ML20207J013

Proposed Radiological Effluent Tech Specs

ML20207J013
Person / Time
Site:	Zion  File:ZionSolutions icon.png
Issue date:	12/16/1986
From:	COMMONWEALTH EDISON CO.
To:
Shared Package
ML20207H957	List: ML20207H951 ML20207J013
References
2493K, NUDOCS 8701080255
Download: ML20207J013 (32)

Text

l ATTACISBNT 1 PROPOSED CHANGES TO APPENDIX A RADIOLOGICAL EFFLUENT TECHNICAL SPECIFICATIONS (RETS)

PAGES MODIFIED:

iv 232a vii 235 viii 236a x

236b xi 237 2

238 5

241 6

241a 223a 251a 223b 253 227 253a 227b 279b 231 322 232 0612N95 Ot00$gg 01010 2493K PDE P

SURVEILLANCE I

LINITING CONDITION FOR OPERATION REQUIREMENT PAGE

~

3.13 Refueling Operations 4.13 243 l

3.13.1 Core Reactivity 4.13.1 243 3.13.2 Protection from Damaged Spent Fuel 4.13.2 244a 3.13.3 Containment Status 4.13.3 246 3.13.4 Radiation Monitoring 4.13.4 246a 3.13.5 Refueling Equipment Operability 4.13.5 246a 3.13.6 Refueling LCO not Met 4.13.6 246a 3.13.7 Spent Fuel Pit Cooling Systems 4.13.7 246a l

3.13.8 Fuel Inspection Program 4.13.8 247 3.13.9 Residual Heat Removal System Operation 4.13.9 247a l

3.13.10 Water Level - Reactor Vessel 247c 3.13.11 Water Level - Storage Pool 4.13.11 247d Bases 248 3.14 Plant Radiation Monitoring 4.14 250 Bases 254 3.15 Auxiliary Electrical Power System 4.15 255 Bases 270 3.16 Radiological Environmental Monitoring 4.16 275 Bases 280c l

3.17.1 Ventilation 4.17.1 281 3.17.2 Aircraft Fire Detection 4.17.2 283 Bases 287 3.18 Steam Generator Activity 4.18 289 Bases 290 l

3.19 Failed Fuel Monitoring 4.19 292 Bases 293 3.20 Radioactive Solids 4.20 295a Bases 295b l

3.21 Fire Protection 4.21 2950 Bases 2950 3.22 Shock Suppressors (Snubbers) 4.22 295W Bases 295Z 3.22.1 Mechanical Snubbers 4.22.1 295W j

3.22.2 Hydraulic Snubbers 4.22.2 295AA1 1

3.22.3 Snubber Service Life Monitoring 4.22.3 295AA7 i

Bases 295AA8 3.23 Special Test Exceptions 4.23 295A6 Bases 295AC TABLE OF CONTENTS (Continued) iv 1

1150t/1151t

1 Fiqure h

3. 3. 2-1 Reactor Coolant System Heatup Limitations 84 l

3.3.2-2 Reactor Coolant System Cooldown Limitations 85 l

3.3.2-3 Effect of Fluence and Copper Content on Shift of 86 l

RTNDT for Reactor Vessel Steels Exposed to 550 degrees F i

Temperature 3.3.2-4 Fluence at 1/4T and 3/4T as a Function of Full Power 87 Service Years 3.3.6-1 Dose Equivalent I-131 RC Limit versus Percent of Rated Thermal Power 124c j

3.4-1 High Steam Line Flow Setpoint 131a 3.11-1 Restricted Area Boundry 226 l

l

[

6.1-1 Minimum Shift Crew Composition 327 l

i l

i i

LIST OF FIGURES (Continued) v11 1150t/1151t i

I

Page Table 6b 1.1 Operational Modes 6c 1.2 Surveillance Frequency Notation Reactor Protection System-Limiting Operations Conditions and Setpoints 3G 3.1 -1 33 t

3.1-2 Reactor Protection System Instrument Numbers 88

3. 3. 2 -1 RTNDT Testing Results 106 i

3. 3. 4 -1 In Service Inspection Program i

122 l

3.3.5-1 Reactor Coolant Systems and Chemistry Specifications 129 Engineered Safeguards Actuation System-Limiting Conditions on 3.4-1 Operation and Setpoints 132 i

3.4-2 Engineered Safeguards System Instrument Numbers l

Neutron Flux High Trip Points with Steam Generator Safety Valves 160a 3.7-1 l

Inoperable - Four Loop Operation l

160b Neutron Flux High Trip Points with Steam Generator Safety Valves

}

3.7-2 Inoperable - Three Loop Operation i

i 188 l

l

3. 8.9 -1 Accident Monitoring Instrumentation Maximum Permissible Concentration of Dissolved or Entrained Noble 226a 3.11-1 Gases Releases from the Site to Unrestricted Areas in Liquid Effluents 228 Radioactive liquid Effluent Monitoring Instrumentation 3.11-2 236 Radioactive Gaseous Effluent Monitoring Instrumentation 3.12-1 251 i

3.14-1 Radiation Monitoring Instrumentation 268 3.15-1 Equipment Requirement with Inoperative 4KV E.S.S. Sus 269 3.15-2 Equipment Inoperable with Inoperative 4KV E.S.S. Bus LIST OF TA8LES

.444

I Table Page f

i l

4.8-5 Component Cooling Pump System 189 l

4.8-6 Service Water Pump System 190 I

l

4. 8-7 Hydrogen Control System 192 4. 8. 9-1 Accident Monitoring Instrumentation Surveillance Requirements 189 4.9-1 Isolation Seal Water System 203 l

i 4.9-2 Penetration Pressurization System 204 i

4.9-3 Containment Isolation valves 205 4.9-4 Main Steam Isolation Valves 208 j

4.11-1 Radioactive Liquid Effluent Sampling and Analysis 227 surveillance i

4.11-2 Radioactive Liquid Effluent Monitoring Instrumentation 228b Surveillance 4.12-1 Radioactive Gaseous Effluent Sampling and Analysis Program 238 4.12-2 Radioactive Gaseous Effluent Monitoring Instrumentation 240 j

Surveillance 1

4.14-1 Plant Radiation Monitoring Instrumentation Surveillance 253 4.15-1 4160-Volt Engineered Safeguard Bus Main, Reserve, and 270 Standby Feeds l

4.16-1 Maximum Valves for the Lower Limits of Detection (LLD) 280 l

l List of Tables (Continued)

X 1160t/1161t

Table Page i

4.17-1 Charcoal Filters 284 4.17-2 HEPA Filters 285 4.19-1 Failed Fuel Monitoring Instruments 295 4.21-1 Fire Protection Instruments 295p 4.21-2 Fire Suppression Water System 295r 4.21-3 Sprinkler Systems 295s 4.21-4 CO2 Systems 295t 4.21-5 Fire Hose Stations 295u l

6.6-2 Special Reports 323 l

6.8-1 Soundary Doors for Flooding Protection 328 e

i j

t l

l I

LIST OF TA8LES (Continued)

I x1 1150t/1151t i

9 1.0 DEFINITIONS 1.9 CHAPNEL FUNCTIONAL TEST 1.11 CONTAINMENT INTEGRITY (Continued) 1 A CHANNEL FUNCTIONAL TEST shall be:

2.

Closed by manual valves, blind flanges, or deactivated automatic valves secured a.

Instruments - The injection of a simulated in their closed positions.

signal (s) into the channel as close to the primary sensor (s) as practicable to verify b.

Equipment hatch is closed.

OPERA 8ILITY, including all channel outputs, as appropriate.

c.

At least one door in each air lock is closed and sealed.

b.

Logics - The application of input signals, or l

the operation of relays or switch contacts, in d.

Containment leakage satisfies Specification j

all the combinations required to produce the 3.10.

required decision outputs including the operation of all ACTUATION DEVICES. Where e.

Penetration pressurization systems are in l

practicable, the test shall include the service as required by Specification 3.9.2.

1 operation of the ACTUATED EQUIPMENT as well (i.e. pumps will be started, valves operated, 1.12 CONTIN'JOUS RELEASE l

etc.).

1.10 COMPOSITE SAMPLE A CONTINUOUS RELEASE is the discharge of liquid or gaseous wastes of a nondiscrete volume; e.g.,

from a volume or system that has an input flow A COMPOSITE SAMPLE is one in which the quantity of during the release.

liquid sample is proportional to the quantity of l

liquid waste discharged and in which the method of 1.13 CONTROLLED LEAKAGE i

sampling employed results in a specimen which is representative of the 11gulds released.

CONTROLLED LEAKAGE shall be the seal water flow from the reactor coolant pump seals.

1.11 CONTAINMENT INTEGRITY t

1.14 CORE ALTERATION CONTAINMENT INTEGRITY shall exist when:

CORE ALTERATION shall be the movement or a.

All penetrations required to be closed during manipulation of any component within the reactor accident conditions are either:

pressure vessel with the vessel head removed and fuel in the vessel. Suspension of CORE 1.

Capable of being closed by an OPERABLE ALTERATION shall not preclude completion of automatic containment isolation valve movement of a component to a safe conservative system, or position.

1150t/1151t 2

i l

i 1.0 DEFINITIONS i !

1.28 OPERATING 1.33 PROCESS CONTROL PROGRAM (PCP)

~

j OPERATING is defined as performing the intended The PROCESS CONTROL PROGRAM (PCP) shall contain f

j function in the intended manner, the current formulas, sampling, analyses, tests 4

and determinations to be made to ensure that the i

fj 1.29 OPERATING CYCLE processing and packaging of solid radioactive l

wastes will be accomplished in such a way as to 1

i 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 I

unit.

of radioactive waste.

i 1.30 OPERATIONAL MODE - MODE 1.34 PROTECTION LOGIC CHANNEL An OPERATIONAL MODE (i.e. MODE) shall correspond A PROTECTION LOGIC CHANNEL shall be an to any one inclusive combination of core arrangement of relays, contacts or other reactivity condition, power level, and average components which operate in response to i

reactor coolant temperature specified in Table INSTRUMENT CHANNEL outputs to produce a decision 1.1, when fuel 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 outp't is the operation of a u

1.31 PHYSICS TESTS sufficient numer of ACTUATION 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 l

the reactor core and related instrumentation as 1) to include the ACTUATION DEVICES.

described in Chapter 14.0 of the FSAR, 2) authorized under the provisions of 10 CFR 50.59, 1.35 PROTECTION SYSTEM or 3) otherwise approved by the Commission.

i The PROTECTION SYSTEM shall consist of both the l.32 PRESSURE BOUNDARY LEAKAGE Reactor Protection System and the Engineered Safeguards System. The PROTECTION SYSTEM PRESSURE BOUNDARY LEAKAGE shall be leakage (except encompasses all electric and mechanical devices steam generator tube leakage) through a and circuitry (from sensors through ACTUATION non-isolable fault in the Reactor Coolant System DEVICES) which are required to operate in order component body, pipe wall, or vessel wall.

to place or restore the Nuclear Steam Supply System to a design safe state.

r 11 CM /11 C1 +

C

j

[

j j 1.0 DEFINITIONS 1.36 PURGE - PURGING 1.40 REFUELING CYCLE OR OUTAGE l

PURGE or PURGING is the controlled process of When REFUELING CYCLE or OUTAGE is used to

]1

}

discharging air or gas from a confinement to designate a surveillance interval, the maintain temperature, pressure, humidity, surveillance shall be performed at least once

!' i concentration or other operating condition, in every 18 months as allowed by general

)i such a manner that replacement air or gas is requirement 4.0.2.

l 1 required to purify the confinement.

jl 1.41 REPORTA8LE EVENT i

2 1.37 OUADRANT POWER TILT RATIO A REPORTA8tE EVENT shall be any of those

• I QUADRANT POWER TILT RATIO shall be the ratio of conditions specified in Specification 6.6.2 or

{ll the maximum upper excore detector calibrated Section 50.73 of 10 CFR Part 50.

output to the average of the upper excore detector

- {

calibrated outputs, or the ratio of the maximum 1.42 SHUTDOWN MARGIN g

lower excore detector calibrated output to the 1

average of lower excore detector calibrated SHUTOOWN MARGIN shall be the instantaneous l

outputs, whichever is greater.

amount of reactivity by which the reactor is 3

subcritical or would be subcritical from its l

1.38 RATED THERMAL POWER present condition assuming all control and lll shutdown banks are fully

• inserted, except for RATED THERMAL POWER shall be a total steady state the single rod cluster assembly of highest i i reactor core heat transfer rate to the reactor reactivity worth which is assumed to be fully coolant of 3250 MWt.

withdrawn.

1.39 REACTOR PRESSURE 1.43 SITE BOUNDARY l

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 i

controlled by the licensee.

1.44 SOLIDIFICATION b

I i

SOLIDIFICATION shall be the conversion of

}

radioactive liquid, resin and sludge wastes from j

11guld systems into a form that meets shipping i

and burial site requirements.

{

1150t/1151t 6

+

.. _ - - _ - - = -.

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.11 Radioactive Liquids (continued) 4.11 Radioactive Liquids (Continued) 2.

Dose 2.

Dose A.

The dose or dose commitment to a A.

Dose Calculations Cumulative dose l

MEMBER OF THE PUBLIC above contributions from liquid effluents background from radioactive shall be determined by calculation at materials in liquid effluents least once per month and a cumulative released from the site to summation of these total body and any UNRESTRICTED AREAS (see Figure organ doses shall be maintained for 3.11-1) shall be limited:

each calendar quarter.

s 1)

During any calendar quarter to less than or equal to 3 mree to the total body and to less than or equal to 10 mrem to any organ, and 2)

During any calendar year to less than or equal to 6 mrem to the total body and to less than or equal to 20 mrem to any organ.

O APPLICA8ILITY:

At all times l

1 l

,,,n.,,,,,.

i i

i i

LINITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT i

i l

l 3.11.2 (Continued) 4.11.2 (Continued) l ACTION:

a)

With the calculated dose from the release of radioactive materiils in liquid effluents r

exceeding twice the limits specified in 3.11.2.A.1 limit the subsequent releases such that the dose or dose commitment to a l

MEN 8ER OF THE PG8LIC from all uranium fuel cycle sources is limited to 125 mres to the total body or any organ (except thyroid, which is limited to 5 75 area) over 12 consecutive months. Demonstrate that radiation exposures to all MEMBERS OF THE PUBLIC from all uranium fuel cycle sources (including all offluent pathways and direct radiation) are less than the 40 CFR Part 190 and 40 CFR Part 141 Standard, otherwise obtain a variance from the Commission to i

permit releases which exceed the 40 CFR Part

)

141 or 190 Standard. The radiation exposure analysis shall use methods prescribed in the 00CM.

I b)

The provisions of specifications 3.0.3 and r

i 3.0.4 are not applicable.

I i

i

]

n snt m sit em

, l l l Minimum (a) l I

~

Liquid Sampling Analysis Type of Lower Limit of Release Type Frequency Frequency Activity Analysis Detection (LLD)(uci/ml) ll A.

Batch Releases Prior To Each Prior To Each Principal Gaauna SE-7 Tanks (c)

Release Release Emitters (el l

'i I-131 1E-6 I

P Dissolved and entrained ll Lake Discharge One Batch /M M

Gases (Gamma emitters)(e) l 1E-5 j

Tank j

P M

Tritium 1E-5 i= l Each Batch Composite (b)

Gross Alpha 1E-7 I

1 P

Q Sr-89, Sr-90 SE-8

!]l Each Batch Composite (b)

Fe-55 1E-6 Weste Neutralizing Prior to' each Prior to each Principal gamma emitter (e)

SE-7 l

l, Tan;:

release release I-131 1E-6 P

M Tritium 1E-5 Each Batch Composite (b) j I

Gross Alpha, IE-7 i

8.

Continuous Principal Gamma

{

Releases (d)

Emittersfe)

SE-7 Continuous W

I-131 1E-6 During Releases Dissolved and entrained gases 1E-5 l

(Gaaru Emitters)

)

l Turbine Building n

ar m um it-3 l

Fire Sump (f)

Continuous Composite (b) i Gross Alpha 1E-7

~

~

y sr-us, sr-su st-a l

Continuous Composite (b)

Fe-55 1E-6 4

l Radioactive Liquid Effluent Sampling & Analysis Surveillance Table 4.11-1 1150t/1151t 227

1 1

l t

TA8LE NOTATION (Continued) jl For certain mixtures of gamma emitters, it may not be possible to measure radionuclides in j

concentrations near their sensitivity limits when other nuclides are present in the sample in much

)

greater concentrations. Under these circumstances, it will be more appropriate to calculate the concentrations of such radionuclides using observed ratios with those radionuclides which are measurable.

l j

b.

A COMPOSITE SAMPLE is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen which is representative of the liquids released.

6 1) i To be representative of the quantities and concentrations of radioactive materials in liquid I

effluents, all samples taken for the composite shall be throughly mixed in order for the comoosite sample to be representative of the effluent release.

l 2)

The weekly and monthly Proportional Composite samples are not required provided that (1) the analysis required for each of these conoosite samples has been run on each batch discharged, l

and (2) a monthly record of radionuclides discharged (isotope and quantity) is maintained.

c.

A BATCH RELEASE is the discharge of liquid wastes of a discrete volume. Prior to sampling for analyses, each batch shall be isolated, and then thoroughly mixed to assure r'epresentative sampling.

d.

A CONTINUOUS RELEASE is the discharge of liquid wastes of a nondiscrete volume;/e.g., from a volume of system that has an input flow during the continuous release.

The principal genna emitters for which the LLO specification applies exclusively are the following e.

radionuclides: Mn-54, Fe-59, Co-58, Co-60, 2n-65, Mo-99, Cs-134. Cs-137. Co-141, and Ce-144. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported.

Nuclides which are below the LLO for the analyses shall be reported as "less than" the nuclide's LLD, and shall not be reported as being present at the LLD level for that nuclide. The "less than" values shall not be used in the required dose calculations.

f)

If the fire sump composite sampler is inoperable grab samples will be taken from the turbine building fire sump once per shift.

Radioactive Liquid Effluent Sampling & Analysis Surveillance Table 4.11-1 Continued 1150t/1151t

???h

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.12 Gaseous Effluents (Continued) 4.12.

Gaseous Effluents (Continued) 2.

Dose 2.

Dose A.

Noble Gas - The air dose due to noble A.

Cumulative dose contributions for the gases released in gaseous effluents current calendar quarter and current f rom the site (see Figure 3.11-1) calendar year shall be determined in shall be limited to the following:

accordance with the 00CM at least once every 31 days for noble gas and 1.

During any calendar quarter:

Less radioiodines, radioactive materials in than or equal to 5 mrad for gamma particulate form and radionuclides (other radiation and less than or equal than noble gas) with half-lives greater to 10 mrad far beta radiation and, than eight days.

2.

During any calendar year:

Less than or equal to 10 mrad for gamma j

radiation and less than or equal to 20 mrad for beta radiation.

I B.

Radiciodine - Particulate - Other Than Noble Gas l

The dose to a MEMBER OF THE PUBLIC from radiciodines and radioactive materials in particulate form, and radionuclides (other than noble gases) with half-lives greater than 8 days in gaseous effluents released from the site (see Figure 3.11-1) shall be limited to the following:

1.

During any calendar quarter:

Less than or equal to 7.5 mrem to any organ and, 2.

During any calendar year:

Less i

than or equal to 15 mrem to any organ.

1157t/1158t 231

[

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.12.2 Gaseous Ef fluents (Continued) 4.12.2 Gaseous Effluents (Continued)

APPLICABILITY: At all times ACTION:

a.

With the calculated air dose f rom gaseous effluents exceeding the above limits, define the corrective action (s) to be taken to ensure that future releases are in compliance with 3.12.2.

b.

With the calculated air dose from radioactive noble gases in gaseous effluents exceeding twice the limits of Specification 3.12.2.4.1 1.

Limit subsequent releases such that the dose or dose j

commitment to a MEMBER OF THE PUBLIC from all uranium fuel cycle sources is limited to less than or equal to 25 mrem i

to the total body or any organ (except thyroid, which is limited to 75 mrem) over 12 i

consecutive months.

2.

Prepare an analysis which demonstrates that radiation exposures to all MEMBERS OF TliE PUBLIC from all uranium fuel cycle sources (including all effluent pathways and direct radiation) are less l

than the 40 CFR Part 190 Standard.

Il57t/ll50t 232

l l

I LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT

~

\\

3.12.2 Gaseous Effluents (Continued) 4.12.2 Gaseous Effluents (Continued) c.

With the calculated dose from the release of Iodine-131. Iodine-133, j

tritium, and all radionuclides in I

particulate form with half-lives greater than 8 days in gaseous effluents exceeding twice the limits of Specification 3.12.2.B.1 1.

Limit subsequent releases such i

that the dose or dose commitment l

l to a MEMBER OF THE PU6LIC from all i

uranium to less than or equal to i

25 mrem to the total body or organ 1

(except the thyroid which is j

limited to less than or equal to 75 mrem) over 12 consecutive months.

t i

2.

Prepare an analysis which demonstrates that radiation exposures to all MEM6ERS OF THE 1

PUBLIC from all uranium fuel cycle

}

sources (including all effluent pathways and direct radiation) are less than the 40 CFR Part 190 1

Standard. Otherwise, request a i

variance from the Commission to permit releases which exceed the 40 CFR Part 190 Standard. The radiation exposure analysis shall use the methods prescribed in the ODCM.

l d.

The provisions of Speck ications 3.0.3 and 3.0.4 are not app 1'. cable.

t 115st/1158t 232a l

I i

j LIMITING CON 0!! ION FOR OPERATION SURVLILLANCE REQUIREMENT l

3.12 Gaseous Effluents (Continued) 4.12 Gaseous Effluents (Continued)

5. Explosive Gas Mix 1.ure 5.

Explosive Gas Mixture l

A.

The concentration of hydrogen or oxygen in A.

The concentrations of hydrogen or l

the waste gas systema shall be limited to oxygen in the weste gas system shall less than or equal to 35 by volume.

be determined to be within limits, at i

I least once per day.

APPLICAPILITY: At all times ACTION:

a.

With the' concentration of hydrogen and oxygen in the waste gas system each greater than 3% by volume but either hydrogen or oxygen less than or equal to 4% by volume, initiate action within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> to restore the concentration of hydrogen or oxygen to within the limit.

i b.

With the concentration of hydrogen and oxygen in the waste gas system each greater than 4% by volume, initiate action within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to reduce the concentration of I

hydrogen or oxygen to less than or equal to 35.

c.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

i The waste gas system consists of the following:

l l

Gas Decay Tank when on fill and the Hold-up Tanks.

i i

ll Ent /ll El f 714 i

Minimum Channels Applicable Instrument Operable Action Modes 4.

Auxiliary 8u11 ding Ventilation and Miscellaneous ventilation Steck A.

Gas Activity Monitor 1.

OR-0014 or 1

6 All 2.

1R-PR25 and 2R-PR25 1

6 All 3.

OR-PR188 Gas 1

6 All 4.

1R-PR49E (Channel 5) 1 6

All 5.

2R-PR49E (Channel 5) 1 6

All 8.

Iodine Monitor l

1.

OR-PR128 1

8 All 2.

1R-PR49C (Channel 3) 1 8

All 3.

2R-PR49C (Channel 3) 1 8

All C.

Particulate Monitor l

1.

OR-PR12A 1

8 All 2.

OR-PRI8A Particulate 1

6 All 3.

1R-PR49A (Channel 1) 1 8

All 4.

2R-PR49A (Channel 1) 1 8

All D.

Flow Rate Monitor 1.

1LP-084 1

9 All 2.

2LP-084 1

9 All E.

Particulate / Iodine Sampler 1.

10PR038 1

8 All 2.

20PR038 1

8 All 5.

Service 8u11 dine Ventilation A.

Gas Activity Monitor 1.

OR-PR22 1

8 All l

B.

Particulate / Iodine Sampler 1.

OR-PR36 1

8 All Radioactive Gaseous Effluent Monitor Instrumentation (Continued)

Table 3.12-1 (Continued) 736a iisnt n init

Minimum j

Channels Applicable i

Instrument Doerable Action Modes 6.

Steam Generator AtmosDheric Relief and Safety Valves 1.

1R-PR58 1

10 1,2,3,7 2.

2R-PR58 1

10 1,2,3,7 3.

1R-PR59 1

10 1,2,3,7 4.

2R-PR59 1

10 1,2,3,7 5.

1R-PR60 1

10 1,2,3,7 6.

2R-PR60 1

10 1,2,3,7 7.

1R-PR61 1

10 1,2,3,7 8.

2R-PR61 1

10 1,2,3,7 7.

Accident Monitoring A.

Containment 1.

1R-PR406 (Channel 7) 1 10 1,2,3,4,7 2.

2R-PR40G (Channel 7) 1 10 1,2,3,4,7 i

3.

1R-PR401 (Channel 9) 1 10 1,2,3,4,7 4.

2R-PR40I (Channel 9) 1 10 1,2,3,4,7 l

B.

Miscellaneous Vent Stack i

1.

1R-PR496 (Channel 7) 1 10 1,2,3,4,7 2.

2R-Pc496 (Channel 7) 1 10 1,2,3,4,7 3.

1R-PR491 (Channel 9) 1 10 1,2,3,4,7 4.

2R-PR491 (Channel 9) 1 10 1,2,3,4,7 C.

Containment Fuel Handling Area Monitor 1.

1R-AR04A 1

11 6

When purging during 2.

1R-AR04B 1

11 6

fuel handling 3.

2R-AR04A 1

11 6

operations l

4.

2R-AR048 1

11 6

Radioactive Gaseous Effluent Monitor Instrumentation (Continued)

Table 3.12-1 (Continued) l l

l 1160t/1161t 236b

l Table N:tation I

ACTION 5 - With the number of channels OPERABLE less than the minimum number required, the contents of the tank may be released to the environment provided that prior to initiating the release:

1.

At least two independent samples of the tank's content are analyzed, and i

8 2.

At least two technically qualified members of the facility staff independently verify the release rate calculations and discharge flow path valving; otherwise, suspend release of radioactive effluents via this pathway.

,h l

8 ACTION 6 - With the number of channels OPERABLE less than the minimum number required, effluent releases via this i

pathway may continue for up to 30 days provided grab samples are taken at least once per shif t and these l

samples are analyzed for gross activity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

l l

ACTION 7 - With the number of channels OPERABLE less than the minimum number required, and no redundant monitor

i OPERABLE in this flow path, 1sumediately suspend PURGING of radioactive effluents via this pathway.

ii I'

ACTION 8 - With the number of channels OPERA 8LE less than the minimum number required, effluent releases via this pathway may continue for up to 30 days provided samples are continuously collected with auxiliary sampling j

]

equipment as required in Table 4.12-1.

ACTION 9 - With the number of OPERA 8LE channels less than the minimum number required, effluent releases via this i

pathway may continue provided the flow rate is estimated at least once per shift wh'11e release is in i

progress.

ACTION 10 - With the number of channels OPERA 8LE less than the minimum number required, restore the inoperable monitor to OPERABLE status within 30 days or establish an alternate means of monitoring the parameter.

ACTION 11 - With the number of OPERABLE channels less than the minimum number required, suspend vent and purge i

operations and close each vent and purge valve providing direct access from the containment atmosphere to the outside atmosphere or suspend the movement of nuclear fuel and reactor components in the vicinity of the reactor, refueling cavity, and transfer canal (containment side).

ACTION 12 - With the number of OPERABLE channels less than the minimum number required, effluent releases via this pathway may continue provided the effluent flow is being accounted fer in the total plant effluent.

Radioactive Gaseous Effluent Monitor Instrumentation (Continued) i TABLE NOTATION Table 3.12-1 11 CO+ /1141 +

117

I l

)

1 I,

4 Minimum

}

Gaseous Sampling Analysis Type of Lower Limit of (e) l Release Type Frequency Frequency Activity Analysis Detection (LLD) (pC1/cc){

l A.

Gas Decay Tank Grab Sample Prior to Each Noble Gases 1E-4 I

Prior to Each Release (c)

Principal Ganna Emitters l

l Release (d) l

)

]

Continuous Sample After Each

'/ articulate 1E-11 l

During Each Release (c)

Principal Ganna Emitters 1

Release (d) i Tritium 1E-6 I-131 (Charcoal Sample) 1E-12 I-133 (Charcoal Sample) 1 E-10 Sr-89 Particulate 1E-11 Composite Quarterly (c)

Sr-90 Particulate 1E-11 Gross Alpha 1E-11 8.

Containment Vent Prior to Each Prior to Each Principal Gaseous Ganna 1E-4 and Purge Release (a)

Release (c)

Emitters (d)

Particulate Gamma Emitters

,l 1E-11 j

(d) l Tritium 1E-6 I-131 (Charcoal) 1E-12 l

f l

1-133 (Charcoal) 1 E-10 Sr-89 Particulate 1E-11 l

l Composite Quarterly (c)

Sr-90 Particulate 1E-11 f

Gross Alpha 1E-11 I

l Radioactive Gaseous Effluent Sampling and Analysis Program Table 4.12-1 1

1 i

n.,,,,,.

Charnel Channel Source Channel Functional Instrument Check Check Calibration (1)

Test (2) 4.

Auxiliary Building Ventilation and Miscellaneous ventilation Stack A.

Gas Activity Monitor 1.

OR-0014 Gas or D

M R

Q 2.

1RT-PR25 and 2RT-PR25 D

M R

Q 3.

OR-PR188 D

M R

Q 4.

1R-PR49E (Channel 5)

D M

R Q

5.

2R-PR49E (Channel 5)

D M

R Q

B.

Icdine Monitor l

1.

OR-PR128 D

M R

Q 2.

1R-PR49C (Channel 3)

D M

R Q

3.

2R-PR49C (Channel 3)

D M

R Q

C.

Particulate Monitor l

1.

OR-PR12A D

M R

Q 2.

OR-PR18A D

M R

Q 3.

1R-PR49A (Channel 1)

D M

R Q

4.

2R-PR49A (Channel 1)

D M

R.

Q D.

Flow Rate Monitor 1.

1LP-084 0

N/A R

Q 2.

2LP-084 D

N/A R

Q l

E.

Particulate / Iodine Sampler i

1.

10PR038 N/A N/A N/A N/A 2.

20PR038 5.

Service Building Ventilation A.

Gas Activity Monitor 1.

OR-PR22 D

M R

Q B.

Particulate / lodine Sampler 1.

OR-PR36 N/A N/A N/A N/A i

l Radioactive Gaseous Effluent Monitor Instrumentation Surveillance (Continued)

Table 4.12-2 1160t/1161t 241 i

Channol Channel Source Channel Functional Instrument Check Check Calibration (1)

Test (2) k 6.

Steam Generator, Atmospheric Relief and Safety Valves 1.

1R-PR58 0

M R

Q 2.

2R-PR58 0

M R

Q 3.

1R-PR59 D

M R

Q 4.

2R-PR59 D

M R

Q 5.

1R-PR60 D

M R

Q 6.

2R-PR60 D

M R

Q

?.

1R-PR61 0

M R

Q 8.

2;2-PR61 D

M R

Q 7.

Accident Monitorina A.

Containment 1.

1R-PR40G (Channel 7)

N/A N/A R

Q i

2.

2R-PR40G (Channel 7)

N/A N/A R

0 3.

1R-PR401 (Channel 9)

N/A N/A R

Q 4.

2R-PR401 (Channel 9)

N/A N/A R

Q 8.

Miscellaneous Vent Stack 1.

1R-PR49G (Channel 7)

N/A N/A R

Q 2.

2R-PR49G (Channel 7)

N/A N/A R

Q 3.

1R-PR491 (Channel 9)

N/A N/A R

Q 4.

2R-PR491 (Channel 9)

N/A N/A R

Q C.

Fuel Handling Area 1.

1R-AR04A D

M(3)

R Q(4)

Q(4) 4)

2.

1R-AR048 0

M(3)

R Q(

3.

2R-AR04A D

M(3)

R 4.

2R-AR04B D

M(3)

R Q(4)

Radioactive Gaseous Effluent Monitor Instrumentation Surveillance (Continued)

Table 4.12-2 (Continued) 1160t/1161t 241a

Minimum Channels Applicable Instrument ODerable Action #

Modes i

F.

Auxiliary Building Area 1.

OR-AR04 1

24 All 2.

OR-AR08 1

24 All 3.

OR-AR09 1

24 All 4.

OR-ARIO 1

24 All 5.

OR-AR11 1

24 All 6.

OR-0006 1

24 All 2.

Process Monitors A.

Containment 1.

Reactor Leak Detection a.

1R-PR12A 1

28 1,2,3,7 b.

1R-PR128 1

28 1,2,3,7 c.

2R-PR12A 1

28 1,2,3,7 d.

2R-PR128 1

28 1,2,3,7 2.

Ventilation c.

1R-PR40A (Channel 1) 1 25 1,2,3,7 c

b.

1P-PR40C (Channel 3) 1 25 1,2,3,7 c.

1R-PR40E (Channel 5) 1 25 1,2,3,7 d.

2R-PR40A (Channel 1) 1 25 1,2,3,7 e.

2R-PR40C (Channel 3) 1 25 1,2,3,7 f.

2R-PR40E (Channel 5) 1 25 1,2,3,7 8.

Component Cooling 4

1.

OR-PR07 1

26 All 2.

1R-0017 1

26 All l

3.

2R-0017 1

26 All RADIATION MONITORING INSTRUMENTATION TA8LE 3.14-1 (Continued) 1160t/1161t 251a

}

I l

l Channel l

Channel 50 s e Channel Functional Instrument Check QLt4 Calibration (1)

Test (2)

I l

1.

Area Monitors l

A.

Fuel Storage Pool Area 1)

OR-0005 D

M3 R

2)

OR-AR03 D

M3 R

Q 3)

OR-AR13 D

M3 g

g B.

Containment Purge Isolatior.

{

1) 1R-AR04A D

M4 R

QS 2) 1R-AR048 D

M4 R

QS 3) 2R-AR04A D

M4 R

QS 4) 2R-AR048 D

M4 R

QS C.-

Containment Area (High Range) i 1) 1R-AR02 D

N/A R

Q 2) 2R-AR02 0

N/A R

Q 3) 1R-AR03 D

N/A R

'Q 4) 2R-AR03 D

N/A R

Q D.

Control Room

./)

1)

OR-0001 D

M R

Q E.

Technical Support Center I

1)

Portable Area Monitor D

M R

Q i

l l

PLANT RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS l

TABLE 4.14-1 1150t/1151t 253

i Channel Channel Source Channel Fur.ctional Instrument Check Check Calibration (1)

Test

'(2)

F.

Auxiliary Building Area 1)

OR-AR04 0

M R

Q 2)

OR-AR08 D

M R

Q 3)

OR-AR09 0

M R

Q 4)

OR-ARIO D

M R

Q 5)

OR-AR11 D

M R

Q 6)

OR-0006 D

M R

Q 2.

Process Monitors A.

Containment 1.

Reactor Leak Detection a) 1R-PR12A D

N/A R

Q b) 1R-PR128 D

N/A R

Q c) 2R-PR12A D

N/A R

Q d) 2R-PR12B D

N/A R

Q f

2.

Ventilation a) 1R-PR40A (Channel 1)

D M

R Q

b) 1R-PR40C (Channel 3)

D M

R Q

c) 1R-PR40E (Channel 5)

D M

R Q

d) 2R-PR40A (Channel 1)

D M

R Q

l e) 2R-PR40C (Channel 3)

D M

R Q

f) 2R-PR40E (Channel 5)

D M

R Q

B.

Component Cooling 1.

OR-PR07 D

M R

Q 2.

1R-0017 D

M R

Q

[

3.

2R-0017 D

M R

Q 4

Plant Radiation Monitoring Instrumentation Surveillance Requirements (Continued) i i

TABLE 4.14-1 (Continued) 1160t/1161t 253a

Exposure Pathway and/or Samole Media Collection Site Tvoe of Analysis Frecuency e

4.

Ingestion (Continued) l C.

Food Products Samples of three different kinds Gamma isotopic and I-131 Monthly when I

of broad leaf vegetation grown analysis.

available.

nearest each of two different I

off-site locations of highest l

predicted annual average ground-level D/Q if milk sampling is not perfonned.

One sample of each of the similiar Gamma isotopic and I-131 Monthly when l

broad leaf vegetation grown 15 to analysis.

available.

30 km distant in the least prevalent wind direction if milk sampling is not performed.

I Footnotes:

l f

1.

81-Weekly shall mean at the frequency of once every other week.

A gamma isotopic analysis shall be performed whenever the gross beta concentration in a sample exceeds by 2.

five times (5x) the average concentration of the preceeding calendar quarter for the sample location.

Zion Standard Radiological Environmental Monitoring Program Table 3.16-1 (Continued) 1150t/1151t 279b l

{C

6.6.2.8.1.a (Coc.tinued)

(IX)

Reports submitted to the Commission in accordance with paragraph (VIII) above Ci

>l j

also meet the effluent release RL.1 i

reporting requirements of 10CFR Part 20 th paragraph 20.405(a)(5) where C is the concentration of the i 1^

radionuclides in the medium and RL is the (X)

Any event that posed an actual threat reporting level of radionuclide 1.

to the safety of the plant or significantly hampered site personnel (2)

If radionuclides other than those in Table in the performance of duties necessary 3.16-2 are detected and are due to plant l

for the safe operation of the nuclear effluents, a reporting level is exceed if power plant including fires, toxic gas the potential annual dose of an individual releases, or radioactive releases.

is equal to or greater than the design objective doses of 10 CFR 50, Appendix 1.

6.6.3 Uniaue Reportina Reauirements (3) This report shall include an evaluation of any release conditions, environmental A.

Non-Routine Reports:

factors, or other aspects necessary to explain the anomalous effect.

4 (1) Environmental Radiological Monitoring l

Program l

If a confirmed measured radionuc'ide I

concentration in an environmental sampling medium averaged over any

)

calendar quarter sampling period i

exceeds the reporting level given in l

Table 3.16-2 and if the radioactivity is attributable to plant operation, a

)

written report shall be submitted to the Director of the NRC Regional Office 1

of Inspection and Enforcement with a i

copy to the Director, Office of Nuclear i

Reactor Regulation within 30 days from

)

the end of the quarter. When more than one of the radionuclides in Table l

3.16-2 are detected in the medium, the reporting level shall have been exceeded if:

i

}

ll60t/1161t 322 I

ATTACHMNT 2 DESCRIPTION OF PROPOSED CHANGES PAGES:

iv, vii, viii, x, xi The changes to these pages all provide minor corrections to the Table of Contents, List of Figures, or the List of Tables.

PAGES: 2, 5, 6, 223a, 223b, 231, 232, 232a These changes all provide clarification to either the definition or usage'of specific terms utilized within the Zion Technical Specifications.

There is no proposed change to the meaning of any definition or Itaiting condition for operation.

PAGES:

227, 227b These pages provide clarification to Table 4.11-1, Radioactive Liquid Effluent Sampling and Analysis Surveillance. There is no functional change to this guidance.

PAGE: 235 This change deletes the word 3 "cr cover gas" from the description of the waste gas system. When a gas decay tank is on cover gas, the holdup tank sample will monitor the concentration of the cover gas system. Thus, t

there is no change to the intent of Section 3.12.5.

PAGES: 236a, 241, 251a, 253a i

i These changes include the addition of auxiliary building ventilation rad monitors OR-PR 12A and B, and 2R-0017. These monitors were inadvertently omitted from the original Technical Specification.

PAGES: 236b, 237, 238, 241a, 279b, 322 These changes all involve minor typographical corrections to the existing Technical Specifications. There is no functional change proposed.

PhGE: 253 l

This change involves the deletion of a monthly source check requirement erroneously listed for the containment high range rad monitors.

i These monitors do not have installed check sources.

l 2493K

,m,

..r._,

-v-..

t,

,_.r._

e.,

7

ATTACHIENT 3 DESCRIPTION OF PROPOSED CHANGES i

EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATION PROPOSED CHANGES TO ZION TECHNICAL SPECIFICATION APPENDIX A i

DESCRIPTION OF AMENDMENT REQUEST An amendment to the Zion Facility Operating License is proposed to provide clarifications and correct typographical errors contained in the Zion Radiological Effluent Technical Specifications (RETS).

BACKGROUND 10 CPR 50.92 states that a proposed amendment will involve a no significant hazards consideration if the proposed amendment does not:

(1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.

In addition, the commission has provided guidance in the practical applica-tion of these criteria by publishing eight examples in 48 FR 14870.

The discussion below addresses each of these three criteria and demonstrates that the proposed amendment involves a no significant hazards consideration.

BASIS FOR NO SIGNIFICANT HAZARDS CONSIDERATION DETEkMINATION

\\

Does the proposed amendment i

(1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or

~...

O

. (3) Involve a significant reduction in a margin of safety?

DISCUSSION ITEM #1 The proposed change involves clarifications to the existing RETS and the correction of minor typographical errors. None of the proposed changes involves the alteration of any definition or of any limiting condition for operation.

None of these administrative clarifications will affect any of Zion's systems or the integrity of any of its structures.

In addition, these clarifications will not affect the probability of occurrence of an external event such as a tornado or an earthquake.

Since these clarification will not involve any system interaction, will not alter Zion's operation, or affect the performance of Zion's structures, the pre-existing safety analysis will remain valid. Thus, this proposed change will not involve a significant increase in the probability or consequences of any accident previously evaluated.

ITEM #2 As discussed above, these administrative clarifications will not affect the performance of any of Zion's systems or structures.

They are administrative clarifications or corrections of minor errors.

These clarifications could not conceivably have any significant effect on Zion's operation during either normal or abnormal conditions. The specific accident sequences contained in Zion safety Analysis have been reviewed. Based upon the lack of interaction discussed above, none of these sequences will be altered by this proposed change.

Thus, the proposed amendments will not create the possibility of a new or different kind of accident from any accident previously evaluated.

ITEM #3 These clarifications and corrections will not affect the performance of any of Zion's systems or structures. Thus, all of Zion's systems will continue to perform their intended functions.

Since all of Zion's individual components and systems will continue to perform their intended safety function, there can be no change in the plant's overall performance during normal and abnormal operations. Thus, this proposed change will not involve a significant reduction in the margin of safety.

. This proposed change consists of administrative clarifications and minor corrections. Thus, example (i) is applicable in this instance.

Example (i) reads as follows:

(i) A purely administrative change to technical specifications: for example, a change to achieve consistency throughout the technical specifications, correction of an error, or a change in nomenclature.

Therefore, since the application for amendment satisfies the criteria specified in 10CFR 50.92 and is similar to examples for which no significant hazards consideration exists, Commonwealth Edison Company has made a determination that the application involves no significant hazards consideration.

l 1

l I

2493K I

.n.

.