ML20045F328

Amends 146 & 134 to Licenses DPR-39 & DPR-48,respectively, Modifying TS in Accordance W/Gl 89-01, Implementation of Programmatic Controls for RETS in Administrative Controls Section of TS & Relocation of Procedural Details..

ML20045F328
Person / Time
Site:	Zion  File:ZionSolutions icon.png
Issue date:	06/29/1993
From:	Dyer J Office of Nuclear Reactor Regulation
To:
Shared Package
ML20045F307	List: ML20045F304 ML20045F328 ML20045F331
References
GL-89-01, GL-89-1, NUDOCS 9307070186
Download: ML20045F328 (43)

Text

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COMMONWEALTH EDISON COMPANY DOCKET NO. 50-295 ZION NUCLEAR POWER STATION. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.146 License No. DPR-39 1.

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

A.

The application for amendment by Commonwealth Edison Company (the licensee) dated March 11, 1993, as supplemented June 21, 1993, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; i

C.

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

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment.

and paragraph 2.C.(2) of Facility Operating License No. DPR-39 is hereby amended to read as follows:

9 9307070186 930629 DR ADOCK 0500 S

' (2)

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 146, are hereby incorporated in the license. The licensee shall operate the facility in l

accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance and is to be implemented within 30 days of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION aw t. O ee f

ames E. Dyer, Director Project Directorate III-2 Division of Reactor Projects - III/IV/V Office of Nuclear Reactor Regulation Attachment-Changes to the Technical Specifications Date of Issuance:

June 29, 1993 i

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UNITED STATES i~

f NUCLEAR REGULATORY COMMISSION g-WASHINGTON, D.C, 20555-0001 COMMONWEALTH EDISON COMPANY DOCKET NO. 50-304 ZION NUCLEAR POWER STATION. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 134 License No. DPR-48 1.

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

A.

The application for amendment by Commonwealth Edison Company (the licensee) dated March 11, 1993, as supplemented June 21, 1993, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter 1:

B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

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

D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 LFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-48 is hereby t

amended to read as follows:

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(2)

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 134, are hereby incorporated in the i

license.

The licensee shall operate the facility in accordance with the Technical Specifications.

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3.

This license amendment is effective as of the date of its issuance and is to be implemented within 30 days of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION i

h.

M

' mes E. Dyer, Director Project Directorate III-2 Division of Reactor Projects - III/IV/V Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

June 29, 1993 f

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ATTACHMENT TO LICENSE AMENDMENTS AMENDMENT NO. 146 TO FACILITY OPERATING LICENSE NO. DPR-39 i

AMENDMENT NO. 134 TO FACILITY OPERATING LICENSE NO. OPR-48 DOCKET NOS. 50-295 AND 50-304 l

Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages.

The revised pages are identified by the captioned amendment number and contain marginal lines i

indicating the area of change.

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Remove Paaes Insert Paaes i

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vii vii ix ix i

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223a l

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225a 225a 226 226a

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279c 280 280a 280b 280c 275-280 295a 295a 295b 295b 302 302 310a 310a 310b 311a 311a 313 313 314 314 315 315' 316 316 322 322 323 323 325 325 326 326 i

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l TABLE OF CONTENTS 1.0 DEFINITIONS

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1.26 Offstte Dose Calculation Manual Page 1 1.1 Act1on 1.2 Actuation Device 1.27 Operable - Operability i

1.3 Actuated Equipment I

1.4 Actuation Logic Test Page 5 1.28 Operating 1.5 Axial Flux Difference 1.29 Operating Cycle l

1.6 DELETED 1.30 Operational Mode - Mode 1.7 Channel Calibration, Instrument 1.31 Physics Tests 1.8 Channel Check 1.32 Pressure Boundary leakage 1.33 Process Control Program (PCP)

Page 2 1.9 Channel Functional Test 1.34 Protection Logic Channel l

1.10 DELETED 1.35 Protection System 1.11 Containment Integrity l

1.12 DELETED Page 6 1.36 Purge - Purging 1.13 Controlled leakage 1.37 Quadrant Power Tilt Ratio 1.14 Core Alteration 1.38 Rated Thermal Power Page 2a 1.14a Core Operating Limits Report 1.39 Reactor Pressure 1.40 Refueling Outage Page 3 1.15 Defined Terms 1.41 Reportable Event 1.16 Degree of Redundancy 1.42 Shutdown Margin 1.17 Dose Equivalent I-131 1.43 Site Boundary 1.18 E - Average Disintegration Energy 1.44 DELETED l

l 1.19 DELETED 1.20 Identified Leakage Page 6a 1.45 Source Check 1.46 Surveillance Frequency Notation Page 4, 1.21 Instrument Channel 1,47 Thermal Power 1.22 Leakage 1.48 Unidentified Leakage 1.23 Master Relay Test 1.49 Unrestricted Area 1.24 Member (s) of the Public 1.50 DELETED l

1.25 Off-Site AC Power 1.51 Venting Sources LIMITING SAFETY SYSTEM SETPOINT SAFETY LIMITS 2.1 Protective Instrumentation Setpoints 1.1 Reactor Core Bases 2.2 Protective Equipment Setpoints 1.2 Reactor Coolant System Pressure Bases 1

Amendment Nos.146 and 134

SURVEILLANCE REQVIREMENT PAGE LI M HI NG_COND ElQN _F_0R_0_P_E R A T IO N 4.7 156 3.7 Steam Generator Emergency Heat Removal 4.7.1 156 3.7.1 Steam Generator Safety Valves 4.7.2 158 3.7.2 Auxiliary Feedwater Pump System 4.7.3 159a 3.7.3 Aux 111ary Feedwater Supply System 162 Bases 4.8 164 3.8 Emergency Core Cooling and Core Cooling Support 4.8.1 164 3.8.1 Centrifugal Charging Pump System 4.8.2 168 3.8.2 Safety Injection Pump System 4.8.3 170 3.8.3 Residual Heat Removal Pump System 4.8.4 173 3.8.4 System Testing of Centrifugal Charging. Safety Injection, and Residual Heat Removal Pump Systems 4.8.5 174 3.8.5 Accumulator System 4.8.6 175 3.8.6 Component Cooling System 4.8.7 178 3.8.7 Service Water System 4.8.8 180 3.8.8 Hydrogen Control Systems 4.8.9 184 3.8.9 Accident Monitoring Instrumentation 193 Bases 4.9 197 3.9 Containment Isolation Systems 4.9.1 197 3.9.1 Isolation Valve Seal Hater System 4.9.2 198 3.9.2 Penetration Pressurization Systems 4.9.3 199a 3.9.3 Containment Isolation Valves 4.9.4 200 3.9.4 Main Steam Isolation Valves and Bypasses 4.9.5 201 3.9.5 Containment Integrity 4.9.6 208a 3.9.6 Containment Ventilation System 209 Bases 4.10 212 3.10 Containment Structural Integrity 4.10.1 212 3.10.1 Containment Leak. age Rate Testing 4.10.2 214a 3.10.2 Containment Air Locks 4.10.3 215 3.10.3 Containment Tendons 4.10.4 217 3.10.4 End Anchorage and Concrete 4.10.5 219 3.10.5 Containment Pressure 4.10.6 219 3.10.6 Containment Temperature 220 Bases 4.11 225a 3.11 Radioactive Liquids 4.11.1 225a

'3.11.1 Outdoor Liquid Tanks 229a Bases 4.12 224 3.12 Gaseous Effluents 4.12.1 234 3.12.1 Gas Decay Tanks 4.12.2 235 3.12.2 Explosive Gas H1xture 242a Bases TABLE OF CONTENTS (Continued) ill Amendment Nos.146 and 134

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SURVEILLANCE LIMILNG_CQNRill0tLEOR_QEERATION REQQI3EMENT PAGE 3.13 Refueling Operations 4.13 243 3.13.1 Core Reactivity 4.13.1 243 3.13.2 Protection from Damaged Spent Fuel 4 13.2 244 3.13.3 Containment Status 4.13.3 245 3.13.4 Radiation Monitoring 4.13.4 246 3.13.5 Refueling Equipment Checkout 4.13.5 246 3.13.6 Refueling Equipment Operability 4.13.6 246 3.13.7 Spent Fuel Pit Cooling Systems 4.13.7 246 3.13.3 Fuel Inspection Program 4.13.8 247 3.13.9 Residual Heat Removal 4.13.9 247a r

System Operation 3.13.10 Hater Level - Reactor Vessel 4.13.10 247c 3.13.11 Hater level - Storage Pool 4.13.11 247d 3.13.13 Spent Fuel Pool Storage 4.13.13 247d 3.13.14 Spent Fuel Storage Pool Boron Concentration 4.13.14 247e Bases 248 3.14 Plant Radiation Monitoring 4.14 250 Bases 254 3.15 Auxillary Electrical Power System 4.15 255 Bases 271 3.16 (Section Deleted) 275 l

3.17.1 Ventilation 4.17.1 281 3.17.2 (Section Deleted) 283 Bases 287 3.18 Steam Generator Activity 4.18 289 Bases 290 3.19 Failed Fuel Monitoring 4.19 292 Bases 293 3.20 (Section Deleted) 4.20 295a l

3.21 Fire Protection 4.21 295D Bases 2950 l

3.22 Shock Suppressors (Snubbers) 4.22 295H Bases 295Z 3.23 (Section Deleted) 2;5AB TABLE OF-CONTENTS (Centinued)

IV Amendment Nos. 146 and 134

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296 5.0 Desian Features 296 5.1 Site 296 5.2 Reactor Coolant System 296 I

5.3 Reactor Core i-297 5.4 Containment System 298 5.5 Fuel Storage 299 5.6 Seismic Design 300 6.0 ADMINISTRATIVE CONTROLS 300 6.1 Organization,' Review, Investigation and Audit....................

309 6.2 Procedures and Programs...............................

1 l

6.3 Actions to be Taken in the event of a Reportable 310b l

Event in Plant Operation..............................

j 6.4 Action to be Taken in the Event A Safety Limit 1s Exceeded

-311 311 6.5 Plant Operating Records...............................

312 6.6 Reporting Requirements 325 6.7 Offsite Dose Calculation Manual (00CN).................... -..

326 6.8 Flooding Protection

..... '326 l

6.9 Process Control Program (PCP)

TABLE OF CONTENTS (Continued)

Amendment Mos.146 and 134 v:

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ElgER Ease 3.3.2-1 Reactor Coolant System Heatup Limitations 84 3.3.2-2 Reactor Coolant System Cooldown Limitations 85 3.3.2-3 Fast Neutron Fluence (E > 1 MeV) as a Function of 86 Full Power Service Life (EFPY) for Zion Unit 1 3.3.2-4 Fast Neutron Fluence (E > 1 HeV) as a Function of 87 Full Power Service Life (EFPY) for Zion Unit 2 3.3.6-1 Dose Equivalent I-131 RC Limit versus Percent of Rated Thermal Power 124c Deleted 131a 3.11-1 Deleted

.226 l

3.13.13-1 Fuel Assembly Burnup Limits in Region 2 247f 5.5.2-1 Spent Fuel Storage Pool 299a 6.1-1 Minimum Shift Crew Composition 327 LIST OF FIGURES (Continued) vil Amendment Nos. 146 and 134

PA22 lahle 208 3.9-4 Deleted 226a 1

3.11-1 Deleted l

228 3.11-2 Deleted 236 3.12-1 Deleted 251 3.14-1 Radiation Monitoring Instrumentation 268 3.15-1 Equipment Requirement with Inoperative 4KV E.S.S. Bus 269 3.15-2 Equipment Inoperable with Inoperative 4KV E.S.S. Bus 279 3.16-1 Deleted 279c 3.16-2 Deleted 35 4.1-1 Reactor Protection System Testing and Calibration Requirements 741 4.3.B-1 Minimum Number of Steam Generators to be Inspected During Inservice Inspection 74) 4.3.B-2 Steam Generator Tube Inspection 124b 4.3.6-1 Primary Coolant Specific Activity Sample and Analysis Program 134 4.4-1 Engineered Safeguards System Testing and Calibration Requirements 136 4.4-2 Deleted 148 4.5-1 Deleted 153 4.6-1 Deleted 160 4.7-1 Steam Generator Safety Valves, Set Pressures Orifice Sizes and Steam Flows LIST OF TABLES (Continued) lx Amendment Nos.146 and 134

=

_ =. _ _ - _ _ - _ _ - _.

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EA91 lable 161 4.7-2 Deleted 161a

-4.7-3 Deleted 185 4.8-1 Deleted i

186 4.8-2 Deleted 187 4.8-3 Deleted 188 4.8-4 Deleted 189 4.8-5 Deleted 190 4.8-6 Deleted 192 4.8-7 Deleted 192b 4.8.9-1 Accident Monitoring Instrumentation Surveillance Requirements 203' 4.9-1 Deleted t

204 4.9-2 Deleted 227 4.11-1 Deleted 228b 4.11-2 Deleted 238 4.12-1 Deleted 240 4.12-2 Deleted 4.14-1 Plant Radiation Monitoring Instrumentation Surveillance 253

-LIST OF 1AOLES (Continued)

Amendment' Mos.146 and 134 -

x

Eage Table 270 4.15-1 Deleted 270a 4.15-2 Diesel Generator Test Schedule 280 l

4.16-1 Deleted 284 4.17-1 Charcoal Filters 285 4.17-2 HEPA Filters 295 4.19-1 Deleted 295p 4.21-1 Fire Protection Instruments 295r 4.21-2 Deleted 295s 4.21-3 Sprinkler Systems 295t 4.21-4 002 Systems 295u 4.21-5 Fire Hose Stations 323 6.6-1 Special Reports 328 6.8-1 Boundary Doors for Flood Conditions LIST OF TABLES (Continued) x1 Amendment Nos.146 and 134

1.0 QEff_IMJ10NS 1.1 ACTLQN 1.5 All_AL_F1VX 0IfLERENCE (AI)

ACTION shall be that part of the Specification which AXIAL FLUX DIFFERENCE (AI) shall be the difference prescribes remedial measures required under designated in normalized flux signals between the top and conditions.

bottom Falves of a two section excore neutron detector 1.2 ACELA110LQEVICI 1.6 DELETED An ACTUATION DEVICE shall be a component or assembly of comporants that directly controls the motive power (ele',tricity, compressed air, etc.) for ACTUATED EQUiPHENT. The following are examples of an ACTUATION DEVICE:

circuit breaker, relay, and a valve (and its operator) used to control compressed air to the operator of a containment isolation valve.

1.3 &CIUMED_EDULPEEE 1.7 Cl! ANN _ELCA11BRATIQN __InstrumeJLt a

ACTUATED EQUIPMENT sSall be a component or assembly of A CHANNEL CALIBRATION shall be-the adjustment, as components that performs or directly contributes to the necessary, of the channel such that it responds performance of a protective function such as a reactor with the necessary range and accuracy to known trip, containment isolation, or emergency coolant values of input. The CHANNEL CALIBRATION shall injection.

The following are examples of ACTUATED encompass the entire channel including the sensors i

EQUIPMENT: an entire control rod and its releast (where possible), alarm interlock and/or trip mechanism, a containment isolation valve and its functions and shall include the CHANNEL FUNCTIONAL operator or a safety injection pump and its prime mover.

TEST. The CHANNEL CALIBRATION may be performed by l

any series of sequential, overlapping, or total l

channel steps such that the entire channel is 1.4 &CIUMLQtLLQGLC_lESI calibrated.

An ACTUATION LOGIC IEST shall be the application of 1.8 CHANNEL _ CHECK various simulated input combinations in conjunction with each possible interlock logic state and A CHANNEL CHECK shall be the qualitative assessment verification of the required logic output.

The of channel behavior during operation by ACTUATION LOGIC TEST shall include a continuity check, observation.

This determination shall include, as a minimum, of output devices.

where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent INSTRUMENT I

l CHANNELS measuring the same parameter.

1 Amendment Nos. 146.and-134

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1.0 DEEINITIONS 1.9 CHANNEL FUNCTIONAL TEST COMIAIEMERLili1EGRIIY shall exist when: (Cont'nued)

A CHANNEL FUNCTIONAL TEST shall be:

1.

Capable of being closed by an OPERABLE automatic containment isolation valve a.

Instruments - The injection of a simulated

system, signal (s) into the channel as close to the or primary sensor (s) as practicable to verify 2.

Closed by manual valves, blind flanges, or OPERABILITY, including all channel outputs, as deactivated automatic valves secured in approgrlate.

their closed positions, except for valves that are open under administrative control b.

Logics - The application of input signals, or as permitted by Specification 3.9.3.

the operation of relays or switch contacts, in all the combinations required to produce the b.

Equipment hatch is closed.

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

At least one door in each air lock is closed practicable, the test shall include the and se& led.

operation of the ACTUATED EQUIPMENT as well (i.e. pumps will be started, valves operated, d.

Containment leakage satisfies Specification etc.).

3.10.

c.

Digital channels - The injection of a e.

Penetration pressurization systems are in simulated signal (s) into the channel as close service as required by Specification 3.9.2.

to the sensor input to the process racks as practicable to verify OPERABILITY including 1.12 DELETED l

alarm and/or trip functions.

1.13 CQRIROLLED LEAKAGE 1.10 DELETED CONTROLLED LEAKAGE shall be the seal water flow from 1.11 CONTAINMENT INTEGRITY shall exist when:

the reactor coolant pump seals.

a.

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

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

2 Amendment Nos. 146 and 134

1.0 DEElHII1QNS 1.15 DEE1HED TERMS 1.19 DELETED The DEFINED TERMS of this section appear in capitalized type throughout these Technical Specifications.

1.16 DEGREE _QE_REQLfRDaliCY 1.20 IDD!UEIED_ LEAR _ AGE Hith reference to redundant instrument or logic IDENTIFIED LEAKAGE shall be:

channels, the DEGREE OF REDUNDANCY is the a.

Leakage (except CONTROLLED LEAKAGE) into closed difference between the number of OPERABLE channels systems, such as pump seal or valve packing and the minimum number of these channels which, leaks that are captured and conducted to a sump when tripped, will cause an automatic system trip.

or collecting tank, or b.

Leakage into the containment atmosphere from 1.17 DQSE_EQu1 VALENT I-111 sources that are both specifically located and known either not to interfere with the DOSE EQUIVALENT I-131 shall be that concentration operation of leakage detection systems or not of I-131 (microcurle/ gram) which alone would to be PRESSURE BCUNDARY LEAKAGE, or produce the same thyroid dose as the quantity and c.

Reactor coolant system leakage through a steam isotopic mixture of I-131, I-132. I-133, I-134, generator to the secondary system.

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 Rev.

1, dated October 1977.

1.18 E - AVERAGE E SINTEGRATION ENERG1 E shall be the average (weighted in proportion to the concentration of each radionucIlde in the sample) of the sum of the average beta and gamma energies per disintegration (in HeV/d) for isotopes, other than todines, with half Ilves greater than 15 minutes, making up at least 95% of the total non-lodine activity in the coolant.

3 Amendment Nos.146 and 134

1.0 DIELMLLIQHS 1.21 INSIEUMENT CHANNEL 1.25 QEE SITE AC POWER SQURCES An INSTRUMENT CHANNEL shall be the combination of sensor (s), signal processing elements, output devices The OFF-SITE A.C. P0HER SOURCES shall be the System and other components and circuitry as required to Auxlitary Transformer - 142 - (Unit 1); 242 - (Unit 2).

The System Auxiliary Transformer from the measure and evaluate a process variable for the opposite unit (242 - Unit 1; 142 - Unit 2).

purpose of observation, control and/or protection of the process system. A channel may produce both 1.26 OffSITE DOSE CALCULATION MANUAL (00CH) analog signal outputs and discrete (signal or electromechanical component operation) outputs.

The The OFFSITE DOSE CALCULATION MANUAL shall contain the channel terminates and loses 'ts identity where methodology and parameters used in the calculation of Individual channel outputs are combined.

offsite doses resulting from radioactive gaseous and i

1.22 LEAKAGE 11guld effluents, in the calculation of gaseous I

11guld effluent monitoring alarm / trip setpoints, and in the conduct of the Environmental Radiological Refer to the following specific paragcaphs.

Monitoring Program.

The ODCH shall also contain (1) 1.13 CONTROLLED LEAKAGE 1.20 1DENTIFIED LEAKAGE the Radioactive Effluent Controls and Radiological 1.32 PRESSURE BOUNDRY LEAKAGE Environmental Monitoring Programs required by Section 1.48 UNIDENTIFIED LEAKAGE 6.2.6 and (2) descriptions of the information that should be included in the Annual Radiological 1.23 BASTER RELAY TEST Environmental Operating and Radioactive Effluent Release Reports required by Specifications 6.6.1.C and 6.6.1.D.

A MASTER RELAY TEST shall be the energization of each master relay and verification of OPERABILITY of each 1.27 OPERABLE - OPERABLLLIX relay.

The MASTER RELAY TEST shall include a continuity check of each associated slave relay.

A system, subsystem, train, component or device shall 1.24 MEMBER (S) 0F THE PUBLIC be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s), and when all necessary attendant Instrumentation, controls, MEMBER (S) 0F THE PUBLIC shall include all persons who electrical power, cooling or seal water, lubrication are not occupationally associated with the plant.

or other auxiliary equipment that are required for This category does not include employees of the the system, subsystem, train, component, or device to utility, its contractors or its vendors. Also excluded from this category are persons who enter the perform its function (s) are also capable of peforming site to service equipment or to make delivertes.

their related support function (s).

This category does include persons who use portions of the site for recreational, occupational or other The OPERABILITY requirement for both normal and purposes not associated with the plant.

emergency AC power supplies does not apply to AC Instrument busses and associated instruments.

4 Amen <inents 146 and 134

1.0 DEEINillONS 1.28 QEERallHG 1.33 P30 CESS _ CONTROL PROGBAM lPCP)

OPERATING is defined as performing the intended The PROCESS CONTROL PROGRAM (PCP) shall contain the function in the intended manner.

current formulas, sampling, analyses, tests and determinations to be made to ensure that the 1.29 REERallhG CYCLE processing and packaging of solid radioactive vastes based on demonstrated processing of actual or The OPERATING CYCLE shall be the interval between the simulated wet solid wastes will be acc pplished in end of one major refueling outage and the end of the such a way as to assure compilance with 10 CFR Parts next subsequent major refueling outage per unit.

20, 61 and 71, State regulations, burial ground requirements and other requirements governing the 1.30 QEE.BalIONAL_MQQE - HODE disposal of radioactive waste.

An OPERATIONAL HODE (i.e. MODE) shall correspond to 1.34 PRQIECT10N_LQGIC_C.HANNEL any one inclusive combination of core reactivity condition, power level, and average reactor coolant A PROTECTION LOGIC CHANNEL shall be an arrangement of temperature specified in Table 1.1, when fuel relays, contacts or other components which operate in assemblies are present in the reactor vessel.

response to INSTRUMENT CHANNEL outputs to produce a decision output. The decision output is the 1.31 ERYSICS TESTS initiation of a protective action signal. At the system level, the decision output is the operation of PHYSICS TESTS shall be those tests performed to a sufficient number of ACTUATION DEVICES and the measure the fundamental nuclear characteristics of associated ACTUATED EQUIPMENT as required to place or the reactor core and related instrumentation and 1) restore the Nuclear Steam Supply System to a design described in Chapter 14.0 of the FSAR, 2) authorized safe state.

The channel is deemed to include the under the provisions of 10 CFR 50.59, or 3) otherwise ACTUATION DEVICES.

approved by the Commission.

1.3S PROIECllDN_SYSIEH 1.32 ERESSURE.EXLHDARY LEAKAGE The PROTECTION SYSTEM shall consist of both the PRESSURE BOUNDARY LEAKAGE shall be leakage (except Reactor Protection System and the Engineered steam generator tube leakage) through a non-isolable Safeguards System.

The PROTECTION SYSTEM encompasses fault in the Reactor Coolant System component body, all electric and mechanical devices and circuitry pipe wall, or vessel wall.

(from sensors through ACTUATION DEVICES) which are required to operate in order to place or restore the Nuclear Steam Supply System to a design safe state.

5 Amendment Nos. 146 and 134

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1.0 DEEINillONS 1.40 RELUELING_ClCl,E_QR_QUIAGE 1.36 EURGE - PURGING When REFUELING CYCLE or OUTAGE is used to PURGE OR PURGING is the controlled process of designate a surveillance interval, the discharging air or gas from a confinement to maintain surveillance shall be performed at least once l

temperature, pressure, humidity, concentration or every 18 months as allowed by general other operating condition, in such a manner that requirement 4.0.2.

replacement air or gas is required to purify the confinement.

1.41 BEPORIABLE EVEMI 1.37 QUAQRANT P0HER TILT RATIQ A REPORTABLE EVENT shall be any of those conditions specified in Specification 6.6.2 or QUADRANT POWER TILT RATIO shall be the ratio of the Section 50.73 of 10 CFR Part 50.

maximum upper excore detector calibrated output to the average of the upper excore detector calibrated 1.42 SHUIDQHN MARGLN outputs, or the ratio of the maximum lower excore detector calibrated output to the average of lower SHUTDOHN MARGIN shall be the instantaneous excore detector calibrated outputs, whichever is amount of reactivity by which the reactor is greater.

subcritical or would be subtritical from its present condition assuming all control and 1.38 BATED _T_REE_ MALE 0HEB shutdown banks are fully inserted, except for the single rod cluster assembly of highest RATED THERMAL POWER shall be a total steady state reactivity worth which is assumed to be fully reactor core heat transfer rate to the reactor withdrawn.

coolant of 3250 MHt.

1.43 SIIE_ BOUNDARY 1.39 REAC10R PRES 1URE The SITE BOUNDARY shall be that line beyond The REACTOR PRESSURE shall be the pressure in 15-which the land is not owned, leased or otherwise steam space of the pressurl er.

controlled by the licensee.

i 1.44 DELETED 6

Amendment Nos. 146 and 134 l

l 1

1.Q QEFINITIONS 1.45 SOURCE CHECK 1.50 DELETED l

A SOURCE CHECK shall be the qualitative assessment of 1.51 VENTING channel response when the Channel sensor is exposed to a radioactive source.

VENTING is the controlled process of discharging air or gas from a confinement to maintain temperature, 1.46 SURVEILLANCE FREQUENCY NOTATION pressure, humidity, concentration or other operating condition, in such a manner that replapement air or The SURVEILLANCE FREQUENCY NOTATION specified for the gas is not provided or required during venting.

performance of Survelliance Requirements shall Vent, used in system names, does not imply a venting correspond to the intervals defined in Table 1.2. and process.

General Surveillance Requirement 4.0.2.

1.47 IliEBri&L.10HER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

1.48 UNIDENTIFIED _LEAEAGE UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE or CONTROLLED LEAKAGE.

1.49 UNRESTRICTED AREA An UNRESTRICTED AREA shall be any area at or beyond i

the SITE BOUNDARY to which access is not controlled by the licensee for purposes of protection of Individuals from exposure to radiation and radioactive materials or any area within the site boundary used for residential quarters or for industrial, commercial, institutional and/or recreational purposes.

l 6a Amendment Nos-146 and 134

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LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.11 Radioactive Liquids 4.11 Radioactive Liquids I.

Quidoor L1aufd Tanks I.

Qu_tdoor L1auldltrt!Ui The quantity of radioactive material A.

The quantity of radioactive material contained in the following tanks shall be excluding tritium and noble gases Ilmited to less than or equal to 10 curles, contained in the listed tanks shall be g

excluding dissolved or entrained gases and determined to be within the limit by tritium:

analyzing a representative sample of the tank (s) contents at least once per a.

Outside temporary radioactive liquid week when radioactive materials are storage tank (s) (when applicable) being added to the tank (s).

These concentrations will be determined as prescribed in the 00CH.

APPLICABILITY: At all times ACTION:

a.

Hith the quantity cf radioactive material in the above listed tank (s) exceeding the above 11mit, immediately suspend all additions of radioactive material to the tank (s) and within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> reduce the tank (s) contents to within the limit.

b.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

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225a Amendment Nos.146 and 134

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3.11.1 00JTD00R LIOUID TANKS 4.11.1 Restricting the quantity of radioactive material contained in those cutdoor tanks specified, that are not surrounded by liners, dikes, or walls capable of holding the tank contents and that do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system, provides assurance that in the event of an uncontrolled release of the tanks contents, the resulting concentrations would be less than the limits of 10 CFR Part 20, Appendix B, Table II, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted area.

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229a Amendment Nos. 146 and 134

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I LIMIIlhG_. CONDITION FOR OPERAI1QN SURVEILL6MCE_ REQUIREMENT 3.12.

Gaseous Effluents 4.12.

Gaseous Effluents 1.

Gu_ Decay Tanks 1.

Gu _Qecav Tanks A.

The quantity of radioactivity A.

The quantity of radioactive material contained in each gas decay tank contained in each gas decay tank shall shall be limited to 22,000 curies be determined to be within the above (considered as Xe-133).

limit at least weekly wheng adioactive r

materials are being added to the tank.

APPLICABILilY: At all times ACIl0ft:

a.

With the quantity of radloactive material in any gas decay tank exceedtr:g the above limit, immediately suspend addition of radioactive material to the tank and within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> reduce the tank contents to within the limit.

b.

The provision of Specifications 3.0.3 and 3.0.4 are not applicable.

234 Amendment Nos. 146 and 134

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LIMITING CONDITION FOR OPERATION SURVELLLANCE REQUIREMENT 3.12 Gaseous Effluents (Continued) 4.12 Gaseous Effluents (Continued) l 2.

Exolosive Gas Hlxture 2.

Exp19111e_ Gas Mixture A.

The concentration of hydrogen or A.

The concentrations of hydrogen or oxygen in the waste gas system

• oxygen in the waste gas system shall shall be limited to less than or be determined to be within limits, at equal to 31 by volume.

least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

APPLICABILLIX:

At all times ACIl0M:

a.

With the concentration of hydrogen and oxygen in the waste gas system each greater than 31 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 Ilmit.

b.

Hith 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 hydrogen or oxygen to less than or equal to 31.

c.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

• The waste gas syytem consists of the following: Gas Decay Tank when on fill or cover gas and the Hold-up Tanks.

235 Amendment Nos. 146 and 134

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3.12 & 4.12 Restricting the quantity of radioactivity contained in each gas decay tank provides anne:.oce that in the event of an uncontrolled release of the tanks contents, the resulting total body exposure to an individual at the nearest unrestricted area boundary will not exceed 0.5 rem.

242a-Amendment Nos. 146 and 134

b DELETED 275 - 280 Amendment Nos. 146 and 134

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t 6.1.7.A.1 (Continued) i During periods when the Superintendent of the Offsite Review and Investigative Function is unavailable, he shall designate this responsibility to an established alternate who satisfies the formal training and experience requirements for the Superintendent of the Offsite Review and Investigative Function.

The responsibilities of the personnel performing this function are stated below. The Offsite Review and Investigative Function shall review:

(a) The safety evaluations for changes to procedures, equipment or systems as described in the safety analysis report and for tests or experiments completed under the provision of 10CFR 50.59 to verify that such actions did not constitute an unreviewed safety question. Proposed changes to the Quality Assurance Program description shall be reviewed and approved by the Manager of QA/NS.

(b) Proposed changes to procedures, equipment or systemt 4tich involve an unreviewed safety question as defined in 10CFR 50.59.

(c) Proposed tests or experiments which involve an unreviewed safety question as defined in 10CFR 50.59.

(d) Proposed changes in Technical Specifications or NRC facility operating licenses.

te) Noncompliance with NRC requirements, or of internal procedures or instructions having nuclear safety significance.

(f) Significant operating abnormalltles or deviations from normal and expected performance of plant equipment that affect nuclear safety as referred to it by the Onsite Review and Investigative Function.

(g) All REPORTABLE EVENTS.

(h) All recognized Indications of an unanticipated deficiency in some aspect of design or operation of safety related structures, systems or components.

(1) All changes to the Generating Stations Emergency Plan prior to implementation of such changes.

(j) All items referred by the Technical Staff Supervisor, Station Manager, Vice President PNR Operations and General Manager Quality Programs and Assessment.

(k) DELETED (1) DELETED 302 Amendment Nos. 146 and 134

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6.2 (Continued) 5.

Drills of the emergency procedures described in Specification 6.2.1.D shall be conducted at the frequency specified in the Generating Station Emergency Plan.

These drills will be planned so that during the course of the year, communication links are tested and outside agencies are contacted.

6.

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

A.

Ra dioa.r ttv elf flue n t_ Con ttc l_5 P ros ram m

A program shall be provided conforming with 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to MEMBERS OF THE PUBLIC from radioactive effluents as low as reasonably achievable.

The program.(1) shall be contained in the ODCH, (2) shall be implemented by station procedures, and (3) shall include remedial actions to be taken whenever the program limits are exceeded.

The program shall include the following elements:

1)

Limitations on the operability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the

ODCM, 2)

Limitations on the concentrations of radioactive material released in liquid effluents to UNRESTRICTED AREAS conforming to 10 CFR Part 20, Appendix B, Table II, Column 2, 3)

Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.106 and with the methodology and parameters in the ODCM, 4)

Limitations on the annual and quarterly doses or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released from each unit to UNRESTRICIED AREAS conforming to Appendix I to 10 CFR Part 50, 5)

Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current calendar year in-accordance with the methodology and parameters in the ODCH at least every 31 days, 6)

Limitations on the operability and use of the liquid and gaseous effluent treatment systems to ensure that the appropriate portions of these systems are used to reduce releases.of radioactivity when the projected doses in a 31-day period would exceed 2 percent of the guidelines for the annual dose or dose commitment conforming to Appendix I to 10 CFR Part 50, 310a Amendment Nos.146 and 134

t 6.2 (Continued) d 7)

Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the SITE BOUNDARY conforming to the doses associated with 10 CFR Part 20 Appendix B, Table II, Column 1 8)

Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50, 9)

Limitations on the annual and quarterly doses to a HEMBER OF 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 beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50,

10) Limitations on the annual dose or dose commitment to any HEH8ER OF THE PUBLIC due to releases of radioactivity and to radiation from urantum fuel cycle sources conforming to 40 CFR Part 190.

B.

Radioloalcal Environmental Monitorina ProatAm A program shall be provided to monitor the radiation and radionuclides in the environs of the plant.

The program shall provide (1) representative measurements of radioactivity in the highest potential exposure pathways, and (2) verification of the accuracy of the effluent monitoring program and modeling of environmental exposure pathways. The program shall (1) be contained in the 00CM, (2) conform to the guidance of Appendix I to 10 CFR Part 50, and (3) include the following:

1)

Honttoring, sampling, analysis, and reporting of radiation and radionuclides in the environment in accordance with the methodology and parameters in the 00CH, 2)

A Land Use Census to ensure that changes in the use of areas at and beyond the SITE BOUNDARY are Identifled and that modifications to the monitoring program are made if required by the results of this census, and 3)

Participation in a Interlaboratory Comparison Program to ensure that independent checks on the precision and accuracy of the measurements or radioactive materials in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring.

6.3 Action to be Taken in the Event of a Reportable Event in Plant Operation:

Any Reportable Event shall be promptly reported to the Vice President PHR Operations or his designated alternate.

The incident shall be promptly reviewed pursuant to Specification 6.1.7.B.2.(k) and a separate report for each reportable event shall be prepared in accordance with the requirements of 10CFR 50.73.

310b Amendment Nos. 146 and 134-

6.5.2 (Continued)

C.

Records of new and irradiated fuel inventory, fuel transfers and assembly burnup histories.

D.

Updated, corrected, and as-built drawings of the plant.

E.

Records of plant radiation and contamination surveys.

F.

Records of offsite environmental monitoring surveys.

G.

Records of radiation exposure for all plant personnel, including all contractors and visitors [o the plant in accordance with 10 CFR 20.

H.

Records of radioactivity in liquid and gaseous wastes released to the environment.

I.

Records of transient or operational cycling for those components that have been designed to operate safely for a limited number of transient or operational cycles.

J.

Records of training and qualification for current members of the station staff.

K.

Inservice inspections and testing performed pursuant to the Technical Specifications.

L.

Minutes of meetings and results of reviews performed by the Offsite and Onsite Review Functions.

i l

H Records of secondary water sampling and water quality.

t l

N.

Records for environmental qualification of components required by the Environmental Qualification (EQ) l Program.

O.

Records of the service lives of all snubbers covered by Specification 3.22 including the date at which the service life commences and associated installation and maintenance records.

P.

Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10CFR 50.59.

Q.

Records of Quality Assurance activities required by the Q.A. Manual.

R.

Records of reactor tests and experiments.

S.

Records of reviews performed for changes made to the OFFSITE DOSE CALCULATION MANUAL and the PROCESS CONTROL PROGRAM.

311a Amendment Nos.146 and 134

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C.

Annual Radioloolcal EnvironmeRial_Qaerating D.

Ba_d10 active Effluent Relene_.Resor_t**

ReDatt' The Radioactive Effluent Release Report covering the The Annual Radiologics1 Environmental operation of the Unit shall be submitted in Operating Report covering operation of the accordance with 10 CFR 50.36a. The report shall Unit during the previous calendar year shall include a summary of the quantitles of radioactive be submitted before May 1 of each year.

The 11guld and gaseous effluents and solid waste released report shall include summartes, from the Unit. The material provided shall be (1)

Interpretations, and analysis of trends of the consistent with the objectives outilned in the ODCM results of the Radiological Environmental and PCP and (2) in conformance with 10 CFR 50.36a and Monitoring Program for the reporting period.

Section IV.B.1 of Appendix I to 10 CFR Part 50.

The material provided shall be consistent with the objectives outilned in (1) the ODCM and (2) Sections IV.B.2, IV.B.3, and IV.C of Appendix I of 10 CFR Part 50.

A single submittal may be made for a multi-unit station.

A single submittal may be made for a multi-unit station.

The submittal should combine those Sections common to both units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.

314 Amendment Nos. 146 and 134

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F.

CORE OPERATING LIMITS REPORT (COLR)_

E.

Bonthly Operatina ReDQE1 Routine reports 1.

Core operating 15 nits shall be established and using the " Operating Data", " Unit Shutdowns documented in the CORE OPERATING LIMITS REPORT -

and Load Reductions" and " Average Dally Unit Power Level" report formats, shall be before each reload cycle or any remaining part submitted on a monthly basis to the Director.

of a reload cycle for the following:

Office of Hanagement and Program Analysis, 1)

Shutdown Bank Insertion Limit for U.S. Nuclear Regulatory Commission, Specifications 3.2.1.C.2 and 0.1.

Washington, D.C. 20555 with a copy to the 2)

Control Bank Insertion Limit for appropriate NRC Regional Office of Inspection Specifications 3.2.1.C.2 and D.1, and and Enforcement, no later than the 15th of 3.2.3.B.4, 5.

each month following the calendar month 3)

AXIAL FLUX DIFFERENCE Limits, Target Band covered by the report.

For Specific 3* tons 3.2.2.A.2.2.c.(2),

3.2.2.A.4, and 3.2.2.A.6.1.

Included witn the report will be a " Narrative 4)

Heat Flux Hot Channei Factor and K(Z) for Summary of Operating Experience" that Specification 3.2.2.A.1.1.

describes the operation of the facility, 5)

Nuclear Enthalpy Rise Hot Channel Factor including major safety related maintenance, and Power Factor Hultiplier for for the monthly report period.

Specification 3.2.2.A.I.1.

2.

The analytical methods used to determine the core operating Ilmits shall be.those previously reviewed and approved by the NRC in the following documents:

l 1)

HCAP-9272-P-A. "Hestinghouse Reload Safety Evaluation Methodology", July 1985 (Westinghouse Proprietary).

(Methodology for Specifications: Shutdown Bank Insertion Limit, Control Bank Insertion Limit, Axlal Flux Difference Heat Flux Hot Channel Factor, and Nuclear Enthalpy Rise Hot Channel Factor.)

316 Amendment Nos. 146 and 134

6.6.2.B.1.a (Continued)

(IX)

Reports submitted to the Conmission in accordance with paragraph (VIII) above also meei the effluent release reporting requirements of 10CFR Part 20 paragraph 20.405(a)(5)

(X)

Any event that posed an actual threat to the safety of the plant or significantly hampered site person"al in the performance of duties necessary for the safe operation of the nuclear power plant including fires, toxic gas releases, or radioactive releases.

4 6.6.3 Unique Reportina Reautrements A.

DELETED l

l l

322 Amendment Nos. 146 and 134

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6.6.3.8 Scaclai Reoorts Reports on the following areas shall be submitted as indicated:

TOPIC SUBHITTAL DATE a.

In-Service Inspection Evaluation In accordance with the requirements of Section XI ASME Boller Code (INA-6000) b.

(Deleted)

Containment Building Structural Testing Hithin 90 days following completion of each test.

c.

Report (ILRT, Tendon) d.

Changes.to,the Offsite Dose Calculation Manual (00CM)

As part of, or concurrent with, the Radioactive Effluent Release Report in accordance with Specification 6.7.C.

Waukegan Regional Airport Expansion The expansion plans and status of such plans for the e.

Plans.

Haukegan Regional.Alrport will be reported yearly in the Annual Report, including FAA form #5010 report.

f.

Low Temperature Overpressure Protection Hithin 30 days of operation.

System Operation g.

Primary Coolant Specific Activity As part of the Annual Report whenever Specification 3.3.6 i

has been exceeded.

See page 124 for required report content.

h.

Pressurizer PORV or Safety Valve Within 30 days as an LER.

l Failure to Clost SPECIAL REPORTS l

Table 6.6-1 323 Amendment Nos. 146 and 134

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6.7 Off$U1_ DOSE CALCULAIl0N MANUAL 100Dn Changes to the OOCH:

a.

Shall be documented and records of b.

reviews performed shall be retained as Shall become effective after review and acceptance by the Onsite Review and Investigative Function, and the required by Specification 6.5.2.5.

The documentation shall contain:

approval of the Station Manager on the date specified by the Onsite Review and Investigative Function.

1)

Sufficient information to support Shall be submitted to the Commission in the form of a c.

the change together with appropriate complete, legible copy of the entire OOCH as a part analyses or evaluations justifying of, or concurrent with, the Radioactive Effluent the change (s); and Release Report for the period of the report in which any change to the ODCM was made effective.

Each 2)

A determination that the change will change shall be identified by markings in the margin maintain the level of radioactive of the affected pages, clearly indicating the area of effluent control required by 10 CFR the page that was changed, and shall Indicata the 20.106, 40 CFR Part 190, 10 CFR date (e.g., month / year) the change was implemented.

50.36a and Appendix I to 10 CFR Part 50 and not adversely impact the accuracy or reliability of effluent, dose or setpoint calculations.

325 Amendment Nos. 146 and 134

o 6.8 ElRQd1Dg_PrQ1ttil2D 1.

In the event of the possibility of flooding, all doors listed in Table 6.8-1 shall be verified closed 6.9 ERQCESS_C011 TROL PROGRAM (PCP)

C,hanges to the PCP:

a.

Shall be documented and records of reviews performed shall be retained as required by Specification 6.S.2.5.

This documentation shall contain:

1)

Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change (s) and 2)

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

Shall become effective after review and acceptance by the Onsite Review and Investigative Function and the approval of the Station Manager.

326 Amendment Nos.146 and 134

- -.