry,

d. Metallurgy,

e. Instrumentation and control,

f. Radiological safety,

g. Mechanical and electrical enginu ring, and

h. Quality assurance practices.

The ORC shall repcet to and advise the Executive Vice President, Nuclear . Enginaring and Ope'ations on those areas 'of responsibility specified in Specifications 6.5.2.7 and 6.5.2.8. I C0,y0SITION l 6.5.2.2 The ORC shall be composed of at least five individuals of whom no more than minority ars members having line responsibility for operations at CPSES. The Chairman and all members will be appointed by the Executive Vice i President, Nuclear Engineering and Operations. t

     +                                                                                   10 I: 0921

{ The ORC members shall hold a Bachelor's d6 gree in an engineering or physical science field or equivalent experience and a sinfeum of 5 years technical experience. It is the responsibility of the Chairman to ensure experience and competence is available to review problems in areas listed in Speci fication 6.5.2.14. through h. To a large esasure, this experience and competence rests with the membership of the ORC. In specialized areas, this experience may be provided by personnel who act as consultants to the ORC. ALTERNAfts 1 6.5.2.3 The Alterante for the Chairman and all alternate members shall be appointed in writing by the Executive Vice President, Nuclear Engineering and l Operations to serve on a temporary basis; however, no more than two alter-l nates shall participate as voting members in ORC activities at any one time.

The Otc C.ks'er m sk ll held a. 6a. hsler 's ci t y<t b% eMt* c int l or phy SIta l S c i %c.t hie.ld o r e p 4' ve.l ht- expteeb et A s d a. minimam OS (* yes re t e c.ks l e.s.\ msss yr;s.I expe.es% cs. l COMANCHE PEAK - UNIT 1 6-h 1 { l l

      -        in-08512                                                 -

ATTACHMENT 17 PAGE 22 0F 41 ADMINISTRA1ILCONTROLS O CONSULTANTS 6.5.2.4 to provideConsultants shall be utilized as determined by the Chairman, ORC expert advice to the ORC. MEETIM FREQUENCY 6.5.2.5 The ORC shall meet at least once per calendar quarter during the initial year of unit operation following fuel loading and at least once per 6 months thereafter. QUORUM 6.5.2.6 The quorum of the ORC n3cessary for the performance of the GRC review and audit functions of these Technical Specifications shall consist of not less than a majority of the appointed individuals (or their alternates) and the Chairman or his dasignated alternate. No more tha? a minority of the l quorum shall have line responsibility for operation of the unit. REVIEW 6.5.2.7 The ORC shall be responsible for'th'e' review of:

a. The safety evaluations for: (1) changes to procedures, equipment, or systems; and (2) tests or experi:sonte comploied under the provision O of 10 CFR 50.59, to verify that such sctions did not constitute an unreviewed safety question;

b. Proposed changes to procedures, equipment, or systees which involve an unreviewed safety question as defined in 10 CFR 50.59;

c. Proposed t sts or experiments which involve an unreviewed safety questior, as defined in 10 CFR 50.59; *

d. Proposed changes to Technical Specifications or this Operating License;

        .           e. Violations of Codes, regulations, orders, Technical Specifications, license requirements, or of internal procedures or instructions having nuclear safety significance;

f. Significant operating abnormalities or deviations from normal and expected performance of unit equipment that affect nuclear safety;

g. All REPORTA8LE EVENTS;

h. All recognized indications of an unanticipatsd deficiency in some aspect of design or operation of structures, systems, or components that could affect nuclear safety; and

i. Reports and meeting minutes of the 50RC.

O COMANCHE PEAK - UNIT 1 6-K 9

IXX 88512 .

ATTACHMENT 11 PAGE 23 0F 41 ADMINISTRATIVE CONTROLS AVOITS 6.5.2.8 Audits of unit activitias shall be performed under the cognizance of the ORC. The audits shall be performed within the specified time interval with a maximum allowable extension not to exceed 25% of the specified interval provided the combined time interval for any three consecutive intervals shall not exceed 3.25 times tPe specified interval. These audits shall encompass:

a. The conformance of unit nptration to provisions contained within the Technical Specifications a.1d applicable license conditions at least once per 12 months;

b. The performance, training, and qualifications of the entire unit staff at least once per 12 months;

c. The results of actio.is taken to carrect deficiencies occurring in unit equipaint, structures, systeas, or method of operation that affect nuclear safety, at least once per 6 months;

d. The performance of activities required by the Operational Quality Assurance Program to meet the cr,1.teria of Appendix 8, 10 CFR 50,.2t least once per 24 months;

e. The fire protection programmatic controls including the implementing procedures at least once per 24 months by qualified ifcensee QA personnel;

f. The fire protection equipment and program impleesntation at least once per 12 months utilizing either a qualified offsite licensee fire protection engineer or an outside independent fire protection consultant. An outside independunt fire protectinn consult 3nt shall be used at laaet every third year; T45str Y w 9* 2f__I f [{};j]ff] h_((j@$jf 5 NC id W U

10 8: 0531

h. The OFFSITE DOSE CALCULATION MANUAL and implementing procedures at least once per 24 months;

i. The PROCESS CONTROL PROGRAM and iuplementing procedures for processing and packaging of radioactive wastes at least once per 24 months;

j. The performance of activities required by the Quality Assurance Program for effluent and environmental monitoring at least once per 12 months; e i

k. Any oth6r area of unit operation considered appropriate by the ORC or the Executive Vice President Nuclear Engineering and Operationga,J TAssar '8'----3p O

10 8: 0534 l

COMANCHE PEAK - UNIT 1 6M 10

AliACEMENT17 i PAGE 24 0F 41 4

Oi j 'INSEdTF FOR NEW PAGE 6-10 (OLD 6-12) SECTION 6.5.2.8 l

nan

l l

 ,                                                                         i

g. The performance of activities in accordance with the RADIOACTIVE EFFLUENT AND ENVIRONMENTAL MONITORING MANUAL at least once per 24 months; l i g.,

1. The performance of activities required by the Technical Specification Improvement Program at least once per 24 months.

O O

TXX-88512 ATTACHMENT 17 PAGE 25 0F 41 ADMINI'STRATIVE CONTROLS RECORDS 6.5.2.9 Records of ORC activities shall be prepared, approved, and distribu-ted as indicated below:

a. Minutes of each ORC meeting shall be prepared, approved, and for-warded to the Vice President, Nuclear Operations and Executive Vice President, Nuclear Engineering and Operations within 14 days following each meeting;

b. Reports of reviews encompassed by Specification 6.5.2.7 shall be prepared, approved, and forwarded to the Vice President Nuclear Operations and Executive Vice President, Nuclear Engineering and Operations within 14 days following completion of the review; and i

c. Audit reports encompassed by Specification 6.5.2.8 shall be for-warded to the Vice President, Nuclear Operations and Executive Vice President, Nuclear Engineering and Operations and to the management l

positions responsible for the areas audited within 30 days after completion of the audit by the auditing organization. 6.5.3 TECHNICAL REVIEW AND CONTp0M 6.5.3.1 Activities which affect nuclear safety shall be conducted as follows:

a. Procedures required by Specification 6.8 and other procedures which affect plant nuclear safety, and changes thereto, shall be prepared, reviewed and ape ~ved. Each such procedure or procedure change I

i shall be review a qualified individual / group.other than the

                          ' ~ individual / group which preparoc the procedure or procedure change, but who say be from the same organization as tha individual / group ID I: 0908 which prepared the procedure or procedure change. The Vice Presi-             l dent, Nuclear Operations, shall approve Station Administrative Procedures, Security Plan Implementing Procedures, and Emergency l                        -

Plan Implementing Procedures. Other procedurgs shall be approved by line manager, g;,, ;7,,,,;,;,, ^- rr ! rMr'ty, r fr';n:%d-by the Vice Presi-as previously designated

                             - dent, tvuclear operations, in writing.        Indiviotals responsible for procedure reviews shall be members of the Nuclear Operations Staff i                               previously designated by the Vice President, Nuclear Operations.

l Changes to procedures which do not change the intent of approved

9eocedures may be approved for implementation by two members of the Nuclear Operations Staff, at least one of whom holds a Senior Operator License, provided such approval is prior to implementation and is documented. Such changes shall be approved by the original approval authority within 14 days of implementation; 1 B. Proposed tests and experiments which affect plant nuclear safety

                               =d r; :t ;ddr;;;;d '- th: "'n:1 Sf:t; 'n !y '; "- rt 07 7ee.;i.;;;.: 4;.iff;etier.; shall be prepared, reviewed, and f

eeting the experience requirements of ANSI N18.1-1971, Sections 4.2, 4.3, 4.4, l

   ~ 4.5.1 (Licensed Operators), 4.5.2, or 4.6 COMANCHE PEAK - UNIT 1                           6 'b( l l

IIX-88512

,             ATTACH"ENT17 PAGE 26 0F 4L a

S U . ..

                                                                                                  .Q.j 4

AONINISTRATIVE CONTROLS TECHNICAL REVIEW ANO CONTROLS (Continued) approved. i i

Each such test or experiment shall be reviewed by a ing 0908

                           \ qualified individual /groupaother than the individual / group which
                          / prepared the proposed test or experiment. Proposed test and experi-                 i I l ments shall be approved before implementation by the Manager, Plant Operations. Individuals responsible for conducting such reviews shall be members of the Nuclear Operations Staff previously l

designated by the Vice President, Nuclear Operations;

c. Proposed changes or modifications to plant nuclear safety-related i

structures, systems and components shall be reviewed as designated 1 by the Vice President, Engineering and Construction. Each such M modification shall be reviewed by a qualified individual /groLp IDI 0908 ! Wother than the individual / group which designed the modification. but who say be from the same organization as the individual / group l

which designed the modifications. Individuals / groups responsible for conducting such reviews shall be previously designated by the i

Vice President, Engineering and Construction. Proposed modifica-j i tions to plant nucleer safety-related structures, systems and

components shall be approved by the Manager, Plant Operations prior i to implementstion;

d. Eachreviewconductedinaccordahes"withtherequirementsoISpecl-fications 6.5.3.la, 6.5.3.lb, and 6.5.3.2c, shall include a deter-mination of whether or not additional cross-disciplinary ' review is necessary. If deseed necessary, such review shall be done in accordance with the appropriate qualification requirements;

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

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 " n:;;r, + g !

                              " hnt ^;;r:thn:. Recommended changes to the Plans shall be                 o      I reviewed pursuant to the requirements of Specifications 6.5.1.6 and         "

6.5.2.8 and approved by the " ::;:r, "h t ^;:r:thn:. NRC approval 2 l shall be obtained as appropriate. t.vmh.um g u m.o.r6%s 6.5.3.2 Records of the above activities described in 6.5.3.1 shall be provided to the Vice President, Nuclear Operations, SORC, and/or ORC as necessary for required reviews.

6. 6 REPORTA8LE EVENT ACTION 6.6.1 The following actions shall be taken for REPORTA8LE EVENTS:

a. The Commission shall be notified and a report submitted pursuant to the requirements of 10 CFR 50.73 and

b. Each REPORTA8LE EVENT shall be reviewed by the 50RC, and the results of this review shall be submitted to the ORC and the Vice President Nuclear Operations.

O, {meetingtheexperiencerequirementsofANSIN18.1-1971 Sections 4.2, 4.3, 4.4, or 4.6 COMANCHE PEAK - UNIT 1 6-K I 2, ' meeting the experience requirements of ANSI N18.1-1971, Section 4.6

IXE88512 ' ATTACHMENT 17 PAGE 27 0F 41 1 1 ADMINISTRATIVE CONTROLS i i 1 ! 6. 7 SAFETY LIMIT VIOLATION l 6.7.1 i The following actions shall be taken in the event a Safety Limit is violated i a. ! In accordance with 10 CFR 50.72, the NRC Operations Center, shall ! be notified by telephone as soon as practical and in all cases i within one hour after the violation has been determined. The Vice j President, Nuclear Operations and the Operations Review Committee (ORC) shall be notified within 24 hours, i b. A Licensee Event Report shall be prepared in accordance with 10 CFR 50.73 i c. 10 1: 0536 The License Event Report shall be submitted to the Commission in accorcani i l W to tu with 10 CFR 50.73 .Vice President, Nuclear Operations 4 and .ne 0 ;erations l Review Goem1ttee (ORC) within 30 days after discovery of the event.

d. .ct t:es o nt p.n. Ae,:w c...aeuO* Q Critical operation of the unit shall not be resumed until authorized $

by the Nuclear Regulatory Commission. ID'I: 0914 6.8 PROCEDURES AND PROGRAMS 6.8.1 Written procedures shall e established, faplemented, and main ained covoring the activities referenc6J below:

a. The applicoble procedures recommended in Appendix A of Regulatory j Guide 1.33, Revision 2, February 1978;

b. The emergency operating procedures required to implement the require-monts of NUREG-0737 and Supplement 1 to NUREG-0737 as stated in Section 7.1 of Generic Letter No. 82-33;

c. Security Plan implementation;

d. Energency Plan implementation;

e. PROCESS CONTROL PROGRAM implementation;

f. OFFSITE DOSE CALCULATION MANUAL implementation; and-g.

Im re.r .A , % Quality Assurance for effluent and environmental monitori( ID I: 0541 t g 6.8.2 h:h;==d=: Of !;=i'f =tf = 5.:.1, rd th=;= th=:::, :h:!' 5: r= ? rd t-; tM 50"" = d :.%1' 5: :;;==d t-j th: " f = "r = i d=0, "= ! = r bII! E5151b.!?

                                                                                                     }               '     '

Iaseer T :. :. : = = =., = =g= = ;= =d. = f !;=ifi= e = :.:.1 => a = = a m.: =

. 'h: 'nt=t :f th: =fgi=1;==d=: i; = 0 : :=;d; 10 i m2

5. h N. =. h hh7IA w'[ w'A A 7- ; ' ' ' ' '

O

                                                                                                             ' h'                 8

t !:=t =: Of ch = Mid: : Sr f = 0;;=tr ' i==: = th; Ci t

f f=t:d; =d -

COMANCHE PEAK - UNIT 1 6-K#3

TXX-88512 ATTACHMENT 17- - PAGE 28 0F 41

INSERTS FOR NEW PAGE 6-13 (OLD 6-15) SECTIONS 6.8.1 & 6.8.2 nan

h. RADIOACTIVE EFFLUENT AND ENVIRONMENTAL MONITORING 10 1 0538 MANUAL implementation; g h Fire Protect Program implementation; and ID l 0537 i

Technical Specification Improvement Program

     )h        implementation.

ID I: 0540 l

                                         .. g .,

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. I l l t i O

TXX-88512 . AtlACHMENT17 PAGE 29 0F 41

     \        -        -                                                               DPF, f ADMINISTRATIVE CONTROLS O

PROCEDURES AND PROGRANS (Continued)

. 'h: ch: ;; f: d::rrted, re;f:r:d by the S^"C, ::d :;;r:;;; by in

                      '";; Pr;;f d;37, ";;h:r 0;:r:tf::: .;f th'- M d:y: Of f-H :::,,t: t:,

6.8.\3 The following progrees 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 systees include the post accident recirculation portions of the Containment Spray Systee, Safety Injection System, Chemical and Volume Control System, RHR System, and RCS Sampling System (Post Accident Sampling Systes portion only). The program shall include the following:

1) Preventive maintenance and periodic visual inspec". ion require-monts, and

2) Integrated leak test requirements for each system at refueling cycle intervals or "ess.

b. In-Plant Radiation Mo'nitorina A program which will ensure the capability to ac:vratoly determine the airborne f odine concentration in vital areas under accident conditions, This, prog.as shall include the following:

1) Training of personnel.

2) Procedures for monitoring, and

                    ;O        Provisions for maintenance of sampling and analysis equipment.

c. hcondsryWaterChemistry 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,

2) Identification of the procedures used to measure the valu es of the critical variables,

3) Identification of process sampling points, which shall include aonitoring the discharge of the condensate pumps for evidence of condenser in-leakage,

4) Procedures for the recording and management of data,

5) Procedures defining corrective actions for all off-control O a 4"' c* ' trv c "d'*4 "a COMANCHE PEAK - UNIT 1 6-h IQ

i IXX-88512 ATTACMN 17 PAGE 30 0F 41 ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued) 1

6) A procedure identifying: (a) the authority responsible for the interpretation of the data, and (b) the sequence and timing of i

administrative events required to initiate corrective action, i

d. 90st-Accident Samplino A program which will ensure the capability to obtain and analyze j, reactor coolant, radioactive iodines and particulates in plant 1 gaseous effluents, and containment atmosphere samples under accident

{ conditions. The program shall include the following: l j 1) Training of personnel,

2) Procedures for sampling and analysis, and
3) Provisions for maintenance of sampling and analysis equipment.

e, RADI0 ACTIVE EFFLUEYr AM) ENVIR000tEMTAL MONITORING MANUAL (REDet) The RADI0 ACTIVE EFFLUENT AND ENVIR0 MENTAL MONITORING MANUAL ' I.REEM).shall outline .the affluent and environountal sampling and analysis program used to detaruina the concentration of radioactive materials in those pathways which lead to radiation O exposures to MEMBER (5) 0F THE PUBLIC froe routine station operation. The a4nual provides the following: ID 1: 0547

1) Input to the 0FFSITE DOSE CALCULATION MANUAL (00CM) methodologies for calculating liquid and gaseous effluent concentrations and offsite doses, 2). Guidellt.as for operati q radioactive waste treatment em,J.,systeinst in' order <that4ffsttEdoses wilt be'kept as' low as reasonably achievable (AL/RA),

3) Definition of the Radiological Environmental Monitoring a.;.. s* . . Progras. (REMP), which provides. confirsation that . -

                                  -(.oncentrations of' padioactive material released from CPSES arte not higher than expected.

4) An outline of information required to be submitted to the NRC through the Semiannual Radioactive Effluent Release Report and the Annual Radiologict1 Environmental Operating Repgrt, and

5) Administrative requirements for implementation, including the review and approval

process for the manual and changes theretu.

O COMANCHE PEAK - UNIT 1 6-K ls- ,

TXX-88512 AllACHMENT17 i PAGE 31 0F 41 ADMINISTRATIVE CONTROLS

6. 9 REPORTING REQUIREMENTS ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Regional Administrator of the Regional Office of the NRC unless otherwise noted.

STARTUP REPORT

6. 9.1.1 A summary report of plant startup and power escalation testing shall be submitted following: (1) receipt of an Operating License, (2) amendment t; the license involving a planned increase in power level (3) installation of fuel that has e different design or has been manufactured by a different fuel supplier, and (4) modifications that say have significantly altered the nuclear, thermal, or hydraulic performance of the unit.

1he initial Startup Report shall address each of the startup tests identi-fied in Chapter 14 of the Final Safety Analysis Report and shall include a description of the measured values of the operating conditions or char.acterjse tics obtained during the tast program and 'a i:~omparison of these values with design predictions and specifications. Any corrective actions that were required to obtain sstisfactory operation shall also be described. Any additional specific details required in license conditions based on other O commitments shall be incluaed in this report. Subsequent Startup Reports shall address startup tests that are necessary to demonstrate the acceptability of changes and/or modifications. Startup Reports shall be submitted within: (1) 90 days following coespletion of the Startup Test Program, (2) 90 days following resumption or commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest. If the Startup Report does not cover all l i three events (i.e., initial criticality, completion of Startup Test Program, and resumption or commencement of commercial operation), suptlementary reports shall be submitted at least every 3 months until all three events have been completed. ANNUAL REPORTS

6.9.1.2 Annual Reports covering the activities of the unit as described below for the previous calendar year shall be submitted prior to March 1 of each year. The initial report shall be submitted prior to March 1 of the year following initial criticality. ,

Reports required on an annual basis shall include:

a. A tabulation on an annual basis of the number of station, utility, and other personnel (including contractors) receiving exposures O *A single submittai may be made for a multiple unit station. The submittal should combine those sections that a?e common to all units at the station.

COMANCHE PEAX - UNIT 1

6-1( 14 l

TXX-88512 ATTACHMENT 17 PAGE 32 0F 41 ADMINISTRATIVE CONTROLS DRAH G ANNUAL REPORTS (Continued) l greater than 100 area /yr and their associated man res exposure according to work and job functions

e.g., reac:or operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling.

The dose assignments to various duty functions may be estimated based on pocket dosimeter, thermoluminescent dosimeter (TLD), or film badge measurements. Small exposures totalling less than 20% of the individual total dose need not be accoanted for. In the aggregate, at least 80% of the total whole-bcdy dose received from external sources should be assigned to specific major work functions;

b. The results of specific activity analyses in which the primary coolant exceeded the limits of Specification 3.4.8. The following information shall be included: (1) Reactor power history starting 48 hours prior to the first sample in which the limit was exceeded (in graphic and tabular forsat); (2) Results of the last isc topic analysis for radio-iodine performed prior to exceeding the Ifmit, results of analysis while limit was exceeded and results of one analysis after the radio-iodine activity was reduced to less than limit. Each result should include date and time of sawling and the radiofodine concentrations; (3) Clean-up flow history starting 48 t.ours prior to the first sample in which the limit was exceeded; (4) Graph of the I-131 conceatra-O tion (pci/gn) and one other radicioine isotopu concentration (pCi/ge) as a function of time for the duration of the specific activity acove

, the steady-state level; and (S) The time duration when the specific activity of the primary coolant exceeded the radiciodine limit. ANNUAL RAOI0 LOGICAL ENVIRONMENTAL OPERATING REPORT ** 6.9.1.3 Routine Annual Radiological Environmental Operat.ing Reports covering the operation of the unit during the previous calendar year shall be sucaitted prier to May 1 of each year. The initial report shall to submitted prior to May 1 of the year following initial criticality and shaiI include copies of reports of the preoperational Radiological Environmental Monitoring Program of the unit for at least two years prior to initial criticality. nese, ryert's Sk*.3 4dv4s the.t informAim del;neded b Ma Rrg ru.

           ' M ^ n. 1 tdt;?:;i;;i ircs'r; c at:1 0;;r ting ti;;rt; ;h;i' 'n:1;d;

m rt::, '-t: pr:::ti;;;, :nd :n :::ly:f: Of tr:nd; of th; r;;;l0; ;f th:

r:di:?:;ic:? :nvir:r ;nt:! : :f'!:::: ::tiviti:: f:n th: r;;;rt ;;rt:d, 10 I: 0550 "This tabulation supplements the requirements of 4M. iO? ;f 10 CFR 20.407.

   ""A single submittal may be made for a multiple unit station.

O CCMANCHE PEAK - UNIT 1 6*'$ 17

Ixx-eral2 AtlACHMEHi 17 PAGE S OF 41 Oh ADMINISTRATIVE CONTROLS 4NNWAL ."*"!OLOO!;^ L !?!!"0'":'v2 L Oa =^v!g egna,av (;;,_,; 5 ;) inci ing a comparison with preoperational studies, with operational cent s, as appr assessmen riate, and with previous environmental surveillance reports, an The reports f the observed impacts of the plant operation on the env oneent. Specification 11 .also . 2. include the results of the Land Use Census quired by

                         .       The Annual Radi               gical Environmental Operating Repor results of analysis of                     1 radiological environmental s lesshall                               include the and of all environmental radiation se urements taken during th eriod pursuant to the locations specified in the t                          e and figures in t Manual, as well as summarized a                                                           Offsite Oose Calculation tabulated res ts of these analyses and measureewnts in the format of the                                   le in t         Radiological Assessment Branch Technical Position, Revision 1, Nov                                   r         9.

dual results are not available for inc1 on with In the event that some indivi-the report, the report shall be submitted noting and explaining t missing data shall be subeitted a oon as po ible the s for missing results. The rea in a supplementary report. The reports shall also lude the following: summary description of the Radiological Enviro al Monitoring Jrogram; at est two legible sans* covering all samp1tng 1 .ations keyed to a table giving stances and directions from the contarline one reactor; the results of license articipation in the Interlaborato onparison Program and the corrective act specified progr- is not being performed as raquired by Specifictaken if the ' ion 3.12.3; reasons for cenducting the Radiological Environmental i4onitoria Program as requirlid b specification 3.12.1, and discussion of all deviations fr the l samplin schedule of Table 3.12-1; discussion of environmental sample me re-ment hat exceed the reporting levels of Table 3.12-2 but are not the resu of art affluents, pursuant to ACTION b. of Specification 3.12.1; and discuss n

                      . . :11 r:! y:n '                .a i :h the-%G . . ,. . . .. ., r.. . . -... . .-. .... ............

j SEMIANNUALRA010ACTIVEEFFLUENTRELEASEREPORT*\ I

6. 9.1. a Routine Semiannual Radioactive Effluent Release Reports covering the operation of the unit during the previous 6 months of operation shall be submitted within 60 days aft?r January 1 and July 1 of each year. The period ,

t shall of the i'irst repo{t *M ", begin M with the date p*f init al criticality. 7"Assa.

                     "*P*%..'Mi*f?_  . . . .. .. . . . . .. . . .. . . '.".'il*,a. i'_1* *2.!,*k. ._ ._. .D, *_

E AA M t , _ _ , m _ r -- y ;f 0.; ;=ntitin :f =0hnth; l';;'d ud ;nn= ;f.nt; =d

lid :=^; n h n;d '7s th: unit n ::0!'n d '- ".:;; Mt:-; 0;id: 1.21,

                     ""      =;r'rg, E-h: ting, ud "- ;rth; "-dunthit-,'- MiidUntu =d
                     "; h = = f "-d h n t h: ":t:rhi; '- Li;;fd = d On n n lff h n t^ fin W 0550 l Light t t r-0= hd "ni n ran:r ."i nt ," "= h f = 1, .';n 10 , i th d-0; es tet.use(AtsA                                                                 {

:. f':^ I.

                       \*A single submittal say be made for a multiple unit station. The submittal i

should combine those sections that are cosumon to all units at the station, however, for units with separate radweste systees, the submittal shall specify the releases of radioactive material from each unit. COMANCHE PEAK - UNIT 1 6-K l 3 . l

IXX-88512 AliACHMENT 17 PAGE 34 0F 41

                     -       ~

ADMINISTRATIVE CONTROLS O 'CM:fC". G:0t.CT '!C CIIL"CNT RCLCMC R=T 'C;, ,t' .W) t s rized on a quarterly basis following the format of Appendix 8 thereof For olid wastes, the format for Table 3 in Appendix 8 shall be supplemen ed with ree additional categories: class of solid wastes (as defined by 0 CFR Part 6 , type of container (e.g., LSA, Type A, Type 8, Large Quantity and l SOLIDIFI TION agent or absorbent (e.g., cement, urea formaldehyde). l i The 5 fannual Radioactive Effluent Release Report to be sube ted within ) 60 days aft January 1 of each year shall include an annual sum ry of hourly j meteorologica data collected over the previous year. This ann 1 summary may ; be either in t form of an hour-by-hour listing on magnetic t e of wind speed, i wind direction, teospheric stability, and precipitation (if asured), or in l the form of joint requency

distributions of wind speed, wi direction, and '

atmospheric stabili . This same report shall include a assessment of the , radiation doses due the radioactive liquid and gaseou effluents released l from the unit or stati during the previous calendar y ar. This same report ' shall also include an a essment of the radiation dos from radioactive liquid ! and gaseous effluents to ERS OF THE PU8LIC due t their activities inside ' the SITE BOUNDARY (Figure .1-3]) during the repo period. All assumptions used in making these assess nts, i.e., specific tivity, exposure ti_me, and location, shall be. included i these reports'l e meteorological conditions concurrent with the time of re ese of radioac ve materials in gassous efflu-ents, as determined by sampling equency an seasurement, shall be used for determining the gaseous pathway do es. The ssessment of radiation doses shall be oerformed in accordance wi'h tho' th O DOSE CALCULATION MANUAL (00CN. ogy and parameters in the OFFSITE The Semiannual Radioactive Efflu Release Report to be submitted within 60 days after January 1 of each year shal also include an assessment of radia-tion doses to the likely most expo. d MEM OF THE PUBLIC free reactor releases and other nearby uranium fuel cyc sources, including doses from primary efflu-

          . ent pathways and direct radiati            , for the pr ious calendar year to show con-

. formance with 40 CFR Part 190, Environmental R iation Protection Standards for l Nuclear Power Operation." A eptable methods fo calculating the dose contribu-l tion from liquid and gaseou effluents are given i Regulatory Guide 1.109, Rev. 1, October 1977. t The Sesiannual Ra osctive Effluent Release Retpor shall include a list and description of u anned releases from the site to ESTRICTED AREAS of l radioactive matarial in gaseous and liquid effluents sad during the reporting ( period. The Seeian al Radioactive Effluent Release Reports'shall include any changes made d ing the reporting period to the PROCESS CONTROL ROGRAM (PCP)

            *In lieu           submission with the Seeiannual Radioactive Effluent Role e Report the Itcensee has the option of retaining this summary of requ red meteo logical data on site in a file that shall be provided to the NR upo request.
            ===                   =                      m                .                           ,

L___.____._____________.

TX1-6'8512 ' ATTACHMENT 17 PAGE 35 0F 41 A0M!i!'57RATIVECONTROLS O CO^!' C 'L "^^IO^O !Y: lP"LUC."' "!LE^5! "'a^^' (C;ntin;;;; and to th SITE DOSE CALCULATION MANUAL (00CM), pursuant to So tions 6.13 and ca-respectively, as well as any major che o Liquid, . Gaseous, or Solid Ra e Treatment Systees pursuant pecification 6.15. W 055O It shall also include a 1 of new locations dose calculations and/or environee. ital monitoring identi the Use Census pursuant to Specifi-cation 3.12.2. 8 The Sesiannual Radio e Effluent Release the following: an e ts shall also include ation as to why the inoperabi f liquid or gaseous effluent monit instrumentation was not corrected within time specified in Specif on 3.3.3.10 or 3.3.3.11, respectively; and des aiptio eve the eading to liquid holdup tanks or gas storage tanks exceeding the

                  ..:i::    Of !;;;i'f::ti:n 2.11.1.^ Or 3.11.2.5, r :;n ;iv;ly.

MONTHLY OPERATING REPORTS

6. 9.1. 5 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the PORVs or safety valves, shall be submitted on a monthly basis to the Director, Office of Resource

'               Management, U.S. Nuclear Regulatory Commission, Washington, D.C. 2055f, with a copy to the Regional Administrator of the Regional Office of the NRC, no later than the 15th of each month fellowing the calendar month covered by the report.

RAOIAL PEAKING FACTOR LIMIT REPORT 6.9.1.6 The F # xy limits for RATED THEAMAL POWER (Fxy ) shall be established for at least each reload core and shall be maintained available in the Control Roos. The limits shall be established and implemented on a time scale consis-tent with normal procedural chaitges. The analytical methods used to generate the F,y limits shall be 4 hove-a: :: 5 reviewed and approved by the NRC,he If changes to these methods are , deteed necessary they will be evaluated in 'accordance with 10 CFR 50.59 and I '- submitted to the NRC for review and approval prior to their use if the change is determined to inv'olve an unreviewed safety question or if such a change for'df(5wouldrequireamendmentofpreviouslysubmitteddocumentation. Spedf.'c 10 8: 0556 I u s e., A report containing the F,y limits for all core planes containing Bank "0" control rods and all unrodded core planes along with the plot of predicted i l F q P axial core height (with the limit envelope for comparison) shall l be provide to the NRC Document Control desk with copies to the Regional Admin-istrator and the Resident Inspector within 30 days of their implementation.

             "'O'" 0205        ""=r "i; trit.ti;n 0;ntr;i :nd L;;d T;11; sing Tre;;4.ree" .nd WCf."
                ?270.^      ""n    1n f;;;; ";?::d.::f:03 5 ;i;:0f:n ":tt:d:::;y."

COMANCHE PEAK - UNIT 1 6-M li ,

i IXX-88512

&liACHMENT17

, PAGE 36 0F 41

               ~       '

o ADMINISTRATIVE CONTROLS SPECIAL REPORTS 4

{ 6.9.2 Special reports shall be submitted to the Regional Administrator of the Regional Office of the NRC within the time period specified for each report. 6.10 RECORD RETENTION 6.10.1 In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated. 6.10.2 The following records shall be retained for at least 5 years:

a. Records and logs of unit operation covering time interval at each power level; b.

Records and logs of principal maintenance activities, inspections, repair, and replacement of principal items of equipment related to nuclear sa'ety;

c. .ill REPORTA8LE EVENTS; ~ ' ' ' '

to'1:0557

d. Records of surveillance activities, inspections, and calibrations required by 4heee-Technical Specifications? :

us

e. Records of changes made to the procedures required by Specification 6.8.1;

f. Records of radioactive shipments;

g. Records of sealed source and fission detector leak tests and results; and

h. Records of annual physical inventory of all sealed source material of record.

ed. M 44, SSe e di. a fis N tM es4Y NM 58F% hNk%#k 3 Pro 1m ud Rnsode.nvs RFFwsMY Adb 24WRMMEdYAL maarcams maaonu, sac.sps as expI;<.;bly covsesd in Spe.c.Mie.4..m G.so.3 O COMANCHE PEAK - UNIT 1 6-M,lo

l

IXX-88512 ATTACHMNT17 PAGE 37 0F 41 Q

                                             .     -                                                                   y ADNT4!STRATIVE CONTROLS REC]RO RETENTION (Continued)

6. '.0. 3 The following records shall be retained for the duration of the unit Oparating License: :

a. Records and drawing changes reflecting unit design modifications made to systems and equipment described in the Final Safety Analysis Report;

b. Records of new and irradiated fuel inventory, fuel transfers, and assembly burnup histories;

c. Records of radiation exposure for all individuals entering radiation control areas;

d. Records of gaseous and liquid radioactive material released to the environs;

e. Records of transient or operational cycles for those unit components identified in Table 5.7-1;

f. Records of reactor tests and experiments;

g. Records of training and qualification for current members of~the -

unit staff;

h. Records of inservice inspections performed pursuant to these Technical Specifications;

i. Records of quality assurance activities required by the Quality Assurance Manual;

j. Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59;

k. Records of meetings of the 50RC and the ORC;

1. Records of the service lives of all hydraulic and mechanical 10 8: 0559 snubbers required by ;;i'f;;ti;.. 2.7.^ including the date at which the service life ommences and associated installation and maintenance records; m %.kJ l Sge: Oca.h Lpmement Proysm

m. Records of secondary water sampling and water quality; and

n. Records of analyses required by the Radiological Environmental Monitoring Program that would permit evaluation of the accuracy of the analysis at a later date. This should include procedures effective at specified times and QA records showing that these l procedures were followed.

1 6.11 RADIATION PROTECTION PROGRAM l ! 6.11.1 Procedures for personnel radiation protection shall be prepared consistent l with the requirements of 10 CFR 20 and shall be approved, maintained, and adhered to for all operations involving personnel radiation exposure. O COMANCHE PEAK - UNIT 1 6-h .2) . 1

Txx-88512 ATTACEENT 17 PAGE 38 0F 41 l ADMINISTRATIVE CONTROLS 6.12 HIGH RADIATION AREA 6.12.1 Pursuant to paragraph 10 CFR 20.203(c)(5), in lieu of the "control device or "alarm signal" required by paragraph 10 CFR 20.203(c), each high radiation area, as defiried in 10 CFR 20, in which the intensity of radiation is equal to or less than 1000 mR/h at 45 cm (18 in.) from the radiation source or from any surface whicn the radiation penetrates shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP). Indi-viduals qualified in radiation protection procedures (e.g., Radiation Protsc-tion Technician) or personnel continuously escorted by such individuals may be exempt from the RWP issuance requirement during the performance of their assigned duties in high radiation areas with exposure rates equal to or less than 1000 nR/h, provided they are othenvise following plant radiation protec-tion proceduns for entry into such high radiation areas. Any individual or group of individuais permitted to enter such areas shall be provided with or accompanied by one or more of the following:

a. A radiation monitoring device which continuously indicates the radiation dose rate in the area; or b.

the A radiation radiation dose monitoring rate in the amadevice which and'alares whencontinuously a preset int integrate e dose is received. Entry into such areas with this monitoring device may be made after the dose rate levels in the ama have been estab-lished and personnel have been made knowledgeable of them; or O c. An individual qualified in radiation protection procedures with a radiation dose rate monitoring device, who is responsible for pro-viding positive control over the activities within the area and shall perform periodic radiation surveillarce at the fMquency specified by the RWP. 6.12.2 In addition to the requirements of Specification 6.12.1, areas seces-sible to personnel with radiation levels gnator than 1000 mR/h at 45 cm (18 in.) from the radiation source or from any surface which the radiation penetrates shall be provided with locked doors to prevent unauthorized entry, and the keys shall be maintair,ed under the administrttive control of the Shif t Supervisor on duty and/or radiation protection supervision. Doors shall remain locked except during periods of access by personnel under an approved RWP which shall specify the dose rate levels in the immediate work areas and the maximum allowable stay time for individuals in that area. In lieu of the stay

'   time specification of the RWP, direct or remoto (such as closed circuit TV cameras) continuous surveillance say be made by personnel qualified in radia-tion protection procedures to provide positive exposure control over the activities being performed within the area.

For individual high radiation areas accessible to personnel with radia-tion levels of greater than 1000 mR/h that are located within large areas, such as PWR containment, where no enclosun exists for purposes of locking, 1 l and where no enclosure can be reasonably constructed around the individual area, that individual an a shall be barricaded, conspicuously posted, and a flashing light shall be activated as a warning device. O COMANCHE PEAK UNIT 1 6-h A;L ,

l

d T1X-88512 ) AllACHMENT 17 i PAGE 39 0F 41 J $ ADMINISTRATIVE CONTROLS . i O- ! 6.13 PROCESS CONTROL PROGRAM (PCP) t 1 1 6.13.1 The PCP shall be approved by the Commission prior to implementation. , l 6.13.2 Licensee-initiated changes to the PCP: a. Shall be submitted to the Commission in the Seelannual Radioactive Effluent Release Report for the period in which the change (s) was

                       .                  made. This submittal shall contain:                                           ;

1) Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information;

2) A determination that the change did not reduce the overall conformance of the solidified waste product to existing criteria for solid wastes; and

3) Documentation of the fact that the change has been reviewed and found acceptable by the 50RC.

b. Shall become effective upon revief and acceptance by the 50RC.

6.14 0FFSITE DOSE CALCULATION MANUAL. (00CM) 6.14.1 The 00CM shall be approved by the Commission prior to implementation. 6.14.2 Licensee-initiated changes to the 00CM: a. Shall be submitted to the Commission in the Seelannual Radioactive Effluent Release Report for the period in which the change (s) was made effective. This suhittal shall contain:

1) Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information. Information submitted should consist of a package of those pages of the 00CM to be changed with each page ntaubered, datsd and containing the revision number, together with appropriate analyses or evMuations justifying the change (s);

2) A determination that the change will not reduce the accuracy or reliability of dose calculations or Setpoint determinations; and

3) Documentation of the fact that the change has been reviewed and found acceptable by the 50RC.

b. Shall become effective upon review and acc6pt m by the 50RC. '

N SERT * 'A' > l ID h 0%2 COMANCHE PEAK - UNIT 1 6- K 13 ,

TXX-88512 ATTACHMENT 17 PAGE 40 0F 41

INSERT FOR NEW PAGE 6-23 (OLD 6-26) NEW SECTION HAH 6.15 RADIOACTIVE EFFLUENT ENVIRONMENTAL MONITORING MANUAL (REEMM) 6.15.1 The REEMM shall be approved by the Commission prior to implementation. 6.15.2 Licensee initiated changes to the REEMM:

a. Shall be submitted to the Commission for approval prior to implementation. This submittal shall cor.tain:

(1) Sufficiently detailed information to totally-support the rationale for the change without benefit of additional or supplemental information. (2) A determination that the change does not reduce the overall effectiveness or reliability of the radiological effluent or environmental monitoring programs. (3) Documentation of the fact that the change has been reviewed and found acceptable by SORC.

b. Shall become effective upon review and acceptance by the Commission O

TXX-88512 AliACHMENT 17 PAGE 41 Of 41 i ADMINISTRATIVE CONTROLS O /6 6.'M MMOR CHANGES 70 LIQUID. GASEOUS, AND SOLIO RA0 WASTE TREATMENT SYSTEMS

16 I

6. 'M.1 Licensee-initiated major changes to the Radwaste Treatment Systens (liquid, Caseous, and solid): ;

a. Shall be reporced to the Commission in the Sesiannual Radioactive Ef fluent Release Report for the period in which the evaluation was reviewed by the 50RC. The discussion of each change shall contain:

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

2) Sufficient detailed information to totally support the reason for the change without benefit of additional or supplemental information;

3) A detailed description of the equipment, mponents, and processes involved and the interfaces with other plant systees;

4) An evaluation of the change, which shows the predicted releases of radioactive materials in liquid and gaseous effluents and/or '

quantity of solid waste that differ from those previously predicted in the License application and asenhnts therets;

5) An evaluation of the change, which shows the. expected wxieue exposures to a MEMBER OF THE PU8LIC in the UNRESTRICTED AREA Oe and to the general population that differ from those previously estimated in the License application and amen h nts thereto; l 6) A comparison of the predicted releases of radioactive matJrials,

! in liquid and gaseous effluents and in solid waste, to the actual releases for the period prior to when the change is to be made;

7) An estimate of the exposure to plant operating personnel as a result of the change; and l 8) Documentation of the fact that the change was reviewed and

     ,                       found acceptable by the 50RC.

b. Shall become effective upon review and acceptance by the SORC.

         "Licensees may choose to submit the information called for in this Specification l          as part of the annual FSAR update.

t tO COMANCHE PEAK - UNIT 1 6-M A4 ' l l l a - _ - - _ . - . - - _ _ _ _ _ ._ - -}}

ML20195G764

Text

SUBJECT:

4.7.3 EachcomponentcoolingwaterloopshalfbedemonstratedOPERA8LE

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