ML20133K816

Proposed Tech Spec Changes to Section 1.0, Definitions, for Inclusion in 850625 Submittal

ML20133K816
Person / Time
Site:	Zion  File:ZionSolutions icon.png
Issue date:	10/07/1985
From:	COMMONWEALTH EDISON CO.
To:
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ML20133K804	List: ML20133K802 ML20133K816
References
NUDOCS 8510220308
Download: ML20133K816 (23)

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ATTACHMENT PROPOSED TECHNICAL SPECIFICATION CIWES ASSOCIATED WITH SECTION 1.0 - [EFINITIONS (Pages listed below should be incorporated with those pluvided in Attachment 1 of the reference)

Pages added to Pages modified from original suomittal original submittal i 226' viii 302 1 303 2

3 4

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6a 6b 'Page 226 will be submitted 6c by Noventer 15, 1985 27a 36 201 243 244 244a 245 246 246a 0587K h

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a TAOLE OF C051ERIS 1.9 SEFle!TItuS .

Page 1 1.1 Attlen 1.26 Offsite Det !alculation Manual 1.2 Acteetten Device 1.27 Operable - Operability 1.3 Acteated Egelpment 1.4 acteatten Logic Test Page 5 1.28 Operating 1.5 Antal Fles Difference 1.2g Operating Cycle 1.6 Batch telease 1.30 Operational Mode - Mode 1.7 Channel Ca11bratten, Instrument 1.31 Physics Tests 1.8 Channel Check 1.32 Pressure Seendary Leakage 1.33 Process Centrol Program (PCP)

Page 2 1.g Channel Tenctlenal Test 1.34 Protection Logic Channel 1.10 Composite Sample 1.35 Protectlen System 1.11 Csetatement Integrity 1.12 Centleeses telease Page 6 1.36 Purge - Purging 1.13 Centre 11ed Leakage 1.37 Ouadrant Power Tilt Ratio 1.14 Core Alteratten 1.38 Rated Thermal Peuer 1.3g Reacter Pressure Page 3 1.15 Defleed Terms 1.40 Refueltag Outage 1.16 segree of Sedundancy 1.41 Reportable Event 1.17 Dese Egetwalent 1-131 -

1.42 Shutdeun Margin 1.19 E - Average 91stategratten Energy 1.43 Site Soundary 1.19 Esseems Radueste Treatment System 1.44 Solidification 1.20 Ideettfled Leekage Page 6a 1.45 Source Check Page 4 1.21 !astrement Channel 1.46 Surveillance Frequency Notation 1.22 Leakage 1.47 Thermal Power 1.23 Rester Belay Test 1.48 Unidentified Leakage 1.24 Rueer(s) of the Public 1.4g Carestricted Area 1.25 Off-Site AC Fewer 1.50 Vent 11atten Exhaust Treatment System Searces 1.51 venting LIMITlWG SAFETY SYSTER SETPOINT SAFETY ListTS 2.1 Protective Instrumentation Setpoints 1.1 Reactor Core Bases 2.2 Protective Equipment setpoints 1.2 Beacter Coelaat Systen Pressere Bases e vrv~wm n

Tab *e M 1.1 Operational Modes 6b 1.2 Serve 111ance Frequency Notation 6c 3.1-1 Reacttc Protection Systes-Limiting Operations Conditions and Setpoints 30 3.1-2 Reactor Protection System Instrument puebers 33 3.3.2-1 RTN01 lesting Results 88 3.3.4-1 In Service Inspection Program 106 3.3.5-1 Reactor Coolant Systems an4 Chemistry Specifications 122 3.4-1 Engineered Safeguards Actuation System-Limiting Conditions on 129 Operation and Setpoints 3.4-2 Engineered Safeguards System Instrument Numbers ,

132 3.7-1 Neutron Flux High Trip Points with Steam Generator Safety Valves 160a Inoperable - Four Loop Operation 3.7-2 Neutron Flux High Trip Points with Steam Generator Safety Valves 160b I operable - Three Loop Operation 3.11-1 Raatsum Permissible Concentration of Olssolved or Entrained Noble 226a Gases Releases from the Site to Unrestricted Areas in Liquid Effluents 3.11-2 Radioactive Liquid Effluent Monitoring Instrumentation 228 3.12-1 Radioactive Gaseous Effluent Monitoring Instrumentation 236 3.14-1 Radiation Monitoring Instrumentation 251 3.15-1 Equipment Requirement with Inoperative 4KV E.S.S. Bus 266 3.15-2 Equipment Inoperable with Inoperative 4KV E.S.S. Bus 267 LIST OF TABLES 09150/09190 v111

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! t j 1.0 MFlWITIONS 1.1 ACTION 1.5 AXIAL FLUE DIFFERENCE (AI)

)

q ACTION shall be that part of the Specification ARIAL FLUX DIFFERENCE (41) shall be the i

which prescribes remedial measures required under difference in normalized flux signals between i j designated conditions. the top and bottom halves of a two section

exccre neutron detector.

i 1.2 ACTUATION MVICE J 1.6 SATCH RELEASE j An ACTUATION MVICE shall be a component or I assembly of components that directly controls the A BATCH RELEASE is the discharge of 11guld I motive power (electricity, compressed air, etc.) wastes of a discrete volume. Prior to sampling  ;

I for ACTUATED E0ulPRENT. The following are for analyses, each batch shall be isolated and  !

) examples of an ACTUATION M VICE: circuit breaker, then thoroughly mixed to assure representative

} relay, and a valve (and its operator) used to sampling. , j i control compressed air to the operator of a .

l containment isolation valve. 1.T CHANNEL CAllSRATION. Instrument l

1.3 ACTUATES EOUIPRENT A CHANNEL CALIBRATION shall be the adjustment, i as necessary, of the channel such that it ACTUATES EQUIPRENT shall be a component or responds with the necessary range and accuracy assembly of components that performs or .11rectly to known values of input. The CHANNEL

, contributes to the performance of a protective Call 8 RATION shall encompass the entire channel j function such as a reactor trip, containment including the sensors (where possible), alarm

! isolation, or emergency coolant injection. the interlock and/or trip functions and shall

{ following are examples of ACTUATES EQUIPRENT: an include the CHANNEL FUNCTIONAL TEST. The 4 entire control red and its release mechanism, a CHANNEL CAllBRATION may be performed by any 1 containment isolation valve and its operator or a series of sequential, overlapping, or total i

safety injection pump and its prime mover. channel steps such that the entire channel is calibrated.

j 1.4 ACTUATION LOGIC TEST 1.8 CHANNEL CHECK J

An ACTUATION LOGIC TEST shall be the application A CHANNEL CHECK shall be the qualitative

! of various simulated input combinations in assessment of channel behavior during operation j . conjunction with each possible interlock logic by observation. This determination shall

state and verification of the required logic include, where possible, comparison of the i output. The ACTUATION LOGIC TEST shall include a channel indication and/or status with other

continuity check, as a minimum, of output devices. Indications and/or status derived from i independent INSTRUMENT CHANNELS measuring the l

same parameter.

i l Og150/0g100 1 l

l.0 DEFINITIONS 1.9 CHAN4EL FUNCTIONAL TEST CONTAINNENT INTEGRITY shall exist when: (Continued)

A CHANNEL FUNCTIONAL TESI shall be: b. Equipment hatch is closed.

a. Instruments - The injection of a simulated c. At least one door in each air lock is closed signal (s) into the channel as close to the and sealed.

primary sensor (s) as practicable to verify OPERABILITY, including all channel outputs, as d. Containment leakage satisfies Specification appropriate. 3.10.

b. Logics - The application of input signals, or e. Penetration pressurization systems are in the operation of relays or switch contacts, in service as required by Specification 3.9.2.

all the combinations required to produce the required decision outputs includtag the 1.12 CONTINUOUS RELEASE operation of all ACTUATION DEVICES. idhere practicable, the test shall include the A CONTINUOUS RELEASE is the discharge of 11guld operation of the ACTUATED EQUIPMENT as well or gaseous wastes of a nondiscrete volume; e.g.,

(i.e. pumps will be started, valves operated, from a volume or system that has an input flow etc.). during the release.

1.10 CORPOSITE SAMPLE 1.13 CONTROLLED LEAKAGE A CORPOSITE SAMPLE is one in which the quantity of CONTROLLED LEAKAGE shall be the seal water flow 11guld sample is proportional to the quantity of from the reactor coolant pump seals.

11guld waste discharged and in which the method of sampling employed results in a specimen which is 1.14 CORE ALTERATION representative of the liquids released.

CORE ALTERATION shall be the movement or 1.11 CONTAINRENT INTEGRITY shall exist when: manipulation of any component within the reactor pressure vessel with the vessel head removed and

a. All penetrations required to be closed during fuel in the vessel. Suspension of CORE accident conditions are either: ALTERATION shall not preclude completion of

1. Capable of being closed by an OPERABLE movement of a component to a safe conservative autonctic containment isolation valve position.

system, or

2. Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions.

09150/09180 2

1.9 MFINITIONS 1.15 KFleES TEARS - 1.19 _ GASEOUS RA8MASTE TREATMENT SYSTEM The MFINED TEWts of this section appear in A GASEOUS RA0MASTE TREATpENT SYSTEM is any capita 11 red type threeghout these Technical system designed and installed to reduce Specificattens. radteactive ga:eous effluents by collecting off. gases from the Reactor Coolant System and 1.16 MEREE OF RESUNDAACY providing for delay or holdup for the purpose of reducing the total radioactivity prior to M1th reference to redundant instrument er logic release to the environment.

channels, the MEREE OF RESUNDANCY 15 the difference between the number of OPERA 8tE channels 1.20 19ENTIFIED LEAKAGE and the minimum number of these channels which,

- when tripped, will cause an aetematic system trip. IMNTIFIED LEAKAGE shall be:

a. Leakage (except CONTROLLED LEAKAGE) into t 1.17 OSSE Etulv4 LEET I-131 closed systems, such as pump seal or valve packing leaks that are captured and 90SE ESWIWALEET I-131 shall be that concentration conducted to a sump or collecting tank, or of I-131 (microcerle/ gram) which alene would b. Leakage into the containment atmosphere from produce the same thyreld dose as the quantity and sources that are both specifically located 1setopic mistere of I-131.1-132.1-133.1-134 and known either not to interfere with the  !

and 1-135 actually present. The thyroid dose operatten of leakage detection systems or conversten facters used for this calculatten shall not to be PRESSURE 800NDARY LEAKAGE, or be these listed in Table III of TIS-14844 c. Reacter coolant system leakage through a "Calculatten of 915tante Facters for Power and steam generator ta the secondary systes.  ;

Yest teacter $1tes' or Table E-T of WRC Regulatory  ;

Emide 1.199 Bew. 1. dated October igTT.

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1.le f - AvtaAGE DISISTEERATles EM tGY i

[ shall be the average (mighted in propertion to  ;

the concentratten of each radienecilde in the sample) of the sum of the average beta and gamma energies per distetegratten (in ReV/d) fer 1seteses, other than ledines, with half lives ,

greater than 15 minutes, making up at least g51L of  !

the total non-ledine activity in the coolant. t 99158/99108 3

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1 1.0 DEFINITI045 l

1.21 INSTaunteT CneauEE 1.25 0FF-SITE AC POWER SOURCES An lasTRumENT CNeumEL shall be the combination of The DFF-SITE A.C. POWER SOURCES shall be the sensor (s), signal processing elements, output devices System Auxiliary Transformer-142 - (Unit 1); 242  :

and other components and circuitry as required to - (Unit 2). The System Aux 111ary Transformer measure and evaluate a process variable for the from the opposite unit (242 - Unit 1; 142 - Unit purpose of shservation, control and/or protection of 2).

the process system. A channel may produce both analog signal setputs and discrete (signal or 1.26 0FFSITE DOSE CALCULATION MANUAL (00CM) electromechanical cesponent operation) outputs. The i

i channel terminates and loses its identity where The DFFSITE DOSE CALCULATION MANUAL shall l 1sdividsal channel outputs are combined. contain the methodology and parameters used in j the calculation of offsite doses due to 1.22 LE AE AGE radioactive gaseous and 11guld effluents, in the i calculation of gaseous 11guld effluent Refer to the following specific paragraphs. monitoring alarm / trip setpoints, and in the 1.13 EceTROLLES LEAKAGE conduct of the Environmental Radiological 1.20 IDENTIFIES LEAKAGE Monitoring Program.

1.32 FRESSURE 80useaY LEAKAGE 1.a8 UNIDENTIFIED LEAKAGE 1.27 OPERA 8tE - OPERA 81LITY 1.23 MASTER SELAY TEST A system, subsystem, train, component or device i shall be OPERABLE or have OPERABILITY when it is A RESTER RELAY TEST shall be the energtration of each capable of performing its specified function (s),

anster relay and vertf1 cation of OPERA 81LITY of each and when all necessary attendant relay. The IWLSTER RELAY TEST shall include a instrumentation, controls, electrical power, 4 continuity check of each associated slave relay. Coollag or seal water, lubrication or other aux 111ary equipment that are required for the 4 1.24 NEngEs(510F TME PUBLIC system, subsystem, train, component, or device i

to perform its function (s) are also capable of

REMBER(S) 0F TME PUBLIC shall include all persons who peforming their related support function (s).

I are not occupationally associated with the plant.

This category does not include egloyees of the The OPERABILITY requirement for both normal and l emergency AC power supplies does not apply to AC

utility. Its contractors or its vendors. Also

]

escluded from this category are persons who enter the instrument busses and associated instruments.

i site to service egulpment or to make de11vertes.

This category does include persons who use portions

of the site for recreational, occupational or other l purposes not associated with the plant.

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1.0 DEFINIT! DES 1.29 OPERATING 1.33 PROCESS CDNTROL PROGRAM CP)

OPERATING 1s defined as performing the intended The PROCESS CDNTROL PROGRAM (PCP) shall contain fonction in the intended menner. the current formulas, sampling, analyses, tests and determinations to be made to ensure that the 1.29 OPERATIBG CYCLE processing and packaging of solid radioactive wastes will be accomplished in such a way as to The OPERATING CYCLE shall be the interval between assure compliance with 10 CFR parts 20, 61 and the end of one major refueling outage and the end 71, and Federal and State regulations and other of the next subsequent major refueling outage per requirements governing the shipment and disposal unit. of radioactive waste.

1.30 DPERATIDEAL N00E - N00E 1.34 PROTECTIDW LDGIC CHANNEL An OPERATIOmAL MODE (1.e. MODE) shall correspond A PROTECTION LOGIC CHANNEL shall be an to any one inclustve combination of core arrangement of relays, contacts or other reactivity condition, power level, and average components which operate in response to reactor coolant temperature specified in Table INSTRURENT CHANNEL outputs to produce a decision 1.1, when feel assemblies are present in the output. The decision output is the initiation reactor vessel. of a protective action signal. At the system level, the decision output is the operation of a 1.31 PHYSICS TESTS sufficient numer of ACTUATIDN DEVICES and the associated ACTUATED EQUIPMENT as required to PHYSICS TESTS shall be those tests performed to place or restore the Nuclear Steam Supply System measure the fundamental nuclear characteristics of to a design safe state. The channel is deemed the reactor core and related instrumentation and to include the ACTUATION DEVICES.

1) described in Chapter 14.0 of the FSAR, 2) authertzed under the provisions of 10 CFR 50.5g, 1.35 PROTECTIDN SYSTEM or 3) otherwise approved by the Cosumission.

The PROTECTION SYSTEM shall consist of both the 1.32 PRESSURE SOURSaRY LEAKAGE Reactor Protection System and the Engineered Safeguards System. The PROTECT!DN SYSTEM PRESSURE BOUe0ARY LEAKAGE shall be leakage (except encompasses all electric and sechanical devices steam generator tube leakage) through a and circuitry (from sensors through ACTUATIDN non-1solable fault in the Reactor Coolant System DEVICES) which are required to operate in order coneonent body, pipe wall, or vessel wall. to place or restore the Nuclear Steam Supply System to a design safe state.

0g150/0g180 $

1.0 DEFINITIONS 1.36 PURGE - PURGIBG 1.40 REFUELING CYCLE OR OUTAGE PURGE OR PURGING is the controlled process of When REFUELING CYCLE or DUTAGE is used to discharging air or gas from a confinement to designate a survelliance interval, the maintain tesserature, pressure, humidity, surveillance shall be performed at lease once concentration or other operating condition, in every 18 meaths as allowed by general such a manner that replacement air or gas is requirteent 4.0.2.

required to purify the confinement.

1.41 REPORTA8tE EVENT 1.37 00AeRANT POWER TILT RATIO A REPORTABLE EVENT shall be any of those QUADRANT POWER TILT RATIO shall be the ratto of conditions specified in Specification 6.6.2 or the maximum upper encore detector calibrated Section 50.73 of 10 CFR Part 50.

output to the average of the upper encore detector calibrated outputs, or the ratto of the maximum 1.42 SHUTOOWN MARGIN lower encore detector calibrated output to the average of lower excere detector calibrated SHUTDOWN MARGIN shall be the instantaneous outputs, whichever is greater. amount of reactivity by which the reactor is subcritical or would be subcritical from its 1.38 RATED THERNAL POWER present condition assuming all control and

! shutdown banks are fully inserted, except for RATED THERRAL POWER shall be a total steady state the single rod cluster assembly of highest reactor core heat transfer rate to the reactor reactivity worth which is assumed to be fully coolant of 3250 MWt. withdrawn.

1.3g REACTOR PRESSURE 1.43 SITE BOUNDARY The REACTOR PRESSURE shall be the pressure in the The SITE BOUNDARY shall be that line beyond steam space of the pressurizer. which the land is not owned, leased or otherwise l controlled by the licensee.

1.44 SOLIDIFICATION SOLIDIFICATION shall be the conversion of radioactive liquid, resin and sludge wastes from

! 11guld systems into a form that meets shipping l and burial site requirements.

09150/09180 6

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I 1.0 KFINITIONS I j 1.45 SOURCE CE CK 1.50 VENTILATION ERHAUST TREATMENT SYSTEM j A SSURCE CNECK shall be the qualitative assessment A VENTILATION ERNAUST TREATMENT SYSTEM is any

of channel response when the Channel sensor is system designed and installed to reduce gaseous

! exposed to a radioactive source. radiciodine or radioactive material in

! particulate form in effluents by passing j 1.46 SWRVEILLANCE FREOUENCY NOTATION ventilation or vent exhaust gases through

charcoal absorbers and/or MEPA filters for the ,

The SWRVEILLANCE FREQUENCY NOTATION specified for purpose of removing lodines or particulates from +

i the performance of Surveillance Requirements shall the gaseous exhaust stream prior to the release correspond to the intervals defined in Table 1.2. to the environment. Such a system is not and General Surve111ance Requirement 4.0.2. considered to have any effect on noble gas

. effluents. Engineered Safety Feature (ESF)

! 1.47 THERRAL POWER atmospheric cleanup systems are not considered l to be VENTILATION ERHAUST TREATMENT SYSTEM i

THERRAL POMER shall be the total reacter core heat components.

transfer rate to the reacter coolant.

1.51 VENTING 1.48 UNIENTIFIED LEAKAGE .

VENTING is the controlled process of discharging i UNIENTIFIED LEAKAGE shall be all leakage which is air or gas from a confinement to maintain j not IN NTIFIES LEAKAGE or CONTROLLES LEAKAGE. temperature, pressure, humidity, concentration j or other operating condition, in such a manner l

! 1.4g UNEESTRICTES AREA that replacement air or gas is not provided or ,

i required during venting. Vent, used in system

! An UNRESTRICTES AREA shall be any area at er names, does not imply a venting process.

i beyond the SITE SOUNGARY to which access is not

! controlled by the licensee for purposes of

! protection of individuals from exposure to

{ radiatten and radteactive materials or any area

within the site boundary used for residential 1 quarters or for industrial, commercial, j institutional and/or recreational purposes.

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_ _ _ . _ _ _ ~ . - _ _ _ _ - . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ , - _ _ _ _ _ _ - _ __ _ ._ ___

REACTIVITY FISSION COOLANT TEMPERATURE l MODE (a K/K) POWER # (Tave.)

1 POWER OPERATION 10 2% $ P $ 100% T oper 2 HOT STANDBY Z0 5 2% T oper 3 HOT SHUTDOWN Fig. 3.2-1 0 350*F < Tavg 5 Toper 4 HOT SHUTDOWN and Tavg 5 350'F Fig. 3.2-1 0 200*F < Tavg 5 350*F 5 COLD SHUTDOWN 5 -1% 0 5 200*F 6 REFUELING ** 5 -10% 0 5 140'F l .

7 LOW POWER PHYSICS TESTS *

• 5 5%

Where: Tavg - Average temperature across a reactor vessel as measured by the hot and cold leg temperature detectors.

Toper - Any temperature at which a reactor is critical, limited by Specifications 3.2.1.C.1 and 3.3.2.A.

NOTE: Where HOT SHUTDOWN is mentioned in any Specification it means either MODE 3 or 4 unless Tavg is specifically stated.

1. Excluding decay heat

• To be stated for specific tests.

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

l TABLE 1.1 OPERATIONAL MODES (See Definition 1.30) 09150/09180 6b

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NOTATION FREQUENCY 5 (Shiftly) At least once per scheduled shift D (Daily) At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

W (Weekly) At least once per 7 days.

M (Monthly) At least once per 31 days.

Q (Quarterly) At least once per 92 days.

SA (Semi-Annually) At least once per 184 days.

R (Refueling Cycle) At least once per 18 months.

S/U (Startup) Prior to reactor startup.

P (Prior) Complete prior to start of release EFPM At least once per effective full power month N.A. Not Applicable. .'

TABLE 1.2 SURVEILLANCE FREQUENCY NOTATIDN (See Definition 1.46) 09150/09180 6c

LIMITING CDMDITIDN FOR OPERATION SURVEILLANCE REQUIRENENT 3.0 GENERAL 4.0 GENERAL 3.0.1 Limiting Conditions for Operation (LCO's) 4.0.1 Surveillance Requirements shall be i

and ACTIDN requirements shall be applicable during the OPERATIONAL RODES applicable for each specification during or other conditions specified for the stated OPERATIONAL MODES or other individual Limiting Conditions for conditions. Operation unless otherwise stated in an l

individual Surveillance Requirement.

3.0.2 Adherence to the requirements of the 4.0.2 Each Surve111ance Requirement shall be Limiting Condition for Operation and/or performed within the specified time associated ACTION within the specified interval with:

time interval shall constitute compliance with the specification. The ACTION a. A maximum allowable extension not statement need not be completed if the to exceed 25% of the surveillance LCO is restored prior to expiration of interval, but

, the time interval.

b. The combined time interval for any ,

three consecutive surveillance 3.0.3 If a limiting Condition for Operation intervals shall not exceed 3.25 4

and/or associated ACTION requirements times the specified surveillance cannot be satisfied, action shall be interval.

initiated within one hour to place the unit in at least HOT SHUTDDWN within the NOTE: See Table 1.2 for an explicit following four hours, and in at least definition of frequency designations.

COLD SHUTDOWN within the following 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. This requirement need not be 4.0.3 Performance of a Surveillance Requirement completed if: within the specified time interval shall constitute compliance with OPERABILITY

a. The reactor is placed in a MODE in requirements for a Limiting Condition for which the specification is not Operation. If there are exceptions to the applicable; or OPERABILITY requirements, they are stated in

b. Corrective measures which permit the individual specifications. Surveillance operation are completed within the Requirements do not have to be performed on specified time interval as measured inoperable equipment.

from initial discovery; or

c. Exceptions are stated in the individual specifications.

09150/09180 27a

Reactor Trip Channel Channel Channel Channel Description Check Calibration [unctional Test Remarks

17. Low Steam Generator Level in Coincidence with Feed Flow Steam Flow Mismatch S R Q

18. Low-Low Steam Generator level S R Q

19. Safety Injection MA NA Q*

• Manual SI function check at R only

20. Turbine Trip. NA NA Q

21. Automatic Reactor Trip Logic NA NA M' ' Including testing of the reactor trip breaker shunt and undervoltage trip pechanisms independently.

PERMISSIVES

22. P-6 NA NA S/U18 2' Not required if performed within the previous thirty days

23. P-7 NA NA M

24. P-8 NA NA M

25. P-10 NA NA M TABLE NOTATION:

See Table 1.2 for definition of terms.

Reactor Protection System Testing and Calibration Requirements TABLE 4.1-1 (Sheet 2 of 2) 09150/09180 36

LIMITING CONDITIDN FOR OPERATION SURVEILLANCE REQUIREMENTS 3.9.4 C. When it is determined that one of the main 4.9.4 C. Not applicable steam isolation valves can be closed but does not respond to automatic actuation, either of the following actions shall be taken:

1. The loop isolation valves in the associated reactor coolant loop shall be closed, or

2. The reactor coolant pump shall be tripped, the main steam isolation valve shall be closed and the feedwater isolation valves shall be closed in the loop associated with the inoperable main steam isolation valve.

D. When it is determined that one of the main D. No applicable steam isolation valves cannot be closed, the loop isolation valves in the associated reactor coolar.t loop shall be closed and three loop operation shall commence (See section 3.3.1).

5. Containment Integrity 5. Containment Integrity A. The CONTAINMENT INTEGRITY shall not be A. Not applicable l

violated whenever a nuclear core is installed in the 0543t/0544t 201 0345A _

LIMITING CONDITION FOR OPERATION , SURVEILLANCE REQUIREMENTS 3.13. REFUELING OPERATIONS 4.13 REFUELING OPERATIONS

1. Core reactivity: 1. Core Reactivity:

A. Core reactivity shall be maintained within A. Surve111ance and testing shall be g control limits during CDRE ALTERATION by performed during CORE ALTERATION as l l 1mplementing the following: follows:

1. A SHUTDOWN MARGIN equal to or greater 1. Reactor coolant boron

• concentration than 10% AK/K shall be maintained and Tavg shall be verified once a using control rods.and/or coolant shift.

boron concentration for reactivity control. Tavg shall be maintained at or less than 140*F.

2. Subcritical neutron flux shall be 2. The operability of the neutron continuously monitored by two source monitor (s) shall be verified once a i range neutron monitors, (permanent shift. Temporary neutron monitors and/or temporary), each with shall be checked with a source prior continuous visual indication and one to installation.

with audible indication, whenever core geometry is being changed. At other 1 times, when fuel is in the reactor neutron flux shall be monitored

continuously by at.least one source range neutron monitor with both visual and audible indication.

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LIMITING CONDITION FOR OPERAT10N SURVEILLANCE REQUIREMENTS 3.13.1. A (Continued 7 4.13.1. A (Continued)

3. The movement of an irradiated fuel 3. Not applicable.

assembly in the reactor core shall not begin until the reactor has been subcritical for a period of at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />.

, 4. A least one RHR pump and heat 4. The operation of at least one RHR exchanger shall be in operation during pump and heat exchanger shall be core alteration operations. verified once a shift.

5. Direct communication between the 5. Communication between the control

i control room and containment shall be room and the containment shall be l OPERABLE. verified before any alteration of the reactor core begins.

6. A licensed fuel handling foreman or 6. Not applicable.

licensed senior reactor operator shall 4

be present at the reactor cavity l during any movement of fuel within the l containment.

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LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENTS t

3.13.2. Protection from Damaged Spent Fuel 4.13.2. Protection from Damaged Spent Fuel A. During irradiated fuel movement or crane A. The charcoal adsorber mode of operation operation with loads over irradiated fuel of the fuel building exhaust system in the fuel building, the fuel building shall be demonstrated to be operable:

exhaust system shall be:

1. Operating with ventilation flow 1. Observe and document shiftly that t

through the HEPA and charcoal the ventilation system is OPERATING filters if there is any irradiated as required by Specification

! fuel stored in the pool with less 3.13.2.A.

] than 60 days decay time. .

2. OPERABLE with automatic initiation 2. When operability is required by l

of flow through the HEPA filters Specification 3.13.2.A.2, the and charcoal adsorbers upon following shall be done at least detection of high radiation at the once per 31 days:

fuel pool if all irradiated fuel stored in'the pool has 60 days or a. Place the Fuel Building greater decay time since Ventilation System in the Fuel irradiation ceased. If automatic Handling Mode for a minimum of actuation is inoperable, the 15 minutes.

system shall be manually placed in the " charcoal adsorber mode".

b Verify flow through the HEPA and APPLICABILITY: All Modes charcoal adsorber train.

ACTION: With the requirements of 3.13.2. A not satisfied, suspend all irradiated fuel c. Verify the Fuel Building is movements or crane operation hith maintained at 1/4 inch of water loads over irradiated fuel after negative pressure with respect first, if applicable, placing loads in to the atmosphere.

a safe condition.

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LIMITING CONDITION FOR OPERATIDN SURVEILLANCE REQUIREMENTS l

3.13.2 8. Ventilation filters for the fuel building 4.13.2 8. For each HEPA or charcoal filter, at including charcoal adsorbers and the least once per 18 months or (1) af ter automatic actuation of the charcoal filter every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber system shall be periodically tested. operation or (2) after any structural maintenance of the filter housings or (3) following painting, fire,or chemical a release ~in any ventilation zone communicating with the system, or (4) after each complete or partial replacement of the filter bank, surveillance will be performed per Table 4.17.

1. Verify that on a high radiation test signal the system automatically starts (unless already in operation) and directs its exhaust flow through the HEPA filters and charcoal adsorber banks. If automatic actuation is inoperable the system shall be manually placed in the charcoal adsorber mode.

09150/09180 245

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENTS 3.13.3 Containment status 4.13.3 Containment status A. During CORE ALTERATION, CONTAINMENT A. Containment door status shall be INTEGRITY shall be maintained as specified verified once a shift.

in section 3.9.5 except as specified in 3.13.3.8.

B. The equipment hatch or both doors on the B. Reactor coolant boron concentration and personnel hatch may be opened during the Tavg shall be verified once a shift when CORE ALTERATION provided the shutdown the equipment hatch is open or both

' l- margin is maintained equal to or greater doors on the personnel hatch are open.

than 10% AK/K and Tavg maintained at or less than 140*F.

l C. During CORE ALTERATION, the containment C. The containment vent and purge system vent and purge system and the radiation and the radiation monitors which monitors which initiate isolation of this initiate isolation of this system shall I system, shall be OPERABLE. be tested and verified to be OPERABLE .

immediately prior to CORE ALTERATION operations.

I l

i 09150/09180 246

= - - - = . ,,erv.--ww=wwe-+,,.-,y- -g-----,,-+vw w ey,y.---g*ee---w-wwww+----- ,,'w+,,,,w,--. e%.wer-----,<--,4,,-, r- -- - -, ..--~,-e- - - - , - -

• LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENTS 3.13 (Continued) 4.13. (Continued)

4. R_adiation Monitoring: 4. Radiation Monitoring:

The radiation monitoring system for the fuel The radiation monitoring system for the fuel building and the refueling cavity shall be in building and refueling cavity shall be

! continuous operation during CORE ALTERATION. tested once a shift during CORE ALTERATION. l

5. Refuelina Equipment OPERABILITY: 5. Not Applicable.

The fuel transfer system and manipulator crane operability shall be verified. All interlocks shall be checked and a load test equivalent to the weight of a fuel assembly shall be nede prior to' refueling.

6. If any of the specified limiting conditions 6. Not Applicable.

for refueling are not net, CORE ALTERATIONS l shall cease until the specified limits are met. No operations which could increase the reactivity of the core, other than Tavg changes, shall be nede untti the specified limits are met.

7. At least one of the spent fuel pit cooling 7. At least one of the two spent fuel pit l

systems trains shall be OPERABLE. cooling systems trains shall be tested and verified to be OPERABLE immediately prior to the CORE ALTERATION.

t l

09150/09180 246a l

's 1

i 6.1.7.A.

6.1.7.A.1 (Continued)

2. Audit function (f) Significant operating abnormalities or deviations from normal and expected- The Audit function shall be the performance of plant equipment that responsibility of the Manager of Quality affect nuclear safety as referred to it Assurance independent of the Production by the Onstte Review and Investigative Department. Such responsibt'11ty is function. delegated to the Operations Quality Assurance Manager, to the Director of (g) Reportable Events requiring 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Quality Assurance Operations and to the notification to the Commission. 01 rector of Quality Assurance Maintenance.

(h) All recognized indications of an Either of the above, or designated Corporate unanticipated deficiency in some aspect Staff or Supervision approved by the Manager of design or operation of safety of Quality Assurance, shall approve the related structures, systems or audit agenda and checklists, the findings components. and the report of each audit. Audits shall be performed in accordance with the Company (1) Review and report findings and Quality Assurance Program and Procedures.

- recommendations regarding all changes ' Audits shall be performed to assure that

! to the Generating Stations Emergency safety-related functions are covered within Plan prior to implementation of such a period of two years or as designated below:

change.

(a) Audit of the conformance of facility (j)' Review and report findings and operation to provisions contained recommendations regarding all items within the Technical Specifications and referred by the Technical Staff applicable license conditions at least Supervisor, Station Manager, Division once per year.

Vice President and General Manager -

Nuclear Stations or Manager of Quality (b) Audit of the adherence to procedures, Assurance. training and qualification of the station staff at least once per year.

(k) Changes to utisite Dose Calculation j Manual (DDCM). (c) Audit of the results of actions taken to correct deficiencies occurring in '

j (1) Changes to the PROCESS CONTROL PROGRAM. facility equipment, structures, systems

or methods of operation that affect nuclear safety at least once per six months.

i

} 0615t/0616t 302

0266A

s i

6.1.7.A.2 (Continued) (m) Report all findings of noncompliance with NRC requirements and (d) Audit of the performance of activities recommendations and results of each required by the Quality Assurance audit to the Station Manager, the Program to meet the Criteria of 10 CFR Division Vice-President and General 50 Appendix "B".. Manager Nuclear Stations, Manager of I Quality Assurance', Vice-President (e) Audit of the Generating Stations (Nuclear Operations), and Manager of Emergency Plan and Implementing Nuclear Safety.

Procedures.

(n) Audit the Offsite Dose Calrulation (f) Audit of the Fac111ty Security Plan and Manual at least once per 24 months.

Implementing procedures.

6.1.7.A.3. Authority (g) Audit Onsite and Offsite Reviews.

The Manager of Quality Assurance (h) Audit the Radiological Environmental reports to the Chairman & President and Monitoring Program at least once per 12 the Supervisor of the Offsite Review months, and Investigative function reports to the Manager of Nuclear Safety who (1) Audit the Facility Fire Protection reports to the Chairman and President.

Program and implementing procedures at Either the Manager of Quality Assurance

, least once per 24 months. or the Supervisor of the Offsite Review and Investigative Function has the (j) An independent fire protection and loss authority to order unit shutdown or prevention program inspection and audit request any other action which he deems shall be performed at least once per 12 necessary to avoid unsafe plant

/ months utilizing either qua11 fled conditions.

offsite licensed personnel or an outside fire protection firm. 4. Records (k) An inspection and audit of the fire (a) Reviews, audits and recommendations protection and loss prevention program shall be documented and distributed as shall be performed by a qualifted covered in 6.1.7.A.1 and 6.1.1.A.2 outside fire consultant at least once per 36 months. (b) Copies of documentation, reports, and correspondence shall be kept on file at (1) The PROCESS CONTROL PROGRAM and the station.

implementing procedures at least once per 24 months.

l 0615t/0616t 303 j 0266A

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