ML20198M802

Amend 104 to License DPR-22,revising TS So That Either 8 or 12 H Shifts Considered Normal & 40 H Considered Nominal Week & Changing Wording for Surveillances Required Once Per Shift to Once Per 12 H

ML20198M802
Person / Time
Site:	Monticello
Issue date:	12/24/1998
From:	Lyon C NRC (Affiliation Not Assigned)
To:
Shared Package
ML20198M809	List: ML20198M802 ML20198M822
References
NUDOCS 9901050361
Download: ML20198M802 (52)

Text

{{#Wiki_filter:__._ "84g p k UNITED STATES g j NUCLEAR REGULATORY COMMISSION e WASHINGTON, D.C. 30006 0001 NORTHERN STATF,S POWER COMPANY DOCKET NO. 50-283 MONTICELLO NUCLEAR GENERATING PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.104 License No. DPR-22 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment by Northem States Power Company (the licensee) dated August 15,1996, as supplemented March 19 and October 12, 1998, 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; 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 inimicht 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 regu'ations and all applicable requirements have been satisfied. 2. Accordingly, the license is amended by changes to the Technical Specifications as L indicated in the attachment to this license amendment, and paragraph 2.C.2 of Facility Operating License No. DPR-22 is hereby amended to read as follows: l i 9901050361 981224 ~ PDR ADOCK 05000263 P Pm a e m. +y e e

w l ; Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 104 , are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications. 3. This license amendment is effective as of the date of issuance, with full implementation within 30 days. FOR THE NUCLEAR REGULATORY COMMISSION l CdF. Carl F. Lyon, Project Manager Project Directorate 111-1 l Division of Reactor Projects - lil/IV { Office of Nuclear Reactor Regulation l

Attachment:

Changes to the Technical l Spacifications l Date of issuance: December 24, 1998 i I e I l I

l g l ATTACHMENT TO LICENSE AMENDMENT NO.104 FACILITY OPERATING LICENSE NO. DPR 22 DOCKET NO. 50-263 Revise Appendix A Technical Specifications by removing the pages identified below and j inserting the attached pages. The revised pages are identified by amendment number and i contain verticallines indicating the areas of change.

REMOVf, INSERT l

i i 11 li lii 111' i iv iv v v vi vi vil vil 22 22 31 31 61 61 62 62 63a 63a 69 69 72 72 82 82 89 89 99 99 102 102 125 125 126 126 164 164 188 188 190 190' 198t 198t 200 200 223 223 227b-227e 227b-227e 229b 229b 229c 229c 229ff 229ff 229i 2291 ' Font change only for document consistency No other changes to these pages. I w

-..... ~... _ _ ATTACHMENT TO LICENSE AMENDMENT NO.104 (Continued) FACILITY OPERATING LICENSE NO. DPR-22 DOCKET NO. 50-263 1 l l REMOVE INSERT i 229u 229u 232 232 233 233 l 2M 2M l 237 237 239 239 240 240 241 241 242 242 243 243 244a 244a 246b 246b 247a 247a 250 250 l r l l i l i li i t

ls l l l TABLE OF CONTENTS l Eagst 1.0 DEFINITIONS 1 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 and 2.3 Fuel Cladding Integrity 6 2.1 Bases 10 2.3 Bases 14 2.2 and 2 4 Reactor Coolant System 21 2.2 Bases 22 2.4 Bases 24 3.0 LIMITING CONDITIONS FOR OPERATION AND 4.0 SURVEILLANCE REQUIREMENTS l 4.0 Surveillance Requirements 25a l 4.0 Bases 25b 3.1 and 4.1 Reactor Protection System 26 3.1 Bases 35 4.1 Bases 42 l l 3.2 and 4.2 Protective Instrumentation 45 A. Primary Containment isolation Functions 45 B. Emergency Core Cooling Subsystems Actuation 46 C. Control Rod Block Actuation 46 D. Other instrumentation 46a E. Reactor Building Ventilation isolation and Standby Gas Treatment System Initiation 47 F. Recirculation Pump Trip initiation and Alternate Rod injection Initiation 48 G. Safeguards Bus Voltage Protection 48 H. Instrumentation for S/RV Low-Low Set Logic 48 l. Instrumentation for Control Room Habitability Protection 48 3.2 Bases 64 4.2 Bases 72 3.3 and 4.3 Control Rod System 76 A. Reactivity Limitations 76 B. Control Rod Withdrawal 77 C. Scram insertion Times 81 D. Control Rod Accumulators 82 l E. Reactivity Anomalies 83 l-F. Scram Discharge Volume 83a G. Required Action 83a 3.3 and 4.3 Bases 84 I l Amendment No. 30, 37, 45, 65, 104 l

u TABLE OF CONTENTS (Cont'd) P.AQft 3.4 and 4.4 Standby Liquid Control System 93 A. System Operation 93 B. Boron Solution Requirements 95 C. 96 3.4 and 4.4 Bases 99 3.5 and 4.5 Core and Containment / Spray Cooling Systems 101 l A. ECCS Systems 101 B. RHR Intertie Return Line Isolation Valves 103 C. Containment Spray / Cooling System 104 D. RCIC 105 E. Cold Shutdown and Refueling Requirements 106 F. Recirculation System 107 3.5/4.5 Bases 110 3.6 and 4.6 Primary System Boundary 121 A. Reactor Coolant Heatup and Cooldown 121 B. Reactor Vessel Temperature and Pressure 122 C. Coolant Chemistry 123 D. Coolant Leakage 126 F. Safety / Relief Valves 127 F. Deleted G. Jet Pumps 128 H. Snubbers 129 3.6 and 4.6 Bases 145 l 3.7 and 4.7 Containment Systems 156 A. Primary Containment 156 B. Standby Gas Treatment System 166 C. Secondary Containment 169 D. Primary Containment isolation Valves 170 E. Combustible Gas Control System 172 l 3.7 Bases 175 4.7 Bases 183 ii Amendment No. 9,35,47,74,77,79,102. SEP 16 1998

w TABLE OF CONTENTS (Cont'd) EaQA j 3.8 and 4.8 Radioactive Effluents 192 A. Liquid Effluents 192 B. Gaseous Effluents 197 . C. Solid Radioactive Waste 198e D. Dose from All Uranium Fuel Cycle Sources 198f 3.8 and 4.8 Bases 198u 3.9 and 4.9 Auxiliary Electrical Systems 199 A. Operational Requirements for Startup 199 B. Operational Requirements for Continued Operation 200 1. Transmission Unes 200 2. Reserve Transformers 201 3. Standby Diesel Generators 201 4. Station Battery System 203 5. 24V Battery Systems 203 3.9 Bases 204 4.9 Bases 205 3.10 and 4.10 Refueling 206 A. Refueling Interlocks 206 B. Core Monitoring 207 C. Fuel Storage Pool Water Level 207 D. Movement of Fuel 207 E. Extended Core and Control Rod Drive Maintenance 208 3.10 and 4.10 Bases 209 3.11 and 4.11 Reactor Fuel Assemblies 211. A. Average Planar Linear Heat Generation Rate 211 B. Linear Heat Generation Rate 212 C. Minimum Critical Power Ratio 213 . 3.11 Bases 216 4.11 Bases 218 l 3.12 and 4.12 Sealed Source Contamination 219 l A. Contamination 219 B. Records 221 3.12 and 4.12 Bases 222 i iii j Amendment No. 29,102 SEP 161998

_. _ ~ a m TABLE OF CONTENTS (Cont'd) PAUt 3.13 and 4.13 Fire Detection Protection Systems 223 A. Fire Detection Instrumentation 223 B. Fire Suppression Water System 224 C. Hose Stations 226 D. Yard Hydrant Hose Houses 227 E. Sprinkler Systems 227a F. Halon Systems 227b s G. Penetration Fire Barriers 227b H. Alternate Shutdown System 227c 3.13 Bases 228 4.13 Bases 228b l 3.14 and 4.14 Accident Monitoring instrumentation 229a j 3.14 and 4.14 Dases 229e 3.15 and 4.15 Inservice Inspection and Testing 229f 3.15 and 4.15 Bases 229g 3.16 and 4.16 Radiation Environmental Monitoring Program 229h A. Sample Collection & Analysis 229h l B. Land Use Census 229j C. Interlaboratory Comparison Program 229k l 3.16 and 4.16 Bases 2291 j 3.17 and 4.17 Control Room Habitability 229u A. Control Room Ventilation System 229u B. Control Room Emergency Filtration System 229v' 3.17 Bases 229y 4.17 Bases 229z 5.0 DESIGN FEATURES 230 5.1 Site 230 5.2 Reactor 230 5.3 Reactor Vessel 230 - 5.4 Containment 230 5.5 Fuel Storage 231 5.6 Seismic Designs 231 6.0 ADMINISTRATIVE CONTROLS 232 6.1 Organization 232 6.2 Review and Audit 237 l 6.3 SpecialInspection and Audits 243 ( 6.4 Action to be taken if a Safety Limit is Exceeded 243 6.5 Plant Operating Procedures 244 6.6 Plant Operating Records 246c l 6.7 Reporting Requirements 248 iv Amendment No. 45,37,-46, 61,65, 104 _ _ _ =

1 i LIST OF FIGURES j Fioure No. Enga 3.4.1 Sodium Pentaborate Solution Volume-Concentration Requirements 97 3.4.2 Sodium Pentaborate Solution Temperature Requirements 98 i 3.6.1 Core Bettline Operating Limits Curve Adjustment vs. Fluence 133 3.6.2 Minimum Temperature vs. Pressure for Pressure Tests 134 3.6.3 Minimum Temperature vs. Pressure for Mechanical Heatup or Cooldown Without the Core Critical 135 3.6.4 . Minimum Temperature vs. Pressure for Core Operation 136 4.6.2 Chloride Stress Corrosion Test Results @ 500 F 137 3.7.1 Differential Pressure Decay Between the Drywell and Wetwel! 191 3.8.1 Monticello Nuclear Generating Plant Site Boundary for Liquid Effluents 1989 3.8.2 Monticello Nuclear Generating Plant Site Boundary for Gaseous Effluents 198h 1 v Amendment No. 9,35,47,74,7-7,79, 104

a LIST OF TABLES Table No. Eagg 3.1.1 Reactor Protection System (Scram) Instrument Requirements 28 4.1.1 Scram Instrument Functional Tests - Minimum Functional Test Frequencies for Safety Instrumentation and Control Circuits 32 4.1.2 Scram Instrument Calibration - Minimum Calibration Frequencies for Reactor Protection Instrument Channels 34 3.2.1 Instrumentation that initiates Primary Containment Isolation Functions 49 3.2.2 Instrumentation that initiates Emergency Core Cooling Systems 52 3.2.3 instrumentation that initiates Rod Block 56 l 3.2.4 Instrumentation that initiates Reactor Building Ventilation isolation and Standby Gas Treatment System Initiation 59 3.2.5 Instrumentation That Initiates a Recirculation Pump Trip and Alternate Rod injection 60 3.2.6 Instrumentation for Safeguards Bus Degraded Voltage and Loss of Voltage Protection 60a 3.2.7 Instrumentation for Safety / Relief Valve Low-Low Set Logic 60b 3.2.8 Other Instrumentation 60d 3.2.9 Instrumentation for Control Room Habitability Protection 60e 4.2.1 Minimum Test and Calibration Frequency for Core Cooling, Rod Block and Isolation instrumentation 61 4.6.1 Snubber Visual Inspection Interval 132a l 3.8.1 Radioactive Liquid Effluent Monitoring Instrumentation 198i 3.8.2 Radioactive Gaseous Effluent Monitoring Instrumentation 198k 4.8.1 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements 198m 4.8.2 Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements 198n 4.8.3 Radioactive Liquid Waste Sampling and Analysis Program 198p 4.8.4 Radioactive Gaseous Waste Sampling and Analysis Program 198s vi Amendment No. 37,39,-44,45,65,74,82, 104

LIST OF TABLES (Cont'd) Table No. Eaga 3.13.1 Safety Related Fire Detection Instruments 227d l 3.14.1 Instrumentation for Accident Monitoring 229b 4.14.1 Minimum Test an'd Calibration Frequency for Accident Monitoring Instrumentation 229d 4.16.1 Radiation Environmental Monitoring Program (REMP) Sample Collection and Analysis 229-l 4.16.2 REMP - Maximum Values for the Lower Limits of Detection (LLD) 229q l 4.16.3 REMP - Reporting Levels for Radioactivity Concentrations in Environmental Samples 229s 6.1.1 Minimum Shift Crew Composition 236 i i i i ) vii Amendment No. 45, 37,-44, 54,70, 104

Bases 2.2: The reactor coolant system integrity is an important barrier in the prevontion of uncontrolled release of fission products. It is essential that the integrity of this system be protected by establishing a pressure limit to be observed for all operating conditions and whenever there is irradiated fuel in the reactor vessel. The pressure safety limit of 1335 psig as measured in the vessel steam space is equivalent to 1375 psig at the lowest elevation of the reactor coolant system. The 1375 psig value was derived from the design pressures of the reactor pressure vessel, coolant piping, and recirculation pump casing. The respective design pressures are 1250 psig at 575 F,1148 psig at 5620F, and 1380 psig l at 575 F. The pressure safety limit was chosen as the lower of the pressure transients permitted by the applicable design codes: l ASME Boiler and Pressure Vessel Code Section ill-A for the pressure vessel, ASME Boiler and Pressure Vessel Code Section ill-C for the recirculation pump casing, and the USAS Piping Code Section B31.1 for the reactor coolant system piping. The ASME Code permits pressure transients up to 10 percent over the vessel design pressure (110% x 1250 = 1375 psig) and the USAS Code permits pressure transients up to 20 percent over the piping design pressure (120% x 1148 = 1378 psig). The design basis for the reactor pressure vessel makes evident the substantial margin of protection against failure at the safety pressure limit of 1375 psig. The vessel has been designed for a general membrane stress no greater than 26,700 psi at an intemal pressure of 1250 psig and temperature of 575oF; this is more than a factor of 1.5 below the yield strength of 42,300 psi at this temperature. At the pressure limit of 1375 psig, the general membrane stress increases to 29,400 psi, still safely below the yield strength. The reactor coolant system piping provides a comparable margin of protection at the established pressure safety limit. 2.2 BASES 22 Amendment No.104

P Table 3.1.1 - Continued e. The high drywell pressure scram functions in the Startup and Run modes when necessary during purging for containment inerting or de-inerting only by closing the manual containment isolation valves. Verification of the bypass condition shall be noted in the control room log. i f. One instrument channel for the functions indicated in the table to allow completion of surveillance testing, provided that: 1. Redundant instrument channels in the same trip system are capable of initiating the automatic function and are demonstrated to be operablo either immediately prior or immediately subsequent to applying the bypass. 2. While the bypass is applied, surveillance testing shall proceed on a continuous basis and the remaining instrument channels initiating the same function are tested prior to any other. Upon completion of surveillance testing, the bypass is removed. 8 3.1/4.1 31 l Amendment No.104 i

P Table 4.2.1 Minimum Test and Calibration Frequency for Core Cooling, Rod Block and Isolation Instrumentation instrument Channel Test (3) Calibration (3) Sensor Check (3) ECCS INSTRUMENTATION 1. Reactor Low-Low Water Level Once/3 months (Note 5) Every Operating Cycle - Transmitter Once/3 months -Trip Unit Once/12 hours l 2. Drywell High Pressure Once/3 months Once/3 months None 3. Reactor Low Pressure (Pump Start) Once/3 months Once/3 months None 4. Reactor Low Pressure (Valve Once/3 monihs Once/3 months None Permissive) 5. Undervoltage Emergency Bus Refueling Outage Refueling Outage None 6. Low Pressure Core Cooling Pumps Once/3 months Once/3 months None Discharge Pressure Interlock 7. Loss of Auxiliary Power Refueling Outage Refueling Outage None 8. Condensate Storage Tank Level Refueling Outage Refueling Outage None 9. Reactor High Water Level Once/3 nionths (Note 5) Every Operating Cycle - Transmitter Every 3 months - Trip Unit Once/12 hours l ROD BLOCKS 1. APRM Downscale Once/3 months (Note 5) Once/3 months None 2. APRM Flow Variable Once/3 months (Note 5) Once/3 months None 3. IRM Upscale Notes (2,5) Note 2 Note 2 4. IRM Downscale Notes (2,5) Note 2 Note 2 5. RBM Upscale Once/3 months (Note 5) Once/3 months None 6. RBM Downscale Once/3 months (Note 5) Once/3 months None 7. SRM Upscale Notes (2,5) Note 2 Note 2 8. SRM Detector Not-Full-in Position Notes (2,9) Note 2 None 9. Scram Discharge Volume-High Level Once/3 months Refueling Outage None MAIN STEAM LINE (GROUP 1) ISOLATION 1. Steam Tunnel High Temperature Refueling Outage Refueling Outage None 2. Steam Une High Flow Once/3 months Once/3 Months Once/12 hours l 3.2/4.2 61 Amendment No. 2,40,37,39,63,66,Si t&3, 104

o 4 Table 4.2.1 Continued Minimum Test and Calibration Frequency for Core Cooling, Rod Block and Isolation Instrumentation Instrument Channel Test (3) Galibration (3) Sensor Check (3) 3. Steam Line Low Pressure Once/3 months Once/3 months None 4. Reactor Low Low Water Level Once/3 months (Note 5) Every Operating Cycle-Transmitter Once/12 hours l Once/3 Months-Trip Unit

j CONTAINMENT ISOLATION (GROUPS 2 & 3) j 1.

Reactor Low Water Level (Note 10) l 2. Drywell High Pressure (Note 10) HPCI (GROUP 4) ISOLATION 1. Steam Line High Flow Once/3 months Once/3 months None 2. Steam Une High Temperature Once/3 months Once/3 months None RCIC (GROUP 5) ISOLATION 1. Steam Line High Flow Once/3 months Once/3 months None 2. Steam Line High Temperature Once/3 months Once/3 months None REACTOR BUILDING VENTILATION & STANDBY GAS TREATMENT 1. Reactor Low Low Water Level Once/3 months (Note 5) Every Operating Cycle - Transmitter Once/12 hours l Once/3 months - Trip Unit 2. Drywell High Pressure (Note 10) 3. Radiation Monitors (Plenum) Once/3 months Once/3 months Once/ day 4. Radiation Monitors (Refueling Floor) Once/3 months Once/3 months Note 4 RECIRCULATION PUMP TRIP AND ALTERNATE ROD INJECTION 1. Reactor High Pressure Once/3 months (Note 5) Once/ Operating Cycle-Transmitter Once/ Day i Once/3 Months-Trip Unit 2. Reactor low Low Water Level Once/3 months (Note 5) Once/ Operating Cycle-Transmitter Once 12 hours l Once/3 Months-Trip Unit SHUTDOWN COOLING SUPPLY 1:iOLATION 1. Reactor Pressure interlock Once/3 months Once/3 Months None 3.2/4.2 62 Amendment No. 74,84,83, % 103, 104

P Table 4.2.1 Continued Minimum Test and Calibration Frequency for Core Cooling, Rod Block and Isolation Instrumentation NOTES: ) (1) (Deleted) (2) Calibrate prior to normal shutdown and start-up and thereafter check once per 12 hours and test once per week until no longer l required. Calibration of this instrument prior to normal shutdown means adjustment of channel trips so that they correspond, within acceptable range and accuracy, to a simulated signal injected into the instrument (not primary sensor). In addition, IRM gain adjustment will be performed, as necessary, in the APRM/lRM overlap region. (3) Functional tests, calibrations and sensor checks are not required wherc the systems are not required to be operable or are tripped. If tests are missed, they shall be performed prior to returning the systems to an operable status. (4) Whenever fuel handling is in process, a sensor check shall be performed once per 12 hours. l (5) A functional test of this instrument means the injection of a simulated signal into the instrument (not primary sensor) to verify the proper instrument channel response alarm and/or initiating action. (6) (Deleted) (7) (Deleted) (8) Once/ shutdown if not tested during previous 3 month period. (9) Testing of the SRM Not-Full-in rod block is not required if the SRM detectors are secured in the full-in position. (10) Uses contacts from scram system. Tested and calibrated in accordance with Tables 4.1.1 and 4.1.2. 3.2/4.2 63a Amendment No. 30,63,83, 104

a l ~ j Bases 3.2 (Continuedh

1-increases core voiding, a negative reactivity feedback. High pressure sensors initiate 'he pump trip in the event of an isolation -

transient. Low level sensors initiate the trip on loss of feedwater (and the resulting MSIV closure). The recirculation pump trip is. only required at high reactor power levels, where the safety / relief valves have insufficient capacity to relieve the steam which continues to be generated after reactor isolation in this unlikely postulated event, requiring the trip to be operable only when in the RUN mode is therefore conservative. The ATWS high reactor pressure and low-low water levellogic also initiates the Alternate Rod Injection System. Two solenoid valves are installed in the scram air header upstream of the hydraulic control units. Each of the two trip systems energizes a valve - i

ii to vent the header and causes rod insertion. This greatly reduces the long term consequences of an ATWS event.

Voltage sensing relays are provided on the safeguards bus to transfer the bus to an alternate source when a loss of voltage t condition or a degraded voltage condition is sensed. On loss of voltage this transfer occurs immediately. The transfer on degraded voltage has a time delay to prevent transfer during the starting of large loads. The degraded voltage setpoint corresponds to the minimum acceptable safeguards bus voltage for a steady state LOCA load that maintains adequate voltage at the 480V essential MCCS. An allowance for relay tolerance is included. Safety / relief valve low-low set logic is provided to prevent any safety / relief valve from opening when there is an elevated water leg in the respective discharge line. A high water leg is formed immediately following valve closure due to the vacuum formed when [ steam condenses in the line. If the valve reopens before the discharge line vacuum breakers act to return water level to normal,- j water clearing thrust loads on the discharge line may exceed their design limit. The logic reduces the opening setpoint and increases the blowdown range of three non-APRS valves following a scram. A 15-second interval between subsequent valve 3 actuations is provided assuming one valve fails to i 3 I I i 3.2 BASES 69 Amendment No. 30,34,45,104

o Bases 4.t The instrumentation in this section will be functionally tested and calibrated at regularly scheduled intervals. Although this l instrumentation is not generally considered to be as important to plant safety as the Reactor Protection System, the same design reliability goals are applied. As discussed in Section 4.1 Bases, monthly or quarterly testing is generally specified unless the - testing must be conducted during refueling outages. Quarterly calibration'is specified unless the calibration must be conducted s during refueling cutages. Where applicable, sensor checks are specified on a once/12 hours or once/ day basis. _l i t I t i t t i ~ i i i t ? i i NEXT PAGE IS 76 Amendment No. 63,84,104 l L

F 3.0 LIMITING CONDITIONS FOR OPE 9ATION 4.0 SURVEILLANCE REQUIREMENTS Any four rod group may contain a control rod which is valved out of service provided the above requirements and Specification 3.3.A are met. D. Control Rod Accumulators D. Control Rod Accumulators Once per 12 hours check the status in the control room of the required Operable accumula!or pressure an Control rod accumulators shall be operable in the Startup, Run, or Refuel modes except as provided below. alarms. 1. In the Startup or Run Mode, a rod accumulator may be inoperable provided that no other control rod in the nine-rod square array around this rod has a: (a) Inoperable accumulator, or (b) Directional control valve electrically disarmed while in a non-fully inserted position. If a contro! rod with an inoperable accumulator is inserted " full-in" and its directional control valves are electrically disarmed, it shall not be considered to have an inopere. ole accumulator. 3.3/4.3 82 Amendment Nc. 5,44,43,54,63 104

e Bases 3.3/4.3 (Continuedh consequences of reactivity accidents are functions of the initial neutron flux. The requirement of at least 3 counts per second 8 assures that any transient, should it occur, begins at or above the initial value of 107 of rated power used in the analyses of transients from cold conditions. One operable SRM channel would be adequate to monitor the approach to criticality using homogeneous patterns of scattered contro! rod withdrawal. A minimum of two operable SRM's are provided as an added conservatism. C. Scram insertion Times The control rod system is designed to bring the reactor subcritical at a rate fast enough to prevent fuel damage; i.e., to prevent the MCPR from becoming less than the Safety Limit (T.S.2.1.A). This requires the negative reactivity insertion in any local region of the core and in the overal! core to be equivalent to at least the scram reactivity curve used in the transient analysis. The required i average scram times for three control rods in all two by two arrays and the required average scram times for all control rods are based on inserting this amount of negative reactivity at the specified rate locally and in the overall core. Under these conditions, the CPR safety limit is never exceeded during any transient requiring control rod scram, and therefore MCPR remains above the Safety l Limit (T.S.2.1.A). L 3.3/4.3 BASES 89 Amendment No. 29,104 i

9 Bases 3.4/4.4: A. The design objective of the standby liquid control system is to provide the capability of bringing the reactor from full power to a cold, xenon-free shutdown assuming that none of the withdrawn control rods can be inserted. To meet this objective, the liquid control system is designed to inject a quantity of boron which produces a concentration of bcron in the reactor core in less than 125 minutes sufficient to bring the reactor from full power to a 3% delta k subcritical condition considering the hot to cold reactivity swing, xenon poisoning and an additional 25% boron concentration margin to allow for leakage and imperfect mixing. l The time requirement (125 minutes) for insertion of the boron solution was selected to override the rate of reactivity insertion due to cooldown of the reactor following the xenon poison peak. The ATWS Rule (10 CFR 50.62) requires the addition of a new design requirement to the generic SLC System design basis. Changes to flow rate, solution concentration or boron enrichment to meet the ATWS Rule do not invalidate the original system design basis. Paragraph (c)(4) of 10 CFR 50.62 states that: "Each boiling water reactor must have a Standby Liquid Control System (SLCS) with a minimum flow capacity and boron content equivalent in control capacity to 86 gallons per minute of 13 weight percent sodium pentaborate solution" (natural boron enrichment). The described minimum system parameters (equivalent to 24 gpm,10.7% concentration and 55 atom percent Boron-10 enrichment) will ensure an equivalent injection capability that meets the ATWS rule requirement. Boron enrichment concentration, solution temperature, and volume (including check of tank heater and pipe heat tracing system) are checked on a frequency to assure a high reliability of operation of the system should it ever be required. Only one of the two standby liquid control pumping circuits is needed for proper operation of the system. If one pumping circuit is found to be inoperable, there is no immediate threat to shutdown capability, and reactor operation may continue while repairs are being made. A reliability analysis indicates that the plant can be operated safely in this manner for ten days. For additional margin, the allowable out of service time has been reduced to seven days. The only practical time to test the standby liquid control system is during a refueling outage cad by initiation from local stations. Components of the system are checked periodically as described above and make a functional test of the entire system on a frequency of less than once each refueling outage unnecessary. A test of explosive charges from one manufacturing batch is made to assure that the replacement charges for the tested system are satisfactory. A continual check of the firing circuit continuity is provided by pilot lights in the control room. The relief valves in the standby liquid control system protect the system piping and positive displacement pumps which are nominally designed for 1500 psi from overpressure. The pressure relief valves discharge back to the standby liquid control solution tank. 3.4/4.4 BASES 99 Amendment No. 56,57,77,104

3.0 LIMITING CONDITIONS FOR OPERATION. - 4.0 SURVEILLANCE REQUIREMENTS i 3. One of the following conditions of inoperability may 4. Perform the following tests: exist for the period specified: Regt - Freauency a. One Core Spray subsystem may be inoperable for 7 days, or Motor Operated Pursuant to i Valve Operability.- Specification

L b.

One RHR pump may be inoperable for 30 days, 4.15.8 or =i ADS Valve Each Operating i c. One low pressure pump or valve (Core Spray or Operability Cycle c RHR) may be inoperable with an ADS valve Note: Safety / relief valve operability is verified by

i inoperable for 7 days, or cycling the valve and observing a compensating d.

One of the two LPCI injection paths may be change in turbine bypass or control valve position. l r !I inoperable for 7 days, or ADS Inhibit Each Operating -{ e.- Two RHR pumps may be inoperable for 7 days, Switch Operability Cycle or Perform a simulated Each Operating automatic actuation test Cycle f. Both of the LPCI injection paths may be inoperable for 72 hours, or (including HPCI transfer to i the suppression pool and g. HPCI may be inoperable for 14 days, provided automatic restart on RCIC is operable, or subsequent low reactor [ water level) [ h. One ADS valve may be inoperable for 14 days, - [ or 5. Perform the following test on the Core Spray Ap l Instrumentation: i. Two or more ADS valves may be inoperable for 12 hours. Check Once/ day i 4. If the requirements or conditions of 3.5.A.1,2 or 3 Test Once/ month cannot be met, an orderly shutdown of the reactor Calibrate Once/3 months shall be initiated and the reactor shall be placed in a l condition in which the affected equipment is not i required to be operable within 24 hours. E I 3.5/4.5 102 i Amendment No. 7J,79, 104 [ t + m

3.0 LILilTING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS 2. (a) The reactor coolant water shall not exceed the 2. During startup and at steaming rates below 100,000 following limits with steaming rates less than pounds per hour, a sample of reactor coolant shall 100,000 pounds per hour except as specified in be taken every four hours and analyzed'ar 3.6.C.2.b. conductivity and chloride content. Conductivity 5 pmho/cm Chloride ion 0.1 ppm (b) For reactu startups the maximum value for I conductivity shall not exceed 10 pmho/cm and the maximum value for chloride ion concentration shall not exceed 0.1 ppm for the first 24 hours after placing the reactor in the power operating condition. 3. Except as specified in 3.6.C.2.b above, the reactor 3.(a) With steaming rates greater than or equal to coolant water shall not exceed the following limits 100,000 lbs. per hour, a reactor coolant sample with steaming rates greater than or equal to shall be taken at least everf 96 hours and when the 100,000 lbs. per hour. continuous conductivity monitors indicate abnormal conductivity (other than short-term spikes) and Conductivity 5 mho/cm analyzed for conductivity and chloride ion content. Chloride ion 0.5 ppm en condnuous con &My monnons 4. If Specifications 3.6.C.1 through 3.6.C.3 are not in Perable, during power operation, a reactor met, an orderly shutdown shall be initiated and the c ! nt sample should be taken once per 12 hours l reactor shall be in the cold shutdown condition an ana e r con &cMy and cNoMe,on i within 24 hours. content. 3.6/4.6 125 Amendment No.104

3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS D. Coolant Leakage ' D. Coolant Lea *Kage 1. Any time irradiated fuel is in the reactor vessel and 1. Any time irradiated fuel is in the reactor vessel and coolant temperature is above 212oF, reactor coolant coolant temperature is above 2120F, the following system leakage, based on sump monitoring, shall surveillance program shall be carried out; be limited to: c Unidentified and Identified Leakage rates chall a. 5 gpm Unidentified Leakage I be recorded once per 12 hours using primary [ b. 2 gpm increase in Unidentified Leakage within containment floor and equipment drain sump any 24 hour period monitoring equioment. c. 20 gpm identified Leakage 2. The reactor coolant system leakage detection d. no pressure boundary leakage systems shall be demonstrated OPERABLE by: 2. With reactor coolant system leakage greater than a. Primary containment atmosphere particulate 3.6.D.1.a or 3.6.D.1.c above, reduce the leakage monitoring systems-performance of a sensor rate to within acceptable limits within four hours or check once per 12 hours, a channel functional l initiate an orderly shutdown of the reactor and test at least monthly and a channel calibration reduce reactor water temperature to less than at least once per cycle. 212oF within 24 hours. b. Primary containment sump leakage 3. With an increase in Unidentified Leakage in excess measurement system-performance of a sensor of the rate specified in 3.6.D.1.b, identify the source check once per 12 hours and a channel l of increased leakage within four hours or initiate an calibration test at least once per cycle. orderly shutdown of the reactor and reduce reactor water tempe'rature to less than 212 F within 24 i hours. 4. If any Pressure Boundary Leakage is detected when the corrective actions outlined in 3.6.D.2 and 3.6.D.3 above are taken, initiate an orderly shutdown of the reactor and reduce reactor water temperature to less than 212oF within 24 hours. i 3.6/4.6 126 h Amendment No.45,47,87,104

-:mummm umum.it-I 3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS 4. Pressure Suppression Chamber-Drywell Vacuum 4. Pressure Suppression Chamber-Drywell Vacuum Breakers Breakers l When primary containment integrity is required, a. a. Operability and full closure of the all eight drywell-suppression chamber vacuum drywell-suppression chamber vacuum breakers breakers shall be operable and positioned in shall be verified by performance of the the closert position as indicated by the position following: indication system, except during testing and except as specified in 3.7.A.4.b through (1) Monthly each operable drywe!I-suppression 3.7.A.4.d below. chamber vacuum breaker shall be exercised through an opening-closing 3 b. Any drywell-suppression chamber vacuum cycle. breaker may be,onfully closed as indicated by the position inc%ation and alarm sistem (2) Once each operating cycle, drywell to provided thr.i drywell to suppression chamber suppression chamber leakage shall be differentir.i pressure decay does not exceed demonstrated to be less than that that shcwn on Figure 3.7.1 equivalent to a one-inch diameter orifice and each vacuum breaker sha!i be visually c. Up to two drywell-suppression chamber vacuum inspected. (Containment access required) breakers may be inoperable provided that: (1) the vacuum breakers are determined to be fully (3) Once each operating cycle, vacuum closed and at least one position alarm circuit is breaker position indication and alarm operable or (2) the vacuum breaker is secured systems shall be calibrated and functionally in the closed position or replaced by a blank tested. (Containment access required) flange. (4) Once each operating cycle, the vacuum d. Drywell-suppression chamber vacuum breakers breakers shall be tested to determine that may be cycled, one at a time, during the force required to open each valve from containment inerting and deinerting opeiations fully closed to fully open does not exceed to assist in purging air or nitrogen from the that equivalent to 0.5 ps, acting on the suppression chamber vent header. suppression chamber face of the valve disc. (Containment access required.) 3.7/4.7 164 Amendment No. 8,36,80, 104 _ o

Bases 4.7 (Continuedk B. Standby Gas Treatment System, and C. Secondary Containment initiating reactor bulkiing isolation and operation of the standby gas treatment system to maintain the design negative pressure within the secondary containment provides an adequate test of the reactor building isolation valves and the standby gas treatment system. Periodic testing gives sufficient confidence of reactor building integrity and standby gas treatment system operational capability. Secondary Containment Capability Test data obtained under non-calm conditions is to be extrapolated to calm wind conditions using information provided in " Summary Technical Report to the United States Atomic Energy Commission, Directorate of Licensing, on Secondary Containment Leak Rate Test", submitted by letter dated July 23,1973, and as described in NSP letter to the NRC dated August 18,1995, with subject," Revision 2 to License Amendment Request Dated June 8,1994, Standby Gas Treatment and Secondary Containment Technical Specifications." The frequency of tests and sample analysis are necessary to show that the HEPA filters and charcoal adsorbers can perform as evaluated. Standby gas treatment system inplace testing procedures will be established utilizing applicable sections of ANSI N510-1989 standard as a procedural guideline only. If painting, fire, or chemical release occurs such that the HEPA filter or charcoal adsorber could become contaminated from the fumes, chemicals, or foreign materials, the same tests and sample analysis should be performed as required for operational use. Replacement adsorbent should be qualified according to the guidelines of Regulatory Guide 1.52 Revision 2 (March 1978) except testing should be IAW D3803-1989. The charcoal l adsorber efficiency test procedures will allow for the removal of a representative sample. The 30 C,95% relative humidity test per ASTM D 3803-89 is the test method to establish the methyl iodine removal efficiency of adsorbent. The sample will be at least two inches in diameter and a length equal to the thickness of the bed. If the iodine removal efficiency test results are unacceptable, all adsorbent in the system will be replaced. High efficiency particulate filters are installed before and after the charcoal filters to prevent clogging of the carbon adsorbers and to minimize potential release of particulates to the environment. An efficiency of 99% is adequate to retain particulates that may be released to the reactor building following an accident. This will be demonstrated by inplace testing with DOP as the testing medium. Any HEPA filters found defective will be replaced with filters qualified pursuant to regulatory guide position C.3.d of Regulatory Guide 1.52 Revision 2 (March 1978). Once per operating cycle demonstration of HEPA filter pressure drop, operability of inline heaters at rated power, automatic initiation of each standby gas treatment system circuit, and leakage tests after maintenance or testing which could affect leakage, is necessary to assure system performance capability. 4.7 BASES 188 Amendment No. 94,104 1

Bases 4.7 (Continued): The containment is penetrated by a large number of small diameter instrument lines. A program for the periodic testing (see Specification 4.7.D) and examination of the valves in these lines has been developed and a report covering this program was submitted to the AEC on July 27,1973. l The main steam line isolation valves are functionally tested on a more frequent interval to establish a high <legree of reliability. E. Combustible Gas Control System The Combustible Gas Control System (CGCS) is functionally tested once every six months to ensure that the recombiner trains i 3 will be available if required. In addition, calibration and maintenance of essential components is specified once each operating cycle. t I 4.7 BASES 190 l Amendment No. 35, 104 m e een.=

s a TABLE 4.8A - RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM (Continued) (Page 2 of 2) Notes: a. The LLD is the smallest concentration of radioactive material in a sample that will be detected with 95% probability with 5% probability of falsely concluding that a blank observation represents a "real" signal. Note (a) of Table 4.8.3 is applicable. b. Grab samples taken at the discharge of tne plant stack and reactor building vent are generally below minimum detectable levels for most nuclides with existing analytical equipment. For this reason, isotopic analysis data, corrected for holdup time, for samples taken at the steam jet air ejector may be used to calculate noble gas ratios. c. Whenever the steady state radioiodine concentration is greater than 10 percent of the limit of Specification 3.6.C.1, daily samp!ing of reactor coolant for radioactive iodines of I-131 through I-135 is required. Whenever a change of 25% or more in calculated Dose Equivalent 1-131 is detected under these condit ons, the iodine and particulate collection devices for all release points shall be removed and analyzed daily until it is shown that a pattem exists which can be used to predict the release rate. Sampling may then revert to weekly. When samples collected for one day are analyzed, the corresponding LLD's may be increased by a factor of 10. Samples shall be analyzed within 48 hours after removal. d. To be representative of the average quantities and concentrations of radioactive materials in particulate form in gaseous effluents. samples should be collected in proportion to the rate of flow of the effluent streams. e. The principal gamma emitters for which the LLD specification will apply are exclusively the following radionuclides: Kr-87, Kr-88, l Xe-133, Xe-133m, Xe-135, and Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144 for particulate emissions. 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. f. Nuclides which are below the LLD for the analyses shall be reported as "less than" the LLD of the nuclide and should not be reported as being present at the LLD leve! for that nuclide. The "less than" values shall not be used in the required dose calculations. When unusual circumstances result in LLD's higher than reported, the reasons shall be documented in the semiannual effluent report, g. The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time period sampled. h. H3 analysis shall not be required prior to purging if the limits of 3.8.B.1 are satisfied for other nuclides. However, the H3 analysis shall be completed within 24 hours after sampling. i. In lieu of grab samples, continuous monitoring with bi-weekly analysis using silica-gel sampleis may be provided. l 3.8/4.8 198t Amendment No. 45, 90, 104

9 3.0 LIMITING CONDmONS FOR OPERATION 4.0 SURVEILLANCE HEQUIREMENTS - 2. Both diesel generators are operable and capable of feeding their designated 4160 voit buses. 3.(a) 4160V Buses #15 and #16 are energized. (b) 480V Load Centers #103 and #104 are energized. 4. All station 24/48,125, and 250 volt batteries are charged and in service, and associated battery chargers are operable. B. When the mode switch is in Run, the availability of electric pcwer shall be as specified in 3.9.A, except as l specified in 3.9.B or the reactor shall be placed in the cold shutdown condition within 24 hours. 1. Transmission Lines From and after the date that incoming power is available from only one line, reactor operation is permissible only during the succeeding seven days unless an additional line is sooner placed in i 3.9/4.9 200 l Amendment No. 54, 104

e 3.0 LIMITING CONDITIONS FOR CPERATION 4.0 SURVEILLANCE REQUIREMENTS 3.13 FIRE DETECTION AND PROTECTION SYSTEMS 4.13 FIRE DETECTION AND PROTECTION SYSTEMS Acclicability: Acolicability: Applies to instrumentation and plant systems used for fire Applies to the periodic testing of instrumentation and plant l detection and protection of the nuclear safety-related systems used for fire detection and protection of the nuclear l structures, systems, and components of the plant. safety related structures, systems, and components. Obiective: Objective: To insure that the structures, systems, and components of To verify the operability of instrumentation and plant systems the plant important to nuclear safety are protected from fire used for fire detection and protection of nuclear safety damage. related structures, systems, and components. Specification: Soecification: A. Fire Detection Instrumentation A. Fire Detection Instrumentation 1. Except as specified below, the minimum fire 1. Fire uetection instrumentation in each of the zones detection instrumentation for each fire detection in Table 3.13.1 shall be demonstrated operable zone shown in Table 3.13.1 shall be operable every six months by performance of functional tests. whenever equipment in that fire detection zone is required to be operable. 2. Alarm circuitry associated with the fire detector j instruments in each of the zones in Table 3.13.1 2. If specification 3.13.A.1 cannot te met, within one shall be demonstrated operable every six months. hour establish a fire watch patrol to inspect the zone (s) with inoperable instruments once per hour (+ 25%). Restore the minimum number of instruments to operable status within 14 days or submit a special report to the Commission within 30 days outlining the cause of the inoperability and the plans and schedule for restoring the instruments to operable status. 3.13/4.13 223 Amendment No. 7,46, 104

~ 3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS F. Halon Systems F. Halon Systems 1. The cable spreading room Halon system shall 1. The cable spreading room Halon system shall be be operable with the storage tanks having demonstrated operable as follows: l at least 95% of full charge weight and 90% of full a. Each valve (manual, power operated, or charge pressure. automatic) in the flow path that is not electrically supervised, locked, sealed or otherwise 2. If specification 3.13.F.1 cannot be met, within one s ured in position, shall be verified to be,in its hour establish a continuous fire watch with backup rrect position every month. fire suppression equipment in the cable spreading room. Restore the system to operable status within b. Verify Ha!on storage tank weight and pressure 14 days or submit a special report to the every six months. Commission within 30 days outlining the cause of c. Perform a system f'metional test every 18 the inoperability and the plans and schedule for months which. includes verifying the system, restoring the system to operable status. includ,ng associated ventilation dampers, i G. Penetration Fire Barriers actuates manually and automatically, upon receipt of a test signal. 1. All penetration fire barriers in fire area boundan.es shall be operable whenever safe shutdown d. Perform an air flow test every 3 years through equipment in that fire area is required to be headers and nozzles to assure no blockage. operable. e. Visually examine headers and nozzles every 18 2. 11 Specification 3.13.G.1 cannot be met, a months. An air flow test shall be performed continuous fire watch shall be established on at upon evidence of obstructions of any Halon least one side of.he affected penetration (s) within system nozzle. l one hour or verify the operability of fire detectors on G. Penetration Fire Barriers at least one side of the non-functional fire barrier l and establish an hourly (+ 25%) fire watch patrol. 1. A visualinspection of penetration fire barriers in fire Restore the inoperable penetration fire barriers to area boundaries protecting safe shutdown Operable status within 14 days or submit a special equipment shall be conducted every 18 months. report to the Commission within 30 days outlining the cause of the inoperability and the plans and 2. Following repair or maintenance of a penetration fire I schedule for restoring the barriers to Operable barrier a visual inspection of the seal shall be status. conducted. 1 3.13/4.13 227b Amendment No. 7,46,61, 104 I

h 3.0 LIMITING CONDITIONS FOR OPERATION '4.0 SURVEILLANCE REQUIREMENTS 4 H. Altemate Shutdown System H. Altemate Shutdown System '1. The system controls on the ASDS panel shall be 1. Switches on the altemate shutdown system panel operable whenever that system / component is shall be functionally tested once per operating required to be operable. 7 cycle. l 2. If system controls required to be operable by -I Specification 3.13.H.1 are made or found 2. The alternate shutdown system panet master 'i inoperable, restore the inoperable system control to transfer switch shall be verified to alami in the. l} operable within 7 days, or perform one of the control room when unlocked once per operating - l following; cycle. ) a. Provide equivalent shutdown capability and within 60 days restore the inoperable system j controls to cperable; or i b. Establish a continuous fire watch in the cable spreading room and the back-panel area of the i - control room and within 60 days restore the inoperable system controls to operable; or c. Verify the operability of the fire detectors in the - cable spreading room and the back-panel area of the control room and establish a hourly fire l i watch patrol and within 60 days restore the E inoperable system controls to operable; or f d. Place the reactor in a condition where the systems for which the system controls at the ASDS are inoperable are not required to be operable within 24 hours. l l 3. The alternate shutdown system panel master transfer switch shall be locked in the normal position except when in use, being tested or being - i maintained. 4 t 3.13/4.13 227c Amendment No. 47,64, 104 i I i

^ TABLE 3.13.1 SAFETY RELATED FIRE DETECTION INSTRUMENTS Fire Detection iocation Minimum Instruments Operable l Zone Heat Flame Smoke 1A "B" RHR Room 3 1B "A" RHR Room 3 1C RCIC Room 3

j 1E HPCI Room 2

1F Reactor Building-Torus Compartment _ 11 1 2A Reactor Bldg.935* elev-TIP Drive Area 1 28 Reactor Bldg. 935' elev - CRD HCU Area East 10 2C Reactor Bldg. 935' elev - CRD HCU Area West 11 2G/2H Reactor Bldg. 935'- LPCI Injection Valve Area 1 l 3B Reactor Bldg. 962' elev - SBLC Area 2 3C Reactor Bldg. 962' elev - South 5 3D Reactor Bldg. 962' elev - RBCCW Pump Area 4 4A Reactor Bldg. 985' elev - South 4 4B Reactor Bldg. 985' elev - RBCCW Hx Area 5 4D SBGT System Room 2 SA Reactor Bldg.1001' elev - South 7 5B Reactor Bldg.1001' elev - North 3 SC Reactor Bldg. - Fuel Pool Cooling Pump Area 1 6 Reactor Building 1027' elev 5 7A Battery Room 1 7B Battery Room 1 7C Battery Room 1 8 Cable Spreading Room 7 227d 3.13/4.13 Amendment No. 64, 104

. e: i 2 .. TABLE 3.13.1 (Continued) SAFETY RELATED FIRE DETECTION INSTRUMENTS Fire Detection Location Minimum Instruments Operable ~l' i- _Z_qng Heat Flame - Smoke . j. I { 12A Turbine Bldg. - 911' - 4.16 KV Switchgear 3 f 13C - Turbine Bldg. - 911' elev - MCC 133 Area .1 l 14A Turbine Bldg. - 931'- 4.16 KV Switchgear 2-j j 15A/15C

1. 12 DG Room & Day Tank Room 3

i f. 15B/15D

1. 11 DG Room & Day Tank Room 3

j 16 Turbine Bldg. 931' elev - Cable Corridor. 3 17 Turbine Bldg. 941' elev - Cable Corridor - 3 j T 19A Turbine Bldg. 931' elev - Water Treatment Area 5 j 19B Turbine Bldg. 931' elev - MCC 142-143 Area ' 1 19C Turbine Bidg. 931' elev - FW Pipe Chase 1 20 Heating Boiler Room 1 23A Intake Structure Pump Room - 3 [ 31A 1st Floor - Reactor Building Addition - Division i 3 31B 1st Floor - Reactor Building Addition - Division li 15 32A 2nd Floor - Reactor Building Addition - Division i 6 328 2nd Floor - Reactor Building Addition - Dwision 11 4 33 3rd Floor - Reactor Building Addition 5 227e 3.13/4.13 . Amendment No. 64, 104 .m .m. f ..m . m. i . m = . ~ +

? Table 3.14.1 i Instrumentation for Accident Monitoring r Function Total No. of Minimum No. of. Required - Instrument Channels Operable Channels Conditions

• Reactor Vessel Fuel Zone Water Level 2

1 A, B ,j Safety / Relief Valve Position ' 2 1 A, C (One Channel Pressure Switch and One Channel 'f Thermocouple Position Indication per Valve) Dryweli Wide Range Pressure 2 1 A, B j Suppression PoolWde Range Level 2 1 A, B Suppression Pool Temperature 2 1 A, D - 1 I Drywell High Range Radiation 2 1. A, D Drywell and Suppression Pool 2 1 A, B Hydrogen and Oxygen Monitor Offgas Stack Wda Range Radiat;an 2 1 A, D Reactor Bldg Vent Wide Range Radiation 2 1 A, D ,t

• Required Conditions j

A. When the number of channels made or found to be inoperable is such that the number of operable channels is less than the total number of channels, either restore the inoperable channels to operable status within seven days, or prepare and submit a special report to the Commission pursuant to Technica! Specification 6.7.D within the next 30 days outlining the action taken, the cause of l i the inoperability, and the plans and schedule for restoring the system to operable status. j t e f 3.14/4.14 229b ~ Amendment No. 2,37,63, 104 f

Table 3.14.1 (Continued) Instrumentation for Accident Monitoring '

• Required Conditions (continued)

B. When the number of channels made or found to be inoperable is such that the number of operable channels is less than the minimum number of operable channels shown, the minimum number of channels shall be restored to operable status within 48 hours .I or be in at least Hot Shutdown within the next 12 hours and Cold Shutdown within the following 24 hours. l ll C. When the number of channels made or found to be inoperable is such that the number of operable channels is less than the minimum number of operable channels shown, the torus temperature shall be monitored once per 12 hours (+25%) to observe any_ l unexplained temperature increase which might be indicative of an open SRV; the minimum number of channels shall be restored to operab's status within 30 days or be in at least Hot Shutdown within the next 12 hours and Cold Shutdown within the following 24 hours. D. When the number of channels made or found to be inoperable is such that the number of operable channels is less than the minimum number of operable channels shown, initiate the preplanned altemate method of monitoring the appropriate parameters in addition to submitting the report required in (A) above. F 3.14/4.14 229c Amendment No. 3,37,63, 104 I

3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS .i .i B. Inservice Testina 1. Inservice Testing of Quality Group A, B, and C pumps and valves shall be performed in accordance with the requirements for ASME Code Class 1,2 and 3 pumps And valves, respectively, contained in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g) except where reliaf has been granted by the Commission pursuant to 10 CFR 50, Section 50.55(a)(g)(6)(i), or where altemate testing is justified in accordance with Generic Letter 89-04. 2. Nothing in the ASME Boiler and Pressure Vessel code shall be construed to supersede the requirements of any Technical Specification. 3.15/4.15 22S4 Amendment No. 6,37, -72. U, 104

3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS 3. Deviations are permitted from the required sampling schedule if samples are unobtainable due to hazardous conditions, seasonable unavailability, or to malfunction of automatic sampling equipment. If the latter occurs, every effort shall be made to complete corrective action prior to the end of the next sampling period. 4. With the level of radioactivity in an environmental sampling medium exceeding the reporting levels of Table 4.16.3 when averaged over any calendar quarter, submit a special report to the Commission within 30 days from the end of the affected calendar quarter pursuant to Specification 6.7.C.2. When l more than one of the radionuclides in Table 4.16.3 are detected in the sampling medium, this report shall be submitted if: concentration (1) + concentration (2) + " >1.0 limit level (1) limit level (2) When radionuclides other than those in Table 4.16.3 are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose to an individual is equal to or greater than the calendar year limits of Specifications 3.8.A.2, 3.8.B.2, or 3.8.B.3. This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiation Environmental Monitoring Report. j 3.16/4.16 229i [ Amendment No. 37,39,46, 104

o i 3.0 LIMITING CONDITIONS FOR OPERATION - 4.0 SURVEILLANCE REQUIREMENTS i 3.17 CONTROL ROOM HABITABILITY 4.17 CONTROL ROOM HABITABILITY f AndiceNiity: Aoolicability: Applies to the control room ventilation system equipment Applies to the periodic testing requirements of systems necessary to maintain habitability. required to maintain control room habitability. e aves: Obiectives: i To assure the control room is habitable both under normal and accident conditions. To verify the operability of equipment related to control room habitability. t Specification: Specification-1 A. Control Room Ventilation System i i a 1. Except as specified in 3.17.A.2 and 3.17.A.3 below, both A. Control Room Ventilation System 1 trains of the control room ventilation system shall be operable, whenever irradiated fuel is in the reactor 1. Once per 12 hours check control room l. vessel and reactor coolant temperature is greater than temperature' 2120F, or during movement of irradiated fuel assemblies in the secondary containment, core alterations or activities having the potential for draining the reactor vessel. 2.a With one control room ventilation train inoperable, restore the inoperable train to operable status within 30 days. 2.b If 2.a is not met, then be in hot shutdown within the next 12 hours following the 30 days and in cold shutdown within 24 hours following the 12 hours. 2.c if 2.a is not met during movement of irradiated fuel r assemblies in the secondary containment, core l' alterations or activities having the potential for draining the reactor vessel then immediately place the operable control room ventilation train in operation or immediately suspend these activities. 3.17/4.17 229u f Amendment No. 66,89, 104

6.0 ADMINISTRATIVE CONTROLS 6.1 Organization A. The Plant Manager shall be responsible for overall unit safe operation and shall have control over those onsite activities necessary for the safe operation and maintenance of the plant. During periods when the Plant Manager is unavailable, this responsibility may be delegated to other qualified supervisory personnel. The Shift Supervisor (or, a designated individual during periods of absence from the control room and shift supervisor's l office) shall be responsible for the control room command function. I B. Offsite and Onsite Organizations Onsite and offsite organizations shall be established for plant operation and corporate management, respectively. The onsite and offsite organizations shall include positions for activities affecting plant safety. 1. Lines of authority, responsibility and communication shall be established and defined for the highest management levels through intermediate levels to and including all operating organization positions. These relationships shall be documented and updated, as appropriate, in the form of organization charts, function descriptions of department responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of l documentation. These requirements are documented in corporate and plant procedures, or the Updated Safety Analysis Report or the Operational Quality Assurance Plan. t 2. The President, NSP Nuclear Generation shall have corporate responsibility for overall plant nuclear safety and shall l t take any measures needed to ensure acceptable performance of the staff in operating, maintaining and providing technical support to the plant to ensure nuclear safety. This position has the responsibility for the Fire Protection l Program. 3. The individuals who train the operating staff and those who carry out health physics and quality assurance functions may report to the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their independence from operating pressures. i 6.1 232 l Amendment No. 7,64,68, 104 i i ?

C. Plant Siaff 1. Each on duty shift shall be composed of at least the minimum shift crew composition shown in Table 6.1.1. 2. At least one licensed operator shall be in the control room when fuel is in the reactor. 3. At least two licensed operators shall be present in the control room during cold startup, scheduled reactor shutdown, and during recovery from reactor trips. 4. An individual qualified in radiation protection procedures shall be onsite when fuel is in the reactor. l 5. All alterations of the reactor core shall be directly supervised by a licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation. 6. A fire brigade of at least five members shall be maintained onsite at all times.* The fire brigade shall not include the l three members of the shift organization required for safe shutdown of the reactor from outside the control room. 7. The General Superintendent, Operations shall be formerly licensed as a Senior Reactor Operator or hold a current Senior Reactor Operator License. 8. At least one member of plant management holding a current Senior Reactor Operator License shall be assigned to the plant operations group on a long term basis (approximately two years). This individual will not be assigned to a rotating shift. i D. Each member of the unit staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except for (1) the General Superintendent Radiation Services who shall meet or exceed the qualifications of l Regulatory Guide 1.8, September 1975, (2) the Shift Technical Advisor who shall have a bachelor's degree or equivalent in a scientific or engineering discipline with specific training in plant design, and response and analysis of the plant for transients and accidents, and (3) the General Superintendent, Operations who shall meet the requirement of ANSI l N18.1-1971 except that NRC license requirements are as specified in Specification 6.1.C.7. The training program shall be under the direction of a designated member of Northern States Power management. l Fire Brigade composition may be less than the minimum requirements for a period of time not to exceed 2 hours in order to accommodate unexpected absence of Fire Brigade members provided immediate action is taken to restore the Fire Brigade to within the minimum requirements. 6.1 233 Amendment No. 46,37, 68,104

l E. A training program for individuals serving in the fire brigade shall be maintained under the direction of a designated member of Northem States Power management. This program shall meet the requirement of Section 27 of the NFPA Code - 1976 with the exception of training scheduling. Fire brigade training shall be scheduled as set forth in the training program. F. Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safety-related functions; e.g., senior reactor operators, reactor operators, health physicists, auxiliary operators, and key maintenance personnel. Procedures shallinclude the following provisions: 1. Adequate shift coverage shall be maintained without routine heavy use of overtime. The objective shall be to have operating personnel work a normal 8 or 12-hour day, nominal 40-hour week while the plant is operating. However, in l the event that unforeseen problems require substantial amounts of overtime to be used, er during extended periods of shutdown for refueling, major maintenance or major plant modifications, on a temporary basis, the following guidelines shall be followed: a. An individual should not be permitted to work more than 16 hours straight, excluding shift tumover time. 5. Overtime should be limited for all nuclear plant staff personnel so that total work time does not exceed 16 hours in any 24-hour period, nor more than 24 hours in any 48-hour period, not more than 84 hours in any seven day period, all excluding shift tumover time. Individuals should not be required to work more than 15 consecutive days without two consecutive days off. c. A break of at least eight hours including shift tumover time should be allowed between work periods. d. Except during extended shutdown periods, the use of overtime should be considered on an individual basis and not for the entire staff on a shift. 6.1 234 Amendment No. 3,46,46,68,104 a

? t I l 6.2 Review and Audit Organizational units for the review and audit of facility operations shall be constituted and have the responsibilities and authorities outlined below-A. Safety Audit Committee (SAC) The Safety Audit Committea provides the independent review of plant operations from a nuclear safety standpoint. Audits of. plant operation are conducted under the cognizance of the SAC. l f 1. Membership f f F a. The SAC shall consist of at least five (5) persons. i. b. The SAC Chairman shall be an NSP representative, not having line responsibility for operation of the plant, appointed by the President, NSP Nuclear Generation. Other members shall be appointed by the President, NSP Nuclear l Generation or by such other person as he may designate. The Chairman shall appoint a Vice Chairman from the SAC membership to act in his absence. j c. No more than two members of the SAC shall be from groups holding line responsibility for operaticn of the plant. t'! d. A SAC member may appoint an attemate to serve in his absence, with concurrence of the Chairman. No more than one altemate shall serve on the SAC at any one time. The attemate member shall have voting rights. I 2. Qualifications I a. The SAC members should collectively have the capability required to review activities in the following areas: nuclear 1 power plant operations, nuclear engineering, chemistry and radiochemistry, metallurgy, instrumentation and control, j radiological safety, mechanical and electrical engineering, quality assurance practices, and other appropriate fields y associated with the unique characteristics of the nuclear power plant. [ t i 6.2 237 Amendment No. 3,46,46, 104 - l L

1 [ f i 1 f. Investigation of all Reportable Events and Events requiring Special Reports to the Commission. { g. Revisions to the Facility Emergency Plan, the Facility Security Plan, and the Fire Protection Program. h. Operations Committee minutes to determine if matters considered by that Committee involve unreviewed or unresolved safety questions. i. Other nuclear safety matters referred to the SAC by the Operations Committee, plant management or company management. F il j. All recognized indications of an unanticipated deficiency in some aspect of design or operation of safety-related l structures, systems, or components. l

j k.

Reports of special inspections and audits conducted in accordance with specification 6.3. I. Changes to the Offsite Dose Calculation Manual (ODCM). l m. Review of investigative reports of unplanned releases of radioactive material to the environs. [ 6. Audit - The operation of the nuclear power plant shall be audited formally under the cognizance of the SAC to assure safe - facility operation. a. Audits of selected aspects of plant operation, as delineated in ANSI N18.7-1976 as modified by the Operational Quality j Assurance Plan, shall be performed with a frequency commensurate with their nuclear safety significance and in a manner to assure that an audit of all nuclear safety-related activities is completed within a period of two years. The audits shall be performed in accordance with appropriate written instructions and procedures. j b. Audits of aspects of plant radioactive effluent treatment and radiological environmental monitoring shall be performed i as follows: l 1. Implementation of the Offsite Dose Calculation Manual and quality controls for effluent monitoring at least once every two years. 2. Implementation of the Process Control Program for solidification of radioactive waste at least once every two years. 3. The Radiological Environmental Monitoring Program and the results thereof, including quality controls, at least once every year. c. Periodic review of the audit program should by performed by the SAC at least twice a year to assure its adequacy. l t d. Written reports of the audits shall be reviewed by the President, NSP Nuclear Generation, by the SAC at a scheduled l meeting, and by members of Management having responsibility in the areas audited. 6.2 239 Amendment No. 46, 46, 46, 69, 104 ?

t 7. Authority The SAC shall be advisory to the President, NSP Nuclear Generation. l 8. Records Minutes shall by prepared and retained for all scheduled meetings of the Safety Audit Committee. ' The minutes shall be distributed within one month of the meeting to the President, NSP Nuclear Generation, the Plant Manager, each member of l the SAC, and others designated by the Chairman or Vice Chairman. There shall be a formal approval of the minutes. 9. Procedures A written charter for the SAC shall be prepared that contains: i a. Subjects within the purview of the group. r i b. Responsibility and authority of the group. c. Mechanisms for convening meeti= gs. _j d. Provisions of use of specialists or subgroups. e. Authority to obtain access to the nuclear power plant operating record files and operating personnel when assigned i audit functions. .i t f. Requirements for distribution of reports and minutes prepared by the group to others in the NSP Organization. 3 I I 4 6.2 240 Amendment No. 3,46,46, 104 i _m.,_ c -..: z. m s m.. m. -m. w

[ B. Operations Committee (OC) 1. Membership The Operations Committee shall consist of at least six (6) regular members drawn from the key supervisors of the onsite supervisory staff. The Plant Manager shall serve as Chairman of the OC and shall appoint a regular member to act as Vice Chairman in his absence. Attemates to the regular members shall be designated in writing by the Chairman, or Vice Chairman in the Chaliman's absence, to serve on a temporary basis. No more than two alternates shall participate as voting members of the Operations Committee at any one time. 2. Meeting Frequency The Operations Committee will meet on call by the Chairman or as requested by individual members and at least monthly. 3. Quorum A quorum shall include a majority of the membership, including the Chairman or Vice Chairman. 4. Responsibilities - The following subjects shall by reviewed by the Operations Committee: a. Proposed tests and experiments and their results. r b. Modifications to plant sy:;tems or equipment as described in the Updated Safety Analysis Report and having nuclear safety significance or which involve an unreviewed safety question as defined in 10 CFR 50.59. c. Proposals which would effect permanent changes to normal and emergency operating procedures and any other proposed changes or procedures that are determined by the Plant Manager to affect nuclear safety. d. Proposed changes to the Technical Specifications or operating license. { e. All reported or suspected violations of Technical Specifications, operating license requirements, administrative procedures, or operating procedures. Results of investigations, including evaluation and recommendations to prevent recurrence, will be reported, in writing, to the President, NSP Nuclear Generation and to the Chairman of l the Safety Audit Committee. 6.2 241 Amendment No. 3,47, 69, 104 l r r

~. g 3 f f. Investigation of all Reportable Events and Events requiring Special Reports to the Commission. g. Drills on emergency procedures (including plant evacuation) and adequacy of communication with off-site support groups. i h. All procedures required by these Technical Specifications, including implementing procedures of the Emergency L Plan and the Security Plan (except as exempted in Section 6.5.F), shall be reviewed with a frequency commensurate with their safety significance but at an interval of not more than two years. i. Perform special reviews and investigations, as requested by the Safety Audit Committee. l j. Review of investigative reports of unplanned releases of radioactive material to the environs. k. All changes to the Process Control Program (PCP) and the Offsite Dose Calculation Manual (ODCM). 5. Authority [ The OC Shall be advisory to the Plant Manager. In the event of disagreement between the recommendations of the OC and the Plant Manager, the course determined by the Plant Manager to be the more conservative will be followed. A written summary of the disagreement will be sent to the President, NSP Nuclear Generation and the Chairman of the l-SAC for review. i 6. Records i Minutes shall be recorded for all meetings of the OC and shallidentify all documentary material reviewed The minutes shall be distributed to each member of the OC, the Chairman and each member of the Safety Audit Committee, the President, NSP Nuclear Generation and others designated by OC Chairman or Vice Chairman. l [ 7. Procedures i A written charter for the OC shall be prepared that contains: I a. Responsibility and authority of the group. b. Content and method of submission of presentations to the Operations Committee. i 6.2 242 Amendment No. 45,25,46, 104 l 6 ,-_ _ _ _ _ _. _ _ _ _ _ _ _ _ = _ _ _ m

c. Mechanism for scheduling meetings d. Meeting agenda i e. Use of subcommittee f. Review and approval, by members, of OC actions

[

g. Distribution of minutes t 6.3 SoecialInspections and Audits A. An independent fire protection and loss prevention inspection and audit shall be performed annually utilizing either qua.'ified offsite Northem States Power Company personnel or an outside fire protection consultant. B. An inspection and audit by an outside qualified fire protection consultant shall be performed at intervals no greater than three years. 6.4 Action to t e Taken if a Safety Umit is Exceeded if a Safety Umit is exceeded, the reactor shall be shut down immediately. An immediate report shall be made to the Commission and i to the President, NSP Nuclear Generation or his designated alternate in his absence. A complete analysis of the circumstances l leading up to and resulting from the situation, together with recommendations by the Operations Committee, shall also by prepared. This report shall be sebmitted to the Commission, to the President, NSP Nuclear Generation and the Chairman of the Safety Audit l { Committee within 14 days of the occurrence. 1 Reactor operation shall not be resumed until authorized by the U.S. Nuclear Regulatory Commission. t 6.2 - 6.4 243 Amendment No. 3, 104 i

1. - ~_ + k I 1 B. Radioloalcal 1.a. A Radiation Protection Program, consistent with the requirements of 10 CFR 20, shall be developed and followed. The f 6 Radiation Protect!on Program shall consist of the following: (1) A Radiation Protection Plan, which shall be a complete definition of radiation protection policy and prograin ( (2) Procedures which implemer.1 the requirements of the Radiation Protection Plan The Radiation Protection Plan and implementing procedures, with the exception of those non-safety related procedures j goveming work activities exclusively applicable to or performed by health physics personnel, sha!I be reviewed by the

i Operations Committee and approved by a member of plant management designated by the Plant Manager. Health physics procedures not reviewed by the Operations Committee shall be reviewed and approved by the General Superintendent i'

Radiation Services. ~ l b. In lieu of the " control device" or " alarm signal

• required by paragraph 20.203(c)(2) of 10 CFR 20, each high radiation area in

( which the intensity of radiation is greater than 100 mrem /hr but less than 1000 mrem /hr shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controiled by requiring issuance of a Radiation. Work Permit.1 Any individual or group of individuals permitted to entar such areas shall be provided with or accompanied by i one or more of the following: [ (1) A radiation monitoring device that continuously indicates the radiation dose rate in the area. l (2) A radiation monitoring device that continuously integrates the radiation dose rate in the area and alarms when a preset I integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rates in the [ area have been determined and personnel have been made knowledgeable of them. (3) An individual cualified in radiation protection procedures with a radiation dose rate monitoring device. This individual is responsible for providing positive radiat5n protection control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified in the radiation protection procedures or the applicable j Radiation Work Permit. + The above procedure shall also apply to each high radiation area in which the intensity of radiation is greater than 1000 c. mrem /hr. In addition doors shall be locked or attended, to prevent unauthorized entry into these areas and the keys or key devices for locked dc - hall be maintained under the administrative control of the Plant Manager. 1. Health Physics personnel or personnel escorted by Health Physics personnel shall be exempt from the Radiation Work Permit issuance requirement during the performance of their assigned radiation protection duties, provided they comply with approved j radiation protection procedures for entry into high radiation areas. This footnote applies only to high radiation areas of 1000 mrem /hr j or less. 6.5 244a Amendment No. 44,39,78,104 1 ... ~ - - - - x- - w

E. Offsite Dose Calculation Manual (ODCM) The ODCM shall be approved by the Commissan prior to initial implementation. Changes to the ODCM shall satisfy the following requirements: 1. Shall by submitted to the Commission with the Semi-Annual Radioactive Effluent release report for the period in which the change (s) were made effective. This submittal shall contain: a. sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplementalinformation. Information submitted should consist of a package of those pages of the ODCM to be changed with each page numbered and provided with a revision date, together with appropriate analyses or evaluations justifying the change (s). 1 b. a determination that the change will not reduce the accuracy or reliability of dose calculations or setpoint determinations; and c. documentation of the fact that the change has been reviewed and found acceptable by the Operations Committee. 2. Shall become effective upon review and acceptance by the Operations Committee. F. Security Procedures shall be developed to implement the requirements of the Security Plan and the Security Contingency Plan. These implementing procedures, with the exception of those non-safety related procedures goveming work activities exclusively I applicable to or performed by security personnel, shall be reviewed by the Operations Committee and approved by a member of plant management designated by the Plant Manager. Security procedures not reviewed by the Operations Committee shall be reviewed and approved by the Superintendent, Security. l G. Temocrary Chances to Procedures Temporary changes to those procedures which are required to be reviewed by the Operations Committee described in A, B, C, D, E and F above, which do not change the intent of the originai precedures may be made with the concurrence of two members of the unit management staff, at least one of whom holds a Senior Operator License. Such changes should be documented, roviewed by the Operations Committee and approved by a member of plant management designated by the Plant Manager within one month. Temporary changes to health physics and security procedures not reviewed by the Operations Committee shall be reviewed by the General Superintendent, Radiation Services for health physics procedures and the Superintendent, Security for security procedures. 6.5 246b Amendment No. 45, 25,39, 68, 104 l

B. Records Retained for Plant Life (continued)

11. Records of the service lives of all safety-related snubbers, including the date at which the service life commences and

' associated installation and maintenance records. 1

i 6.6 247a Amendment No. 9, 104

o .w. i t B. Reportable Events 7 The following actions shall be taken for Reportable Events: a. The Commission shall be notified by a report submitted pursuant to the requirements of Section 50.73 to + 0 CFR Part 50

and, i

b. Each Reportable Event shall be reviewed by the Operations Committee and the results of this review shall be submitted to .' ;I the Safety Audit Committee and the President, NSP Nuclear Generation. l l t 'I l 6.7 250 Amendment No. 45,46, 104 I -.}}