ML093060414

Report of Changes to Technical Specifications Bases

ML093060414
Person / Time
Site:	Palisades
Issue date:	11/02/2009
From:	Patricia Anderson Entergy Nuclear Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML093060414 (71)

Text

Entergy Nuclear Operations, Inc.

Palisades Nuclear Plant 27780 Blue Star Memorial Highway Covert, MI 49043 Tel 269 764 2000 Paula K Anderson Licensing Manager November 2,2009 U. S .. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

SUBJECT:

Report of Changes to Technical Specifications Bases Palisades Nuclear Plant Docket 50-255 License No. DPR-20

Dear Sir or Madam:

This report is submitted in accordance with Palisades Technical Specification 5.5.12.d, which requires that changes to the Technical Specifications Bases, implemented without prior Nuclear Regulatory Commission (NRC) approval, be provided to the NRC on a frequency consistent with 10 CFR 50.71 (e). Attachment 1 provides a listing of all bases changes since issuance of the previous report, dated August 14, 2008, and identifies the affected sections and nature of the changes. Attachment 2 provides page change instructions and a copy of the current Technical Specifications Bases List of Effective Pages, Title Page, Table of Contents, and the revised Technical Specifications Bases sections listed in Attachment 1.

Summary of Commitments This letter identifies no new commitments and no revisions to existing commitments.

Sincerely, t)Z pka/jlk Attachment(s): 1. Technical Specifications Bases Change Chronology

2. Revised Technical Specifications Bases cc: Administrator, Region III, USNRC Project Manager, Palisades, USNRC Resident Inspector, Palisades, USNRC

1 TECHNICAL SPECIFICATIONS BASES CHANGE CHRONOLOGY DATE AFFECTED BASES CHANGES SECTION(S) 02/19/2009 B 3.7.15 and Bases revised to reflect License B3.7.16 Amendment 236 that modified the Spent Fuel Pool Region I storage requirements.

10129/2009 B 3.7.8 Bases revised to reflect the service water load outside containment is no longer the air compressors C-2A and C-2C, but instead is the after-coolers associated with the air compressors.

10/29/2009 B 3.3.1 Bases revised to reflect the turbine control system modification described in engineering change EC 5861.

10129/2009 B 3.4.12 Bases revised to reflect an editorial change to a Technical Specifications figure number.

Page 1 of 1

2 REVISED TECHNICAL SPECIFICATIONS BASES Page Change Instructions List of Effective Pages Title Page Table of Contents Bases Sections B 3.3.1 B 3.4.12 B 3.7.8 B 3.7.15 B 3.7.16

...

68 Pages Follow

TECHNICAL SPECIFICATIONS BASES CHANGES DOCKET 50-255 RENEWED FACILITY OPERATING LICENSE DPR-20 Page Change Instructions Revise your GOpy of the Palisades Technical Specifications Bases with the attached revised pages, ,The revised section pages are identified by an amendment number or the revision date at the bottom of the pages. Vertical lines in the margin indicate the. a reElS of chc:nge.

REMOVE INSERT List of Effective Pages List of Effective Pages Title Page Title Page T able of Contents Table of Contents Section B 3.3.1 Section B 3.3.1 Section B 3.4.12 Section B 3.4.12 Section B 3.7.8 Section B 3.7.8 Section B 3.7.15 Section B 3.7.15 Section B 3.7.16 Section B 3.7.16 Page 1 of 1

PALISADES TECHNICAL SPECIFICATIONS BASES 1 LIST OF EFFECTiVE PAGES COVERSHEET Title Page 236 - Revised 02/19/09 TABLE OF CONTENTS Pages i and ii Revised 02/19/09 TECHNICAL SPECIFICATIONS BASES Bases 2.0 Pages B 2.1.1 B 2.1.1-4 Revised 09/28/01 Pages B 2.1.2 B 2.1.2-4 189 Bases 3.0 Pages B 3.0 B 3.0-16 Revised 02/24/05 Bases 3.1 Pages B 3.1.1 B 3.1.1-5 189 Pages B3.1.2 B3.1.2-6 Revised 09/09/03 Pages B 3.1.3 B 3.1.3-4 189 Pages B 3.1.4 B 3.1.4-13 Revised 07/18/07 Pages B 3.1.5 B 3.1.5-7 Revised 07/02/04 Pages B 3.1.6 B 3.1.6-9 Revised 07/30103 Pages B 3.1.7 B 3.1.7-6 Revised 05/15/07 Bases 3.2 Pages B 3.2.1 B 3.2.1-11 Revised 08/06/04 Pages B 3.2.2 B 3.2.2-3 Revised 09/28/01 Pages B 3.2.3 B 3.2.3-3 Revised 09/28/01 Pages B 3.2.4 B 3.2.4-3 189 - Revised 08/09/00 Bases 3.3 Pages B 3.3.1 B 3.3.1-35 Revised 10/29/09 189 - Revised 02/12/01 I,

Pages B 3.3.2 B 3.3.2-10 Pages B 3.3.3 B 3.3.3-24 Revised 03/20108 Pages B 3.3.4 B 3.3.4-12 Revised 09109/03 Pages B 3.3.5 B 3.3.5-6 Revised 01/26/04 Pages B 3.3.6 B 3.3.6-6 189 - Revised 02/12/01 Pages B 3.3.7 B 3.3.7-12 Revised 04/19/05 Pages B 3.3.8 B 3.3.8-6 Revised 02/24/05 Pages B 3.3.9 B 3.3.9-5 189 - Revised 08/09/00 Pages B3.3.10-1-B3.3.10-4 189 Bases 3.4 Pages B 3.4.1 B 3.4.1-4 Revised 08/24/04 Pages B 3.4.2 B 3.4.2-2 189 f:Jages B 3.4.3 B 3J1..3-7 Revisod 01l27.'05 Pages B 3.4.4 B 3.4.4-4 Revised 09/21/06 Pages B 3.4.5 B 3.4.5-5 Revised 09/21/06 Pages B 3.4.6 B 3.4.6-6 Revised 07/31/07 Pages B 3.4.7 B 3.4.7-7 Revised 07/31/07 Pages B 3.4.8 B 3.4.8-5 Revised 07/31/07 Pages B 3.4.9 B 3.4.9-6 189 Pages 83.4.10-1-B3.4.10-4 189 Pages B3.4.11-1-B3.4.11-7 Revised 02/24/05 Pages B 3.4.12 B 3.4.12-13 Revised 10/29/09 Pages 83.4.13 B 3.4.13-7 Revised 03/20108 Pages B3.4.14-1-B3.4.14-8 189 - Revised 08/09/00 Pages 8 3.4.15 B 3.4.15-6 Revised 02/24/05 Pages B 3.4.16 B 3.4.16-5 Revised 02/24/05 Pages B 3.4.17 B 3.4.17-7 223 Revised 10/29/2009

PALISADES TECHI\JICAL SPECIFICATIONS BASES 2 LIST OF EFFECTIVE PAGES Bases 3.5 Pages B 3.5.1-1 - B 3.5.1-5 189 Page B 3.5.1-6 191 Page B 3.5.1-7 189 Page B 3.5.1-8 191 Pages B 3.5.2-1 - B 3.5.2-12 228 Pages B 3.5.3-1 - B 3.5.3-4 Revised 07/22/02 Pages B 3.5.4-1 - B 3.5.4-7 227 Pages B 3.5.5-1 - B 3.5.5-5 227 Bases 3.6 Pages B 3.6.1-1 - B 3.6.1-4 Revised 12/10102 Pages B 3.6.2-1 - B 3.6.2-8 Revised 08/12/03 Pages B 3.6.3-1 - B 3.6.3-12 Revised 03/02/04 Pages B 3.6.4-1 - B 3.6.4-3 Revised 04/27/01 Pages B 3.6.5-1 - B 3.6.5-3 Revised 09/09/03 Pages B 3.6.6-1 - B 3.6.6-12 227 Bases 3.7 Pages B 3.7.1 B 3.7.1-4 Revised 08/06/04 Pages B 3.7.2 B 3.7.2-6 Revised 12/02/02 Pages B 3.7.3 B 3.7.3-5 Revised 12102/02 Pages B 3.7.4 B 3.7.4-4 Revised 07/16/08 Pages B 3.7.5 B 3.7.5-9 Revised 02/24/05 Pages B 3.7.6 B 3.7.6-4 Revised 07/31/07 Pages B 3.7.7 B 3.7.7-9 Revised 06/07/05 Pages B 3.7.8 B 3.7.8-8 Revised 10/29/09 Pages B 3.7.9 B 3.7.9-3 Revised 07/16/01 Pages B 3.7.10 B 3.7.10-8 230 Pages B3.7.11-1-8'3.7.11-5 189 Pages B 3.7.12 B 3.7.12-7 Revised 07/16/03 Pages B3.7.13-1 ~ 'B 3.7:13-3 189 - Revised 08/09/00 Pages B 3.7.14 B 3.7.14-3 Revised 09/09/03 Pages B 3.7.15 B 3.7.15-2 236 Pages B 3.7.16 B 3.7.16-3 236 Pages B 3.7.17 B 3.7.17-3 Revised 07/22/02 Bases 3.8 Pages B 3.8.1 B 3.8.1-24 Revised 02/24/05 Pages B 3.8.2 B 3.8.2-4 Revised 11/06/01 Pages B 3.8.3 B 3.8.3-7 Revised 07/22/02 Pages B 3.8.4 B 3.8.4-9 Revised 07/13/06 Pages B' 3.8.5 B 3.8.5-3 Revised 11106/01 Pages B 3.8.6 B 3.8.6-6 189 - Revised 08/09/00 Pages B 3.8.7 B 3.8.7-3 189 Pages B 3.8.8 B 3.8.8-3 Revised 11/06/01 Pages B 3.8.9 B 3.8.9-7 Revised 11/06/01 Pages B 3.8.10 B 3.8.10-3 Revised 11106/01 Bases 3.9 Pages B 3.9.1-1 - B 3.9.1-4 189 - Revised 08/09/00 Pages B 3.9.2-1 - B 3.9.2-3 189 - Revised 02/12/01 Pages B 3.9.3-1 - B 3.9.3-6 189 - Revised 08/09/00 Pages B 3.9.4-1 - B 3.9.4-4 Revised 07/31/07 Pages' B 3:9.5-1 - B 3.9.5-4 Revised 07/31/07 Pages B 3.9.6-1 - B 3.9.6-3 189 - Revised 02/27/01 Revised 10/29/2009

PALISADES PLANT FACILITY OPERATING LICENSE DPR-20 APPENDIX A TECHNICAL SPECIFICATIONS

BASES As Amended Through Amendment No. 236 Revised 2/19/2009

B 2.0 SAFETY LIMITS (SLs)

B 2.1.1 Reactor Core SLs B 2.1.2 Primary Coolant System (PCS) Pressure SL B 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY B 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.1 SHUTDOWN MARGIN (SDM)

B 3.1.2 Reactivity Balance B 3.1.3 Moderator Temperature Coefficient (MTC)

B 3.1.4 Control Rod Alignment B 3.1.5 Shutdown and Part-Length Rod Group Insertion Limits B 3.1.6 Regulating Rod Group Position Limits B 3.1.7 Special Test Exceptions (STE)

B 3.2 POWER DISTRIBUTION LIMITS B 3.2.1 Linear Heat Rate (LHR)

B 3.2.2 TOTAL RADIAL PEAKING FACTOR (FRT)

B 3.2.3 QUADRANT POWER TILT (Tq)

B 3.2.4 AXIAL SHAPE INDEX (ASI)

B 3.3 INSTRUMENTATION B 3.3.1 Reactor Protective System (RPS) Instrumentation B 3.3.2 Reactor Protective System (RPS) Logic and Trip Initiation B 3.3.3 Engineered Safety Features (ESF) Instrumentation B 3.3.4 Engineered Safety Features (ESF) Logic and Manual Initiation B 3.3.5 Diesel Generator (DG) - Undervoltage Start (UV Start)

B 3.3.6 Refueling Containment High Radiation (CHR) Instrumentation B 3.3.7 . Post Accident Monitoring (PAM) Instrumentation B 3.3.8 Alternate Shutdown System B 3.3.9 Neutron Flux Monitoring Channels B 3.3.10 Engineered Safeguards Room Ventilation (ESRV) Instrumentation B 3.4 PRIMARY COOLANT SYSTEM (PCS)

B 3.4.1 PCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits B 3.4.2 PCS Minimum Temperature for Criticality B 3.4.3 PCS Pressure and Temperature (PIT) Limits 83.4.4 PCS Loops - fvl0DES 1 and 2 B 3.4.5 PCS Loops - MODE 3 B 3.4.6 PCS Loops - MODE 4 B 3.4.7 PCS Loops - MODE 5, Loops Filled B 3.4.8 PCS Loops - MODE 5, Loops Not Filled B 3.4.9 Pressurizer B 3.4.10 Pressurizer Safety Valves B 3.4.11 Pressurizer Power Operated Relief Valves (PORVs)

B 3.4.12 Low Temperature Overpressure Protection (LTOP) System B 3.4.13 PCS Operational LEAKAG E B 3.4.14 PCS Pressure Isolation Valve (PIV) Leakage B 3.4.15 PCS Leakage Detection Instrumentation B 3.4.16 PCS Specific Activity B 3.4.17 Steam Generator (SG) Tube Integrity Palisades Nuclear Plant Revised 2/19/2009

B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

B 3.5.1 Safety Injection Tanks (SITs)

B 3.5.2 ECCS - Operating B 3.5.3 ECCS - Shutdown B 3.5.4 Safety Injection Refueling Water Tank (SIRWT)

B 3.5.5 Containment Sump Buffering Agent and Weight Requirements B 3.6 CONTAINMENT SYSTEMS B 3.6.1 Containment B 3.6.2 Containment Air Locks B 3.6.3 Containment Isolation Valves B 3.6.4 Containment Pressure B 3.6.5 Containment Air Temperature B 3.6.6 Containment Cooling Systems B 3.7 PLANT SYSTEMS B 3.7.1 Main Steam Safety Valves (MSSVs)

B 3.7.2 Main Steam Isolation Valves (MSIVs)

B 3.7.3 Main Feedwater Regulating Valves (MFRVs) and MFRV Bypass Valves B 3.7.4 Atmospheric Dump Valves (ADVs)

B 3.7.5 Auxiliary Feedwater (AFW) System B 3.7.6 Condensate Storage and Supply B 3.7.7 Component Cooling Water (CCW) System B 3.7.8 Service Water System (SWS)

B 3.7.9 Ultimate Heat Sink (UHS)

B3.7.10 Control Room Ventilation (CRV) Filtration B3.7.11 Control Room Ventilation (CRV) Cooling B3.7.12 Fuel Handling Area Ventilation System B 3.7.13 Engineered Safeguards Room Ventilation (ESRV) Dampers

'B 3.7.14 Spent Fuel Pool (SFP) Water L e v e l "

B3.7.15 Spent Fuel Pool (SFP) Boron Concentration B3.7.16 Spent Fuel Pool Storage B 3.7.17 Secondary Specific Activity B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.1 AC Sources - Operating B 3.8.2 AC Sources - Shutdown B 3.8.3 Diesel Fuel, Lube Oil, and Starting Air B 3.ii.4 DC Sources - Operating B 3.8.5 DC Sources - Shutdown B 3.8.6 Battery Cell Parameters B 3.8.7 Inverters - Operating B 3.8.8 Inverters - Shutdown B 3.8.9 Distribution Systems - Operating B 3.8.10 Distribution Systems - Shutdown B 3.9 REFUELING OPERATIONS B 3.9.1 Boron Concentration B 3.9.2 Nuclear Instrumentation B 3.9.3 Containment Penetrations B 3.9.4 Shutdown Cooling (SDC) and Coolant Circulation - High Water Level B 3.9.5 Shutdown Cooling (SDC) and Coolant Circulation - Low Water Level B 3.9.6 Refueling Cavity Water Level Palisades Nuclear Plant ii Revised 2/19/2009

RPS Instrumentation B 3.3.1 B 3.3 INSTRUMENTATION B 3.3.1 Reactor Protective System (RPS) Instrumentation BASES BACKGROUND The RPS initiates a reactor trip to protect against violating the acceptable fuel design limits and breaching the reactor coolant pressure boundary during Anticipated Operational Occurrences (AOOs). (As defined in 10 CFR 50, Appendix A, "Anticipated operational occurances mean those conditions of normal operation which are expected to occur one or more times during the life of the nuclear power unit and include but are not limited to loss of power to all recirculation pumps, tripping of the turbine generator set, isolation of the main condenser, and loss of all offsite power.") By tripping the reactor, the RPS also assists the Engineered Safety Features (ESF) systems in mitigating accidents.

The protection and monitoring systems have been designed to ensure safe operation of the reactor. This is achieved by specifying Limiting Safety System Settings (LSSS) in terms of parameters directly monitored by the RPS, as well as LCOs on other reactor system parameters and equipment performance.

The LSSS, defined in this Specification as the Allowable Values; in conjunction with the LCOs, establish the threshold for protective system action to prevent exceeding acceptable limits during Design Basis Accidents (DBAs).

During AOOs, which are those events expected to occur one or more times during the plant life, the acceptable limits are:

• The Departure from Nucleate Boiling Ratio (DNBR) shall be maintained above the Safety Limit (SL) value to pre,-:entdeparture from Ilucleate boiling;

• Fuel centerline melting shall not occur; and

• The Primary Coolant System (PCS) pressure SL of 2750 psia shall not be exceeded.

Maintaining the parameters within the above values ensures that the offsite dose will be within the 10 CFR 50 (Ref. 1) and 10 CFR 100 (Ref. 2) criteria during AOOs.

Palisades Nuclear Plant B 3.3.1-1 Revised 10/29/2009

RPS Instrumentation B 3.3.1 BASES BACKGROUND Accidents are events that are analyzed even though they are not (continued) expected to occur during the plant life. The acceptable limit during accidents is that the offsite dose shall be maintained within an acceptable fraction of 10 CFR 100 (Ref. 2) limits. Different accident categories allow a different fraction of these limits based on probability of occurrence. Meeting the acceptable dose limit for an accident category is considered having acceptable consequences for that event.

The RPS is segmented into four interconnected !l1.odules. These mbdules are:

• Measurement channels;

• RPS trip units;

• Matrix Logic; and

• Trip Initiation Logic.

This LCO addresses measurement channels and RPS trip units. It also addresses the automatic bypass removal feature for those trips with Zero Power Mode bypasses. The RPS LogiQ and. Trip Initiation Logic are addressed in LCO 3.3.2, "Reactor Protective System (RPS) Logic and- Trip Initiation." The role of the- measurement channels, RPS trip units, and RPS Bypasses is discussed below. .-

Measurement Channels Measurement channels, consisting of pressure switches, field transmitters, or process sensors and associated instrumentation, provide a measurable electronic signal based upon the physical characteristics of the parameter being measured.

With the exception of Hi Startup Rate, which employs two instrument channels, and Loss of Load, which employs a single pressure sensor, four identical measurement channels with electrical andphysical separation are provided for each parameter used in the direct generation of trip signals. These are designated channels A through D.

Some measurement channels provide input to more than one RPS trip unit within the same RPS channel. In addition, some measurement channels may also be used as inputs to Engineered Safety Features (ESF) bistables, and most provide indication in the control room.

Palisades Nuclear Plant B 3.3.1-2 Revised 10/29/2009

RPS Instrumenta:tion B 3.3:1 BASES BACKGROUND Measurement Channels (continued)

(continued)

In the case of Hi Startup Rate and Loss of Load, where fewer than four sensor channels are employed, the reactor trips provided are not relied upon by the plant safety analyses. The sensor channels do however, provide trip input signals to all four RPS channels.

When a channel monitoring a parameter exceeds a predetermined setpoint, indicating an abnormal condition, the bistable monitoring the parameter in that channel will trip. Tripping two or more channels of bistable trip units monitoring the same parameter de-energizes Matrix Logic, (addressed by LCO 3.3.2) which in turn de-energizes the Trip Initiation Logic. This causes all four DC clutch power supplies to de-energize, interrupting power to the control rod drive mechanism clutches, allowing the full length control rods to insert into the core.

For those trips relied upon in the safety analyses, three of the four measurement and trip unit channels can meet the redundancy and testability of GDC 21 in 10 CFR 50, Appendix A (Ref. 1). This LCO requires, however, that four channels be OPERABLE. The fourth channel provides additional flexibility by allowing one channel to be removed *from service (trip channel bypassed) for maintenance OJ testing while still maintaining a minimum two-out-of-three logic.

Since no single failure will prevent a protective system actuation, this arrangement meets the requirements of IEEE Standard 279-1971 (Ref. 3).

Most of the RPS trips are generated by comparing a single measurement to a fixed bistable setpoint. Two trip Functions, Variable High Power Trip and Thermal Margin Low Pressure Trip, make use of more than one measurement to provide a trip.

The required RPS Trip Functions utilize the following input instrumentation:

• Variable High Power Trip (VHPT)

The VHPT uses Q Power as its input. Q Power is the higher of NI power from the power range NI drawer and primary calorimetric power (11 T power) based on PCS hot leg and cold leg temperatures. The measurement channels associated with the VHPT are the power range excore channels, and the PCS hot and cold leg temperature channels.

Palisades Nuclear Plant B 3.3.1-3 Revised 10/29/2009

RPS Instrumentation B 3.3.1 BASES BACKGROUND Measurement Channels (continued)

II Variable High Power Trip (VHPT) (continued)

The Thermal Margin Monitors provide the complex signal processing necessary to calculate the TM/LP trip setpoint, VHPT trip setpoint and trip comparison, and Q Power calculation. On power decreases the VHPT setpoint tracks power levels downward so that it is always within a fixed increment above current power, subject to a minimum value.

On power increases, the trip setpoint remains fixed unless manually reset, at which point it increases to the new setpoint, a fixed increment above Q Power at the time of reset, subject to a maximum value. Thus, during power escalation, the trip setpoint must be repeatedly reset to avoid a reactor trip.

II High Startup Rate Trip The High Startup Rate trip uses the wide range Nuclear Instruments (f:.Jls) to provide an input signal. There are only two wide range NI channels. The wide range channel signal processing electronics are physically mounted in RPS cabinet channels C (NI-1/3) and D (NI-2/4). Separate bistable trip units mounted within the NI-1/3 wide range channel drawer supply High Startup Rate trip signals to RPS channels A and C. Separate bistable trip units mounted within the NI-2/4 wide range channel drawer provide High Startup Rate trip signals to RPS channels B and D.

II Low Primary Coolant Flow Trip The Low Primary Coolant Flow Trip utilizes 16 flow measurement channels which monitor the differential pressure across the primary side of the steam generators. Each RPS channel, A, B, C, and D, receives a signal which is the sum of four differential pressure signals. This totalized signal is compared with a setpoint in the RPS Low Flow bistable trip unit for that RPS channel.

Palisades Nuclear Plant B 3.3.1-4 Revised 10/29/2009

RPS Instrumentation B 3.3.1 BASES BACKGROUND Measurement Channels (continued)

(continued)

G Low Steam Generator Level Trips There are two separate Low Steam Generator Level trips, one for each steam generator. Each Low Steam Generator Level trip monitors four level measurement channels for the associated steam generator, one for each RPS channel. '$",

G Low Steam Generator Pressure Trips There are also two separate Low Steam Generator Pressure trips, one for each steam generator. Each Low Steam Generator Pressure trip monitors four pressure measurement channels for the associated steam generator, one for each RPS channel.

G High Pressurizer Pressure Trip The High Pressurizer Pressure Trip monitors four pressurizer pressure channels, one for each RPS channel.

G Thermal Margin Low Pressure (TM/LP) Trip The TM/LP Trip utilizes bistable trip units. Each of these bistable trip units receives a calculated trip setpoint from the Thermal Margin Monitor (TMM) and compares it to the measured pressurizer pressure signal. The TM/LP setpoint is based on Q power (the higher of NI power from the power range NI drawer, or LH power, based on PCS hot leg and cold leg temperatures) pressurizer pressure, PCS cold leg temperature, and Axial Shape Index. The TMM provide the complex signal processing -,

necessary to calculate the TM/LP trip setpoint, TM/LP trip comparison signal, and Q Power. " , -. ..,

Palisades Nuclear Plant B 3.3.1-5 Revised 10/29/2009

RPS Instrumentation B 3.3.1 BASES BACKGROUND Measurement Channels (continued)

(continued) e Loss of Load Trip The Loss of Load Trip is initiated by two-out-of-three logic from pressure switches in the turbine auto stop oil circuit that sense a turbine tripJor input to all four RPS auxiliary trip units. The Loss of Load Trip isactoated by turbine auxiliary relays 305L and 305R.

Relay 305L proVides'input to RPS channels A and C; 305R to channels Band D: Relays 305L and 305R are energized on a turbine trip. Their inputs are the same as the inputs to the turbine solenoid trip valve, 20ET.

If a turbine trip is generated by loss of auto stop oil pressure, the auto stop oil pressure switches, by two-out-of-three logic, will actuate relays 305L and 305R and generate a reactor trip. If a turbine trip is generated by an input to the solenoid trip valve, relays 305L and 305R, which are wired in parallel, will also be actuated and will generate a reactor trip.

e Containment High Pressure Trip The Containment High Pressure Trip is actuated by four pressure switches; one for each RPS channel. '

e Zero Power Mode Bypass Automatic Removal The Zero Power Bypass allows manually bypassing (i.e., disabling) four reactor trip functions, Low PCS Flow, Low SG A Pressure, Low SG B Pressure, and TMILP (low PCS pressure),

when reactor power (as indicated by the wide range nuclear instrument channels) is below 10-4 %. This bypassing is necessary to allow PIPS te,,:;?tingand control rod drive mecllanism testing when the reactor is' shutdown and plant conditions would cause a reactor trip to be f)(e~ent.

The Zero Power Mode Bypass removal interlock uses the wide range nuclear instruments (Nls) as measurement channels.

There are only twowlde range NI channels. Separate bistables are provided to actuate the bypass removal for each RPS channel. Bistables in the NI-1/3 channel provide the bypass removal function for RPS channels A and C; bistables in the NI-2/4 channel for RPS channels Band D.

Palisades Nuclear Plant B 3.3.1-6 Revised 10/29/2009

RPS Instrumentation B 3.3.1 BASES BACKGROUND Several measurement instrument channels provide more than one (continued) required function. Those sensors shared for RPS and ESF functions are identified in Table B 3.3.1-1. That table provides a listing orthose shared channels and the Specifications which they affect.

RPS Trip Units Two types of RPS trip units are used in the RPS cabinets; bistable trip units and auxiliary trip units:

A bistable trip unit receives a measured process Signal from its instrument channel and compares it to a setpoint; the trip unit actuates three relays, with contacts in the Matrix Logic channels, when the measured signal is less conservative than the setpoint.

They also provide local trip indication and remote annunciation.

An auxiliary trip unit receives a digital input (contacts open or closed); the trip unit actuates three relays, with contacts in the Matrix Logic channels, when the digital input is received. They also provide local trip indication and remote annunciation ..

Each RPS channel has four auxiliary trip units and seven bistable trip units.

The contacts from these trip unit relays are arranged into six coincidence matrices, comprising the Matrix Logic. If bistable trip units monitoring the same parameter in at least two channels trip, the Matrix Logic will generate a reactor trip (two-out-of-four logic).

Four of the RPS measurement channels provide contact outputs to the RPS, so the comparison of an analog input to a trip setpoint is not necessary. In these cases, the bistable trip unit is replaced with an auxiliary trip unit. The Quxiliary trip units provide contact muitlplicQtion so the single input contact opening can provide multiple contact outputs to the coincidence logic as well as trip indication and annunciation.

Palisades Nuclear Plant B 3.3.1-7 Revised 10/29/2009

RPS Instrumentation B 3.3.1 BASES BACKGROUND RPS Trip Units (continued)

(continued)

Trips employing auxiliary trip units include the VHPT, which receives contact inputs from the Thermal Margin Monitors; the High Startup Rate trip which employs contact inputs from bistables mounted in the two wide range drawers; the Loss of Load Trip which receives contact inputs from one of two auxiliary relays which are operated by two-out-of-three logic switches sensing turbine auto stop oil pressure; and the Containment High Pressure (CHP) trip, which employs containment pressure switch contacts.

There are four RPS trip units, designated as channels A through D, each channel having eleven trip units, one for each RPS Function. Trip unit output relays de-energize when a trip occurs.

All RPS Trip Functions, with the exception of the Loss of Load and CHP trips, generate a pretrip alarm as the trip setpoint is approached.

The Allowable Values are specified for each safety related RPS trip Function which is credited in the safety analysis. Nominal trip setpoints are specified in the plant procedures. The nominal setpoints are selected to ensure plant parameters do not exceed the Allowable Value if the instrument loop is performing as required. The methodology used to determine the nominal trip setpoints is also provided in plant documents. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable. Each Allowable Value specified is more conservative than the analytical limit determined in the safety analysis in order to account for uncertainties appropriate to the trip Function. These uncertainties are addressed as described in plant documents. A channel is inoperable if its actual setpoint is not within its Allowable Value.

Setpoints in accordance with the Allowable Value will ensure that SLs of Chapter 2.0 are not violated during AOOs and the consequences of DBAs will be acceptable, providing the plant is operated from within the LCOs at the onset of the AOO or DBA and the equipment functions as designed.

Note that in the accompanying LCO 3.3.1, the Allowable Values of Table 3.3.1-1 are the LSSS.

Palisades Nuclear Plant B 3.3.1-8 Revised 10/29/2009

RPS Instrumentation B 3.3.1 BASES BACKGROUND Reactor Protective System Bypasses (continued)

Three different types of trip bypass are utilized in the RPS, Operating Bypass, Zero Power Mode Bypass, and Trip Channel Bypass. The Operating Bypass or Zero Power Mode Bypass prevent the actuation of a trip unit or auxiliary trip unit; the Trip Channel Bypass prevents the trip unit output from affecting the Logic Matrix. A channel which is bypassed, other than as allowed by the Table 3.3.1-1 footnotes, cannot perform its specified safety function and must be considered to be inoperable.

Operating Bypasses The Operating Bypasses are initiated and removed automatically during startup and shutdown as power level changes. An Operating Bypass prevents the associated RPS auxiliary trip unit from receiving a trip signal from the associated measurement channel. With the bypass in place, neither the pre-trip alarm nor the trip will actuate if the measured parameter exceeds the set pOint. An annunciator is provided for each Operating Bypass. The RPS trips with Operating Bypasses are:

a. High Startup Rate Trip bypass. The High Startup Rate trip is automatically bypassed when the associated wide range channel indicates below 1 E-4%.RTP, and when the a$~ociated power range excore channel indicates above 13% RTP. These bypasses are automatically removed between 1 E-4% RTP and 13% RTP.

b. Loss of Load bypass. The Loss of Load trip is automatically bypassed when the associated power range excore channel indicates below 17% RTP. The bypass is automatically removed when the channel indicates above the set point. The same power range excore cil8.nnel bistable is used to bypass tl"le Higrl Startup Rate trip and the Loss of Load trip for that RPS channel: "

Palisades Nuclear Plant B 3.3.1-9 Revised 10/29/2009

RPS Instrumentation B 3.3.1 BASES BACKGROUND Operating Bypasses (continued)

(continued)

Each wide range channel contains two bistables set at 1 E-4% RIP, one bistable unit for each associated RPS channel. Each of the two wide range channels affect the Operating Bypasses for two RPS channels; wide range channel NI-1/3 for RPS channels A and C, wide range channel NI-2/4 for RPS channels Band D. Each of the four power range excore channel affects the Operating Bypasses for the associated RPS channel. The power range excore channel bistables associated with the Operating Bypasses are set at a nominal 15%, and are required to actuate between 13% RTP and 17% RTP.

Zero Power Mode (ZPM) Bypass The ZPM Bypass is used when the plant is shut down and it is desired to raise the control rods for control rod drop testing with PCS flow, pressure or temperature too low for the RPS trips to be reset. ZPM bypasses may be manually initiated and removed when wide range power is below 1 E-4% RTP, and are automatically removed if the associated wide range NI indicated power exceeds 1E-4% RTP. A ZPM bypass prevents the RPS trip unit from actuating if the measured parameter exceeds the set point. Operation of the pretrip alarm is unaffected by the zero power mode bypass. An annunciator indicates the presence of any ZPM bypass. The RPS trips with ZPM bypasses are:

a. Low Primary Coolant System Flow.

b. Low Steam Generator Pressure.

c. Thermal Margin/Low Pressure.

Tile ,:vide range NI cllannels provide contact closure permissive signals when indicated power is below 1~-4% RTP. The ZPM bypasses may then be manually initiated or removed by actuation of key-lock switches.

One key-lock switch located on each RPS cabinet controls the ZPM Bypass for the associated RPS trip channels. The bypass is automatically removed if the associated wide range NI indicated power exceeds 1 E-4% RTP. The same wide range NI channel bistables that provide the ZPM Bypass permissive and removal Signals also provide the high startup rate trip Operating Bypass actuation and removal.

Palisades Nuclear Plant B 3.3.1-10 Revised 10/29/2009

RPS Instrumentation B 3.3.1 BASES BACKGROUND Trip Channel Bypass (continued)

A Trip Channel Bypass is used when it is desired to physically remove an individual trip unit from the system, or when calibration or servicing of a trip channel could cause an inadvertent trip. A trip Channel Bypass may be manually initiated or removed at any time by actuation of a key-lock switch. A Trip Channel Bypass prevents the trip unit output from affecting the RPS logic matrix. A light above the bypass switch indicates that the trip channelllas been bypassed. Each RPS trip unit has an associated trip channel bypass:

The key-lock trip channel bypass switch is located above each trip unit.

The key cannot be removed when in the bypass position. Only one key for each trip parameter is provided, therefore the operator can bypass only one channel of a given parameter at a time. During the bypass condition, system logic changes from two-out-of-four to two-out-of-three channels required for trip.

APPLICABLE Each of the analyzed accidents and transients can be detected by one SAFETY ANALYSES or more RPS Functions. The accident analysis contained in Reference 4 takes credit for most RPS trip Functions. The High Startup Rate and Loss of Load Functions, which are not specifically credited in the accident analysis-are part of the NRC approved licensing basis for the plant. The High Startup Rate and Loss of Load trips are purely equipment protective, and their use minimizes the potential for equipment damage.

The specific safety analyses applicable to each protective Function are identified below.

1. Variable High Power Trip (VHPT)

The VHPT provides reactor core protection against positive reactivity excursions.

The safety analysis assumes that this trip is OPERABLE to terminate excessive positive reactivity insertions during power operation and while shut down.

Palisades Nuclear Plant B 3.3.1-11 Revised 10/29/2009

RPS Instrumentation B 3.3.1 BASES APPLICABLE 2. High Startup Rate Trip SAFETY ANALYSIS (continued) There are no safety analyses which take credit for functioning of the High Startup Rate Trip. The High Startup Rate trip is used to trip the reactor when excore wide range power indicates an excessive rate of change. The High Startup Rate trip minimizes transients for events such as a continuous control rod withdrawal or a boron dilution event from low power levels. The trip may be operationally bypassed when THERMAL POWER is

< 1E-4% RTP, when poor counting statistics may lead to erroneous indication. It may also be operationally bypassed at

> 13% RTP, where moderator temperature coefficient and fuel temperature coefficient make high rate of change of power unlikely.

There are only two wide range drawers, with each supplying contact input to auxiliary trip units in two RPS channels.

3. Low Primary Coolant System Flow Trip The Low PCS Flow trip provides DNB protection during events which suddenly reduce the PCS flow rate during power operation, a

such as loss of power to, or seizure of, primary coolant pump.

.. -. .

Flow in each of the four PCS loops is determined from pressure drop from inlet to outlet of the SGs. The total PCS flow is determined, for the RPS flow channels, by summing the loop pressure drops across the SGs and correlating this pressure sum with the sum of SG differential pressures which exist at 100% flow (four pump operation at full power Tave). Full PCS flow is that flow which exists at RTP, at full power T ave , with four pumps operating.

4, 5. LO'.AJ Steam Generator [ evel Trip The Low Steam Generator Level trips are provided to trip the reactor in the event of excessive steam demand (to prevent overcooling the PCS) and loss of feedwater events (to prevent overpressurization of the PCS).

The Allowable Value assures that there will be sufficient water inventory in the SG at the time of trip to allow a safe and orderly plant shutdown and to prevent SG dryout assuming minimum AFW capacity.

Palisades Nuclear Plant B 3.3.1-12 Revised 10/29/2009

RPS Instrumentation B 3.3.1 BASES APPLICABLE 4,5. Low Steam Generator Level Trip (continued)

SAFETY ANALYSIS (continued) Each SG level is sensed by measuring the differential pressure in the upper portion of the downcomer annulus in the SG. These trips share four level sensing channels on each SG with the AFW actuation signal.

6,7. Low Steam Generator Pressure Trip The Low Steam Generator Pressure trip provides protection against an excessive rate of heat extraction from the steam generators, which would result in a rapid uncontrolled cooldown of the PCS. This trip provides a mitigation function in the event of an MSLB.

The Low SG Pressure channels are shared with the Low SG Pressure signals which isolate the steam and feedwater lines.

8. High Pressurizer Pressure Trip The High Pressurizer Pressure trip, in conjunction with pressurizer

. safety valves and Main Steam Safety Valves (MSSVs), provides protection against overpressure conditions in the PCS when at operating temperature. The safety aflalyses assume the High Pressurizer Pressure trip is OPERABLE during accidents and transients which suddenly reduce PCS cooling (e.g., Loss of Load, Main Steam Isolation Valve (MSIV) closure, etc.) or which suddenly increase reactor power (e.g., rod ejection accident).

The High Pressurizer Pressure trip shares four safety grade instrument channels with the TM/LP trip, Anticipated Transient Without Scram (A TWS) and PORV circuits, and the Pressurizer Low Pressure Safety Injection Signal.

Palisades Nuclear Plant B 3.3.1-13 Revised 10/29/2009

RPS Instrumentation B 3.3.1 BASES APPLICABLE 9. Thermal Margin/Low Pressure (TM/LP) Trip SAFETY ANALYSIS (continued) The TM/LP trip is provided to prevent reactor operation when the DNBR is insufficient. The TM/LP trip protects against slow reactivity or temperature increases, and against pressure decreases.

The trip is initiated whenever the PCS pressure signal drops

~,- below a minimum value (Pmin) or a computed value (Pvar) as described below, whichever is higher.

The TM/LP trip uses Q Power, ASI, pressurizer pressure, and cold leg temperature (Tc) as inputs.

Q Power is the higher of core THERMAL POWER (1:1 T Power) or nuclear power. The 1:1T power uses hot leg and cold leg RTDs as inputs. Nuclear power uses the power range excore channels as inputs. Both the 1:1 T and excore power signals have provisions for calibration by calorimetric calculations.

The ASI is calculated from the upper and lower power range excore detector signals, as explained in SeCtion 1.1, "Definitions."

The signal is corrected for the difference between the flux at the core periphery and the flux at the detectors. -.

The Tc value is the higher of the two cold leg signals.

The Low Pressurizer Pressure trip limit (Pvar)is calculated using the equations given in Table 3.3.1-2.

The calculated limit (Pvar) is then compared to a fixed Low Pressurizer Pressure trip limit (Pmin). The auctioneered highest of tllese signals becolTles tile trip limit (PlliP)' Phil-' is compared to tIle measured PCS pressure and a trip signal is generated when the measured pressure for that channel is less than or equal to Ptrip . A pre-trip alarm is also generated when P is less than or equal to the pre-trip setting, Ptrip + I:1P.

The TM/LP trip setpoint is a complex function of these inputs and represents a minimum acceptable PCS pressure for the existing temperature and power conditions. It is compared to actual PCS pressure in the TM/LP trip unit.

Palisades Nuclear Plant B 3.3.1-14 Revised 10/29/2009

RPS Instrumentation B 3.3.1 BASES APPLICABLE 10. Loss of Load Trip SAFETY ANALYSIS (continued) There are no safety analyses which take credit for functioning of the Loss of Load Trip.

The Loss of Load trip is provided to prevent lifting the pressurizer and main steam safety valves in the event of a turbine generator trip while at power. The trip is equipment protective. The safety analyses do not assume that this trip functions during any -

accident or transient. The Loss of Load trip uses two-out-of-three logic from pressure switches in the turbine auto stop oil circuit to sense a turbine trip for input to all four RPS auxiliary trip units.

11. Containment High Pressure Trip The Containment High Pressure trip provides a reactor trip in the event of a Loss of Coolant Accident (LOCA) or Main Steam Line Break (MSLB). The Containment High Pressure trip shares sensors with the Containment High Pressure sensing logic for Safety Injection, Containment Isolation, and Containment Spray.

Each of these sensors has a single bellows which actuates two microswitches. One microswitch on each of four sensors provides an input to the RPS .

.:&;...-"

12. Zero Power Mode Bypass Removal The only RPS bypass considered in the safety analyses is the Zero Power Mode (ZPM) Bypass. The ZPM Bypass is used when the plant is shut down and it is desired to raise the control rods for control rod drop testing with PCS flow or temperature too low for the RPS Low PCS Flow, Low SG Pressure, or Thermal Margin/Low Pressure trips to be reset. ZPM bypasses are automatically rernoved if tile wide range NI indicated power J" .'

exceeds 1 E-4% RTP.

Palisades Nuclear Plant B 3.3.1-15 Revised 10/29/2009

RPS Instrumentation B 3.3.1 BASES APPLICABLE 12. Zero Power Mode Bypass Removal (continued)

SAFETY ANALYSIS (continued) The safety analyses take credit for automatic removal of the ZPM Bypass if reactor criticality due to a Continuous Control Rod Bank Withdrawal should occur with the affected trips bypassed and PCS flow, pressure, or temperature below the values at which the RPS could be reset. The ZPM Bypass would effectively be removed when the first wide range NI channel indication reached

-'

1E-4% RTP. With the ZPM Bypass for two RPS channels removed, the RPS would trip on one of the un-bypassed trips.

This would prevent the reactor reaching an excessive power level.

If a reactor criticality due to a Continuous Control Rod Bank Withdrawal should occur when PCS flow, steam generator pressure, and PCS pressure (TM/LP) were above their trip setpoints, a trip would terminate the event when power increased to the minimum setting (nominally 30%) of the Variable High Power Trip. In this case, the monitored parameters are at or near their normal operational values, and a trip initiated at 30% RTP provides adequate protection.

The RPS design also includes automatic removal of the Operating Bypasses for the High Startup Rate and Loss of Load trips. The safety analyses do not assume fllnctioning of either these trips or the automatic removal of their bypasses.

The RPS instrumentation satisfies Criterion 3 of 10 CFR 50.36(c)(2).

LCO The LCO requires all instrumentation performing an RPS Function to be OPERABLE. Failure of the trip unit (including its output relays), any required portion of the associated instrument channel, or both, renders the affected channel(s) inopernble and reduces the reliability of the affected Functions. Failure of an automatic ZPM bypass removal channel may also impact the associated instrument channel(s) and reduce the reliability of the affected Functions.

Palisades Nuclear Plant B 3.3.1-16 Revised 10/29/2009

RPS Instrumentation B 3.3.1 BASES LCO Actions allow Trip Channel Bypass of individual channels, but the (continued) bypassed channel must be considered to be inoperable. The bypass key used to bypass a single channel cannot be simultaneously used to bypass that same parameter in other channels. This interlock prevents operation with more than one channel of the same Function trip channel bypassed. The plant is normally restricted to 7 days in a trip channel bypass, or otherwise inoperable condition before either restoring the Function to four channel operation (two-out-of-four logic) or placing the channel in trip (one-out-of-three logle). -

The Allowable Values are specified for each safety related RPS trip Function which is credited in the safety analysis. Nominal trip setpoints are specified in the plant procedures. The nominal setpoints are selected to ensure plant parameters do not exceed the Allowable Value if the instrument loop is performing as required. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, is acceptable. Each Allowable Value specified is more conservative than the analytical limit determined in the safety analysis in order to account for uncertainties appropriate to the trip Function.

These uncertainties are addressed as described in plant documents.

Neither Allowable Values nor setpoints are specified for the non-safety related. RPS Trip Functions, since no safety analysis assumptions would be violated if they are not set at a particular value.

The following Bases for each trip Function identify the above RPS trip Function criteria items that are applicable to establish the trip Function OPERABILITY.

1. Variable High Power Trip (VHPT)

This LCO requires all four channels of the VHPT Function to be OPERABLE.

The Allowable Value is high enough to provide an operating envelope that prevents unnecessary V,HPT trips during normal plant operations. The Allowable Value is low enough for the system to function adequately during reactivity addition events.

Palisades Nuclear Plant B 3.3.1-17 Revised 10/29/2009

RPS Instrumentation B 3.3.1 BASES LCO 1. Variable High Power Trip (VHPT) (continued)

(continued)

The VHPT is designed to limit maximum reactor power to its maximum design and to terminate power excursions initiating at lower powers without power reaching this full power limit. During plant startup, the VHPT trip setpoint is initially at its minimum value, :5 30%. Below 30% RTP, the VHPT setpoint is not required to "track" with Q Power, i.e., be adjusted to within 15% RTP. It remains fixed until manually reset, at which point it increases to

5 15% above existing Q Power.

The maximum allowable setting of the VHPT is 109.4% RTP.

Adding to this the possible variation in trip setpoint due to calibration and instrument error, the maximum actual steady state power at which a trip would be actuated is 113.4%, which is the value assumed in the safety analysis.

2. High Startup Rate Trip This LCO requires four channels of High Startup Rate Trip Function to be OPERABLE in MODES 1 and 2 .

. The High Startup Rate trip serves as a backup to the administratively enforced startup rate limit. The FUDction is not credited in the accident analyses; therefore, no Allowable Value for the trip or operating bypass Functions is derived from analytical limits and none is specified.

The four channels of the High Startup Rate trip are derived from two wide range NI Signal processing drawers. Thus, a failure in one wide range channel could render two RPS channels inoperable. It is acceptable to continue operation in this condition because the Higll Startup Rate trip is not credited in Clny safety analyses.

The requirement for this trip Function is modified by a footnote, which allows the High Startup Rate trip to be bypassed when the wide range NI indicates below 10E-4% or when THERMAL POWER is above 13% RTP. If a High Startup Rate trip is bypassed when power is between these limits, it must be considered to be inoperable.

Palisades Nuclear Plant B 3.3.1-18 Revised 10/29/2009

RPS Instrumentation B 3.3.1 BASES LCO 3. Low Primary Coolant System Flow Trip (continued)

This LCO requires four channels of Low PCS Flow Trip Function to be OPERABLE.

This trip is set high enough to maintain fuel integrity during a loss of flow condition. The setting is low enough t6~allow for normal operating fluctuations from.offsite power.

The Low PCS Flow trip setpoint of 95% of full PCS flow insures that the reactor cannot operate when the flow rate is less than 93% of the nominal value considering instrument errors. Full PCS flow is that flow which exists at RTP, at full power Tave, with four pumps operating.

The requirement for this trip Function is modified by a footnote, which allows use of the ZPM bypass when wide range power is below 1E-4% RTP. That bypass is automatically removed when the associated wide range channel indicates 1E-4% RTP. If a trip channel is bypassed when power is above 1E-4% RTP, it must be considered to be inoperable.

4, 5. Low Steam Generator Level Trip This LCO requires four channels of Low Steam Generator Level Trip Function per steam generator to be OPERABLE.

The 25.9% Allowable Value assures that there is an adequate water inventory in the steam generators when the reactor is critical and is based upon narrow range instrumentation. The 25.9%

indicated level corresponds to the location of the feed ring.

6,7. Lo'vv Stearn Generator Pressure Trip This LCO requires four channels of Low Steam Generator Pressure Trip Function per steam generator to be OPERABLE.

The Allowable Value of 500 psia is sufficiently below the full load operating value for steam pressure so as not to interfere with normal plant operation, but still high enough-to provide the required protection in the event of excessive steam demand.

Since excessive steam demand causes the PCS to cool down, resulting in positive reactivity addition to the core, a reactor trip is required to offset that effect.

Palisades Nuclear Plant B 3.3.1-19 Revised 10/29/2009

RPS Instrumentation B 3.3.1 BASES Leo (continued) 8. High Pressurizer Pressure Trip This LeO requires four channels of High Pressurizer Pressure Trip Function to be OPERABLE.

The Allowable Value is set high enough to allow for pressure increases in the pes during normal operation (i.e., plant transients) not indicative of an abnormal condition. The setting is below the lift setpoint of the pressurizer safety valves and low enough to initiate a reactor trip when an abnormal condition is indicated.

9. Thermal Margin/Low Pressure (TM/LP) Trip This LeO requires four channels of TM/LP Trip Function to be OPERABLE.

The TM/LP trip setpoints are derived from the core thermal limits through application of appropriate allowances for measurement uncertainties and processing errors. The allowances specifically account for instrument drift in both power and inlet temperatures, calorimetric power measurement, inlet temperature measurement, and primary system pressure measurement.

Other uncertainties including allowances for assembly power tilt, fuel pellet manufacturing tolerances, core flow measurement uncertainty and core bypass flow, inlet temperature measurement time delays, and ASI measurement, are included in the development of the TM/LP trip setpoint used in the accident analysis.

Tile requirement for tliis trip Function is modified by a footnote, which allows use of the ZPM bypass when wide range power is below 1E-4% RTP. That bypass is automatically removed when the associated wide range channel indicates 1E-4% RTP. If a trip channel is bypassed when power is above 1E-4% RTP, it must be considered to be inoperable.

Palisades Nuclear Plant B 3.3.1-20 Revised 10/29/2009

RPS Instrumentation B 3.3.1 BASES LCO 10. Loss of Load Trip (continued)

The LCO requires four Loss of Load Trip Function channels to be OPERABLE in MODE 1 with THERMAL POWER 2: 17% RTP.

The Loss of Load trip may be bypassed or be inoperable with THERMAL POWER < 17% RTP, since it is no longer needed to prevent lifting of the pressurizer safety valves or steam generator safety valves in the eventof a Loss of Load. Loss of Load Trip unit must be considered inoperable if it is bypassed when THERMAL POWER is above 17% RTP.

This LCO requires four RPS Loss of Load auxiliary trip units, relays 305L and 305R, and pressure switches 63/AST-1, 63/AST-2, and 63/AST-3 to be OPERABLE. With those components OPERABLE, a turbine trip will generate a reactor trip.

The LCO does not require the various turbine trips, themselves, to be OPERABLE.

The Nuclear Steam Supply System and Steam Dump System are capable of aCGornmodating the Loss of Load without requiring the use of the above equipment.

The Loss; of Load Trip Function is not credited in the accident analysis; therefore, an Allowable Value for the trip cannot be derived from analytical limits, and is not specified.

11. Containment High Pressure Trip This LCO requires four channels of Containment High Pressure Trip Function to be OPERABLE.

Tile Allowable Value is high enougtl to allo'vv for small pressure increases in containment expected during normal operation (i.e., plant heatup) that are not indicative of an abnormal condition.

The setting i~ low enough to initiate a reactor trip to prevent containment pressure from exceeding desigp pressure following a DBA and ensures the reactor is shutdown before initiation of safety injection and containment spray.

Palisades Nuclear Plant B 3.3.1-21 Revised 10/29/2009

RPS Instrumentation B 3.3.1 BASES LCO (continued) 12. ZPM Bypass The LCO requires that four channels of automatic Zero Power Mode (ZPM) Bypass removal instrumentation be OPERABLE.

Each channel of automatic ZPM Bypass removal includes a shared wide range NI channel, an actuating bistable in the wide range drawer, and a relay in the associated RPS cabinet. Wide Range NI channel 1/3 is shared between ZPM Bypass removal channels A and C; Wide Range NI channel 2/4, between ZPM Bypass removal channels Band D. An operable bypass removal channel must be capable of automatically removing the capability to bypass the affected RPS trip channels with the ZPM Bypass key switch at the proper setpoint.

APPLICABILITY This LCO requires all safety related trip functions to be OPERABLE in accordance with Table 3.3.1-1.

Those RPS trip Functions which are assumed in the safety analyses (all except High Startup Rate and Loss of Load), are required to be operable in MODES 1 and 2, and in MODES 3, 4, and 5 with more than one full-length control rod capable of being withdrawn and PCS boron concentration less than REFUELING BORON CONCENTRATION.

These trip Functions are not required while in MODES 3, 4, or 5, if PCS boron concentration is at REFUELING BORON CONCENTRATION, or when no more than one full-length control rod is capable of being withdrawn, because the RPS Function is already fulfilled. REFUELING BORON CONCENTRATION provides sufficient negative reactivity to assure the reactor remains subcritical regardless of control rod position, and the safety analyses assume that the highest worth withdrawn full-length control rod will fail to insert on a trip. Tilerefore, under these conditions, the safety analyses assumptions will be met without the RPS trip Function.

The High Startup Rate Trip Function is required to be OPERABLE in MODES 1 and 2, but may be bypassed when the associated wide range NI channel indicates below 1 E-4% power, when poor counting statistics may lead to erroneous indication. In MODES 3, 4, 5, and 6, the High Startup Rate trip is not required to be OPERABLE. Wide range channels are required to be OPERABLE in MODES 3, 4, and 5, by LCO 3.3.9, "Neutron Flux Monitoring Channels," and in MODE 6, by LCO 3.9.2, "Nuclear Instrumentation."

Palisades Nuclear Plant B 3.3.1-22 Revised 10/29/2009

RPS Instrumentation B 3.3.1 BASES APPLICABILITY (continued) The High Startup Rate Trip Function is required to be OPERABLE in MODES 1 and 2, but may be bypassed when the associated wide range NI channel indicates below 1 E-4% power, when poor counting statistics may lead to erroneous indication. In MODES 3, 4, 5, and 6, the High Startup Rate trip is not required to be OPERABLE. Wide range channels are required to be OPERABLE in MODES 3, 4, and 5, by LCO 3.3.9, "Neutron Flux Mon:itoring C~annels," and in MODE 6, by LCO 3.9.2, "Nuclear Instrumentation." ~,

The Loss of Load trip is required to be OPERABLE with THERMAL POWER at or above 17% RTP. Below 17% RTP, the ADVs are capable of relieving the pressure due to a Loss of Load event without challenging other overpressure protection.

The trips are deSigned to take the reactor subcritical, maintaining the SLs during AOOs and aSSisting the ESF in providing acceptable consequences during accidents.

ACTIONS The most common causes of channel inoperability are outright failure of loop components or drift of those loop components which is sufficient to exceed the tolerance provided in the plant setpoint analysis. Loop component failures are typically identified by the' actuation of alarms due to the channel failing to the "safe" condition, during CHANNEL CHECKS (when the instrument is compared to the redundant channels), or during the CHANNEL FUNCTIONAL TEST (when an automatic component might not respond properly). Typically, the drift of the loop components is found to be small and results in a delay of actuation rather than a total loss of function. Excessive loop component drift WOUld, most likely, be identified during a CHANNEL CHECK (when the instrument is compared to the redundant channels) or during a CHANNEL CALJBFi!-I.nON (wlleninstrurnent loop components are checked against reference standards).

In the event a channel's trip setpoint is found nonconservative with respect to the Allowable Value, or the transmitter, instrument loop, signal processing electronics, or RPS bistable trip unit is found inoperable, all affected Functions provided by that channel must be declared inoperable, and the plant must enter the Condition for the particular protection Functions affected.

Palisades Nuclear Plant B 3.3.1-23 Revised 10/29/2009

RPS Instrumentation B 3.3.1 BASES ACTIONS (continued) When the number of inoperable channels in a trip Function exceeds that specified in any related Condition associated with the same trip Function, then the plant is outside the safety analysis. Therefore, LCO 3.0.3 is immediately entered if applicable in the current MODE of operation.

A Note has been added to the ACTIONS to clarify the application of the Completion Time rules. The Conditions of this Specification may be entered independently for each Function. The Completion Times of each inoperable Function will be tracked separately for each Function, starting from the time the Condition was entered.

Condition A applies to the failure of a single channel in any required RPS Function, except High Startup Rate, Loss of Load, or ZPM Bypass Removal. (Condition A is modified by a Note stating that this Condition does not apply to the High Startup Rate, Loss of Load, or ZPM Bypass Removal Functions. The failure of one channel of those Functions is addressed by Conditions B, C, or D.)

If one RPS bistable trip unit or associated instrument channel is inoperable, operation is allowed to continue. Since the trip unit and associated instrument channel combine to perform the trip function, this Condition is also appropriate if both the trip unit and the associated instrument channel are inoperable. Though not required, the inoperable channel may be bypassed. The provision of four trip channels allows one channel to be bypassed (removed from service) during operations, placing the RPS in two-out-of-three coincidence logic. The failed channel must be restored to OPERABLE status or placed in trip within 7 days.

Required Action A.i places the Function in a one-out-of-three configuration. In this configuration, common cause failure of dependent channels cannot prevent trip.

The Completion Time of 7 days is based on operating experience, which has demonstrated that a random failure of a second channel occurring during the 7 day period is a low probability event.

Palisades Nuclear Plant B 3.3.1-24 Revised 10/29/2009

RPS Instrumentation B 3.3.1 BASES ACTIONS A.1 (continued)

(continued)

The Completion Time of 7 days is based on operating experience, which has demonstrated that a random failure of a second channel occurring during the 7 day period is a low probability event.

Condition B applies to'the failure of a single High Startup Rate trip unit or associated instrument channel.

If one trip unit or associated instrument channel fails, it must be restored to OPERABLE status prior to entering MODE 2 from MODE 3. A shutdown provides the appropriate opportunity to repair the trip function and conduct the necessary testing. The Completion Time is based on the fact that the safety analyses take no credit for the functioning of this trip.

Condition C applies to the failure of a single Loss of Load or associated instrument channel.

Its:me trip unit or associated instrument channel fails, it must be restored to OPERABLE status prior to THERMAL POWER ~ 17% RTP following a shutdown. If the plant is shutdown at the time the channel becomes inoperable, then the failed channel must be restored to OPERABLE status prior to THERMAL POWER ~ 17% RTP. For this Completion Time, "following a shutdown" means this Required Action does not have to be completed until prior to THERMAL POWER ~ 17% RTP for the first time after the plant has been in MODE 3 following entry into the Condition. The Completion Time trip assures that the plant will not be restarted witll ali inoperaqle Loss of Load trip cllannel.

Palisades Nuclear Plant B 3.3.1-25 Revised 10/29/2009

RPS Instrumentation B 3.3.1 BASES ACTIONS D.1 and D.2 (continued)

Condition D applies when one or more automatic ZPM Bypass removal channels are inoperable. If the ZPM Bypass removal channel cannot be restored to OPERABLE status, the affected ZPM Bypasses must be immediately removed, or the bypassed RPS trip Function channels must be immediately declared to be inoperable. Unless additional circuit failures exist, the ZPM Bypass may be removed by placing the associated "Zero Power Mode Bypass" key operated switch in the normal position.

A trip channel which is actually bypassed, other than as allowed by the Table 3.3.1-1 footnotes, cannot perform its specified safety function and must immediately be declared to be inoperable.

E.1 and E.2 Condition E applies to the failure of two channels in any RPS Function, except ZPM Bypass Removal Function. (The failure of ZPM Bypass Removal Functions is addressed by Condition D.).

Condition E is modified by a Note stating that thjs Condition does not apply to the ZPM Bypass Removal Function.

Required Action E.1 provides for placing one inoperable channel in trip within the Completion Time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Though not required, the other inoperable channel may be (trip channel) bypassed.

Palisades Nuclear Plant B 3.3.1-26 Revised 10/29/2009

RPS Instrumentation B 3.3.1 BASES ACTIONS E.1 and E.2 (continued)

(continued)

This Completion Time is sufficient to allow the operator to take all appropriate actions for the failed channels while ensuring that the risk involved in operating with the failed channels is acceptable. With one channel of protective instrumentation bypassed or inoperable in an untripped condition, the RPS is in a two-out-of-three logic for that function; but with another channel failed, the RPS may be operating in a two-out-of-two logic. This is outside- the-assumptions made in the analyses and should be corrected. To correct the problem, one of the inoperable channels is placed in trip. This places the RPS in a one-out-of-two for that function logic. If any of the other unbypassed channels for that function receives a trip signal, the reactor will trip.

Action E.2 is modified by a Note stating that this Action does not apply to (is not required for) the High Startup Rate and Loss of Load Functions.

One channel is required to be restored to OPERABLE status within 7 days for reasons similar to those stated under Condition A. After one channel is restored to OPERABLE status, the provisions of Condition A still apply to the remaining inoperable channel. Therefore, the channel that is still inoperable after completion of Required Action E.2 must be placed in trip if more than 7 days have elapsed since the initial channel failure. .

F.1 The power range excore channels are used to generate the internal ASI signal used as an input to the TMILP trip. They also provide input to the Thermal Margin Monitors for determination of the Q Power input for the TMILP trip and the VHPT. If two power range excore channels cannot be restored to OPERABLE status, power is restricted or reduced during subsequent operations because of increased uncertainty associated with inoperable power range excore channels which provide input to those trips.

The Completion Time of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> is adequate to reduce power in an orderly manner without challenging plant systems.

Palisades Nuclear Plant B 3.3.1-27 Revised 10/29/2009

RPS Instrumentation B 3.3.1 BASES ACTIONS G.1! G.2.1! and G.2.2 (continued)

Condition G is entered when the Required Action and associated Completion Time of Condition A, B, C, 0, E, or F are not met, or-if the control room ambient air temperature exceeds 90°F.

If the control room ambient air temperature exceeds 90°F, all Thermal Margin Monitor channels are rendered inoperable because their operating temperature limit is exceeded. In this condition, or if the Required Actions and associated Completion Times are not met, the reactor must be placed in a condition in which the LCO does not apply.

To accomplish this, the plant must be placed in MODE 3, with no more than one full-length control rod capable of being withdrawn or with the PCS boron concentration at REFUELING BORON CONCENTRATION in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

The Completion Time is reasonable, based on operating experience, for placing the plant in MODE 3 from full power conditions in an orderly manner and without challenging plant systems. The Completion Time is also reasonable to ensure that no more than one full-length control rod is capable of being withdrawn or that the PCS boron concentration is at REFUELING BORON CONCENTRATION.

SURVEILLANCE The SRs for any particular RPS Function are found in the SR column of REQUIREMENTS Table 3.3.1-1 for that Function. Most Functions are subject to CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION.

SR 3.3.1.1 Performance of the CHANNEL CHECK once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> ensures triat gross failure of instrumentation Ilas not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the two instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. Under most conditions, a CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying that the instrumentation continues to operate properly between each CHANNEL CALIBRATION.

Palisades Nuclear Plant B 3.3.1-28 Revised 10/29/2009

RPS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.1.1 (continued)

REQUIREMENTS (continued) Agreement criteria are determined by the plant staff based on a-combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the transmitter or the signal processing equipment has drifted outside its limits.

The ContainiTlent High Pressure and Loss of Load channels are pressure switch actuated. As such, they have no associated control room indicator and do not require a CHANNEL CHECK.

The Frequency, about once every shift, is based on operating experience that demonstrates the rarity of channel failure. Since the probability of two random failures in redundant channels in any 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> period is extremely low, the CHANNEL CHECK minimizes the chance of loss of protective function due to failure of redundant channels. The CHANNEL CHECK supplements less formal, but more frequent, checks of channel OPERABILITY during normal operational use of the displays associated with the LCO required channels.

SR 3.3.1.2 This SR verifies that the control room ambient air temperature is within the environmental qualification temperature limits for the most restrictive RPS components, which are the Thermal Margin Monitors. These monitors provide input to both the VHPT Function and the TMILP Trip Function. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency is reasonable based on engineering judgement and plant operating experience.

SR 3.3.1.3 A Gaily cali,~Ja,tion (Ileat balallce) is performed wilen THER[\,1AL POWER is;::: 15%. The daily calibration consists of adjusting the "nuclear power calibrate" potentiometers to agree with the calorimetric calculation if the absolute difference is :2: 1.5%. Nuclear power is adjusted via a potentiometer, or THERMAL POWER is adjusted via a Thermal Margin Monitor bias number, as necessary, in accordance with the daily calibration (heat balance) procedure. Performance of the daily calibration ensures that the two inputs to the Q power measurement are indicating accurately with respect to the much more accurate secondary calorimetric calculation.

Palisades Nuclear Plant B 3.3.1-29 Revised 10/29/2009

RPS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.1.3 (continued)

REQUIREMENTS (continued) The Frequency of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is based on plant operating experience and takes into account indications and alarms located in the control room to detect deviations in channel outputs.

The Frequency is modified by a Note indicating this Surveillance must be performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after THERMAL POWER is;::: 15% RTP.

The secondary calorimetric is inaccurate at lower power levels. The 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allows time requirements for plant stabilization, data taking, and instrument calibration.

SR 3.3.1.4 It is necessary to calibrate the power range excore channel upper and lower subchannel amplifiers such that the measured ASI reflects the true core power distribution as determined by the incore detectors. ASI is utilized as an input to the TMILP trip function where it is used to ensure that the measured axial power profiles are bounded by the axial power profiles used in the development of the Tinlet limitation of LCO 3.4.1. An adjustment of the excore channel is necessary only if reactor power is greater than 25% RTP and individual excore channel ASI differs from AXIAL OFFSET, as measured by the incores, outside the bounds of the follOwing table; Allowed Group 4 Group 4 Reactor Rods ;::: 128" withdrawn Rods <128" withdrawn Power

5 100% -0.020:5 (AO-ASI):5 0.020 -0.040:5 (AO-ASI):5 0.040

< 95 -0.033 :5 (AO-ASI) :5 0.020 -0.053 :5 (AO-ASI) :5 0.040

< 90 -0.046 :5 (AO-ASI) :5 0.020 -0.066:5 (AO-ASI) :5 0.040

< 85 -0.060 :5 (AO-ASI) :5 0.020 -0.080 :5 (AO-ASI) :5 0.040

< 80 -0.120 s (I\O-/\SI) s 0.080 -0.140 S (I\O**)l,SI) s 0.100

< 75 -0.120 :5 (AO-ASI) :5 0.080 -0.140 :5 (AO-ASI) :5 0.100

< 70 -0.120 :5 (AO-ASI) :5 0.080 -0.140 :5 (AO-ASI) :5 0.100

< 65 -0.120 :5 (AO-ASI) :5 0.080 -0.140 :5 (AO-ASI) :5 0.100

< 60 -0.160 :5 (AO-ASI) :5 0.120 -0.180 :5 (AO-ASI) :5 0.140

< 55 -0.160 :5 (AO-ASI) :5 0.120 -0.180 :5 (AO-ASI) :5 0.140

< 50 -0.160 :5 (AO-ASI) :5 0.120 -0.180 :5 (AO-ASI) :5 0.140

< 45 -0.160 :5 (AO-ASI) :5 0.120 -0.180 :5 (AO-ASI) :5 0.140

< 40 -0.160 :5 (AO-ASI) :5 0.120 -0.180 :5 (AO-ASI) :5 0.140

< 35 -0.160 :5 (AO-ASI) :5 0.120 -0.180 :5 (AO-ASI) :5 0.140

< 30 -0.160:5 (AO-ASI):5 0.120 -0.180 S (AO-ASI):5 0.140

< 25 Below 25% RTP any AO/ASI difference is acceptable Table values determined with a conservative P'Idr gamma constant of -9505.

Palisades Nuclear Plant B 3.3.1-30 Revised 10/29/2009

RPS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.1.4 (continued)

REQUIREMENTS (continued) Below 25% RTP any difference between ASI and AXIAL OFFSET is acceptable. A Note indicates the Surveillance is not required to have been performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after THERMAL POWER is ~ 25% RTP.

Uncertainties in the excore and incore measurement process make it impractical to calibrate when THERMAL POWER is < 25% RTP. The

. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allows time for plant stabilization, data taking, and instrument calibration.

The 31 day Frequency is adequate, based on operating experience of the excore linear amplifiers and the slow burnup of the detectors. The excore readings are a strong function of the power produced in the peripheral fuel bundles and do not represent an integrated reading across the core. Slow changes in neutron flux during the fuel cycle can also be detected at this Frequency.

SR 3.3.1.5 A CHANNEL FUNCTIONAL TEST is performed on each RPS instrument channel, except Loss of Load and High Startup Rate, every 92 days to ensure the entire channel will perform its intended function when needed. For the TM/LP *Function, the constants associated with the Thermal Margin,Monitors must be verified to be within tolerances.

A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions.

Any setpoint adjustment must be consistent with the assumptions of the current setpoint analysis.

The Frequency of 92 days is based on the reliability analysis presented in topical report CEN-327, "RPS/ESFAS Extended Test Interval Evaluation" (Ref. 5).

Palisades Nuclear Plant B 3.3.1-31 Revised 10/29/2009

RPS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.1.6 REQUIREMENTS (continued) A calibration check of the power range excore channels using the internal test circuitry is required every 92 days. This SR uses an-internally generated test signal to check that the 0% and 50% levels read within limits for both the upper and lower detector, both on the analog meter and on the TMM screen. This check verifies that neither the zero point nor the amplifier gain adjustment have undergone excessive drift since the previous complete CHANNEL CALIBRATION.

The Frequency of 92 days is acceptable, based on plant operating experience, and takes into account indications and alarms available to the operator in the control room.

SR 3.3.1.7 A CHANNEL FUNCTIONAL TEST on the Loss of Load and High Startup Rate channels is performed prior to a reactor startup to ensure the entire channel will perform its intended function.

A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions.

The High Startup Rate trip is actuated by either of the Wide Range Nuclear Instrument Startup Rate channels. NI-1/3 sends a trip signal to RPS channels A and C; NI-2/4 to channels Band D. Since each High Startup Rate channel would cause a trip on two RPS channels, the High Startup Rate trip is not tested when tile reactor is critical.

The four Loss of Load Trip channels are all actuated by a single pressure switch monitoring turbine auto stop oil pressure which is not tested when the reactor is critical. Operating experience has shown that these components usually pass the Surveillance when performed at a Frequency of once per 7 days prior to each reactor startup.

Palisades Nuclear Plant B 3.3.1-32 Revised 10/29/2009

RPS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.1.8 REQUIREMENTS (continued) SR 3.3.1.8 is the performance of a CHANNEL CALIBRATION every 18 months.

CHANNEL CALIBRATION is a complete check of the instrument channel including the sensor (except neutron detectors). The Surveillance verifies that the channel responds to a measured par~meter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drift between successive calibrations to ensure that the channel remains operational between successive tests. CHANNEL CALIBRATIONS must be consistent with the setpoint analysis.

The bistable setpoints must be found to trip within the Allowable Values specified in the LCO and left set consistent with the assumptions of the setpoint analysis. The Variable High Power Trip setpoint shall be verified to reset properly at several indicated power levels during (simulated) power increases and power decreases.

The as-found and as-left values must also be recorded and reviewed for consistency with the assumptions of the setpoint analysis .

. ' .As part of the CHANNEL CALIBRATION of the wide range Nuclear Instrumentation, automatic removal of the ZPM Bypass for the Low PCS Flow, TMILP must be verified to assure that these trips are available when required.

The Frequency is based upon the assumption of an 18 month calibration interval for the determination of the magnitude of equipment drift.

This SR is modified by a Note vV!licil states that it is not necessary to calibrate neutron detectors because they are passive devices with minimal drift and because of the difficulty of simulating a meaningful signal. Slow changes in power range excore neutron detector sensitivity are compensated for by performing the daily calorimetric calibration (SR 3.3.1.3) and the monthly calibration using the incore detectors (SR 3.3.1.4). Sudden changes in detector performance would be noted during the required CHANNEL CHECKS (SR 3.3.1.1).

Palisades Nuclear Plant B 3.3.1-33 Revised 10/29/2009

RPS Instrumentation B 3.3.1 BASES REFERENCES 1. 10 CFR 50, Appendix A, GOC 21

2. 10CFR100

3. IEEE Standard 279-1971, AprilS, 1972

4. FSAR, Chapter 14

5. CEN-327, June 2, 1986, including Supplement 1, March 3, 1989 Palisades Nuclear Plant B 3.3.1-34 Revised 10/29/2009

RPS Instrumentation B 3.3.1 Table B 3.3.1-1 (page 1 of 1)

Instruments Affecting Multiple Specifications Required Instrument Channels Affected Specifications Nuclear Instrumentation Source Range NI-1/3, Count Rate Indication @ C-150 Panel 3.3.8 (#1)

Source Range NI-1/3 & 2/4, Count Rate Signal 3.3.9 & 3.9.2 4

Wide Range NI-1/3 & 2/4, Flux Level 10- Bypass 3.3.1 (#3, 6, 7, 9, & 12)

Wide Range NI-1/3 & 2/4, Startup Rate 3.3.1 (#2)

Wide Range NI-1/3 & 2/4, Flux Level Indication @EC-06 Panel for 3.3.7 3.3.7 (#3) & 3.3.9 Power Range NI-5, 6, 7, & 8, Tq 3.2.1 & 3.2.3 Power Range NI-5, 6, 7, & 8, 0 Power 3.3.1 (#1 & 9)

Power Range NI-5, 6, 7, & 8, ASI 3.3.1 (#9) & 3.2.1 & 3.2.4 Power Range NI-5, 6, 7, & 8, Loss of Load/High Startup Rate Bypass 3.3.1 (#2 & 10)

PCS T-Cold Instruments TT-0112CA, Temperature Signal (SPI ilT Power for PDIL Alarm Circuit) 3.1.6 TT-0112CA & 0122CA, Temperature Signal (C-150) 3.3.8 (#6 & 7)

TT-0122CB, Temperature Signal (PIP ilT Power for PDIL Alarm Circuit) 3.1.6 TT -0112CA & 0122CB, Temperature Signal (L TOP) 3.4.12.b.1 TT-0112CC & 0122CD (PTR-0112 & 0122) Temperature Indication 3.3.7 (#2)

TT-0112 & 0122 CC & CD, Temperature Signal (SMM) 3.3.7 (#5)

TT-0112 & 0122 CA, CB, CC, & CD, Temperature Signal (0 Power & TMM) 3.3.1 (#1 & 9) & 3.4.1.b PCS T-Hot Instruments TT -0112HA, Temperature Signal (SPI ilT Power for PDIL Alarm Circuit) 3.1.6 TT-0112HA & 0122HA, Temperature Signal (C-150) 3.3.8 (#4 & 5)

TT -0122HB, Temperature Signal (PIP AT Power for PDIL Alarm Circuit) 3.1.6 TT -0112 & 0122 HC & HD, Temperature Signal (SMM) 3.3.7 (#5)

TT-0112HC & 0122HD (PTR-0112. & 0122) Temperature Indication 3.3.7 (#1)

TT-0112 & 0122 HA, HB, HC, & HD, Temperature Signal (0 Power & TMM) 3.3.1 (#1 & 9)

Thermal Margin Monitors PY-0102A, B, C, & D 3.3.1 (#1 & 9)

Pressurizer Pressure Instruments PT-0102A, B, C, & D, Pressure Signal (RPS & SIS) 3.3.1 (#8 & 9) &

3.3.3 (#1.a & 7a)

PT -01 04A & B, Pressure Signal (LTOP & SDC Interlock) 3.4.12.b.1 & 3.4.14 PT -01 05A & B, Pressure Signal (WR Indication & LTOP) 3.3.7 (#5) & 3.4.12.b.1 PI-0110, Pressure Indication @ C-150 Panel 3.3.8 (#2)

SG Level Instruments LT-0751 & 0752 A, B, C, & D, Level Signal (RPS & AFAS) -------------------'--i3.1(#4--&-S)&--- .--.-.-.

I . 3.3.3 (#4.a & 4.b)

I LI-0757 & 0758 A & B, Wide Range Level Indication 3.3.7 (#11 & 12)

I LI-0757C & 0758C, Wide Range Level Indication @ C-150 Panel 3.3.8 (#10 & 11)

! SG Pressure Instruments I PT-0751 & 0752 A, B, C, & D. Pressure Signal (RPS & SG IS.Qlation) I 3.3.1 (#6 & 7) &

1-=-=-::-::::-:::-:-::-::-:==-=-=-=--=_ _--:----:-:---:-:-______________13.3.3 (#2a, 2b, 7b, 7c) ~

PIC-0751 & 0752 C & D, Pressure Indication PI-0751 E & 0752E, Pressure' Indication @ C-150 Panel i 3.3.7 (#13 & 14)

! 3.3.8 (#8 & 9) l Containment Pressur_e__I_n_st_f_um_e_nt_s___..___________ ---1

~S-1801, 1802A, 1803, & 1804A, Switch Output (ESF)

.P?-1 E3.~1A, 1802, 180~'-"_, & 1804, Switch Ou~put (ESF). _ _

t PS-1801, 1802, 1803, & 1804, Switch Output(~R_P=S"-=)c = c - - - - - - - - 13.3.1 (#11) ~

3.3.3 (#5.a)

~.3.3 (#5.b)

~

~J

---J Note: The information provided in this table is intended for use as an aid to distinguish those instrument channels which provide more than one required function and to describe which specifications they affect. The information in this table should not be taken as inclusive for a!1 instruments nor affected specifications.

Palisades Nuclear Plant B 3.3.1-35 Revised 10/29/2009

LTOP System B 3.4.12 B 3.4 PRIMARY COOLANT SYSTEM (PCS)

B 3.4.12 Low Temperature Overpressure Protection (LTOP) System BASES BACKGROUND The LTOP System controls PCS pressure at low temperatures so the integrity of the Primary Coolant Pressure Boundary (PCPB) is not compromised by violating the Pressure and Temperature (PIT) limits of 10 CFR 50, Appendix G (Ref. 1). The reactor vessel is the limiting PCPB c9mponent requirin.g such protection. LCO 3.4.3, "PCS Pressure and Temperature (PIT) Limits," provides the allowable combinations for operational pressure and temperature during cooldown, shutdown, and heatup to keep from violating the Reference 1 requirements during the LTOP MODES.

The toughness of the reactor vessel material decreases at low temperatures. As the vessel neutron exposure accumulates, the material toughness decreases and becomes less resistant to pressure stress at low temperatures (Ref. 2). PCS pressure, therefore, is maintained low at low temperatures and is increased only as temperature is increased.

The potential for vessel overpressurization is most acute when the PCS is water solid, which occurs only while shutdown. Under that condition, a pressure fluctuation can occur more quickly than an operator can react to relieve the condition. Exceeding the PCS pn: limits by a significant amount could cause brittle fracture of the reactor vessel. LCO 3.4.3 requires administrative control of PCS pressure and temperature during heatup and cooldown to prevent exceeding the PIT limits.

This LCO provides PCS overpressure protection by limiting coolant injection capability and requiring adequate pressure relief capacity.

Limiting coolant injection capability requires all High Pressure Safety Injection (HPSI) pumps be incapable of injection into the PCS when any PCS cold leg temperature is < 300°F. The pressure relief capacity requires either two OPERABLE redundant Power Operated Relief Valves (PORVs) or the PCS depressurized and a PCS vent of sufficient size.

One PORV or the PCS vent is the overpressure protection d.evice that acts to terminate an increasing pressure event.

Palisades Nuclear Plant B 3.4.12-1 Revised 10/29/2009

LTOP System B 3.4.12 BASES BACKGROUND With limited coolant injection capability, the ability to provide core (continued) coolant addition is restricted. The LCO does not require the chemical and volume control system to be deactivated or the Safety Injection Signals (SIS) blocked. Due to the lower pressures in the LTOP MODES and the expected core decay heat levels, the chemical and volume control system can provide adequate flow via the makeup control valve. If conditions require the use of an HPSI pump for makeup in the event of loss of inventory, then a pump can be made available through manual actions.

The LTOP System for pressure relief consists of two PORVs with temperature dependent lift settings or a PCS vent of sufficient size.

Two PORVs are required for redundancy. One PORV has adequate relieving capability to prevent overpressurization for the allowed coolant injection capability.

PORV Requirements As designed for the LTOP System, an "open" signal is generated for each PORV if the PCS pressure approaches a limit determined by the LTOP actuation logic. The actuation logic monitors PCS pressure and cold leg temperature to determine when the LTOP overpressure setting is approach.ed. If the indicated pressure meets or exceeds the calculated value, a PORV is opened.

The LCO presents the PORV setpoints for LTOP by specifying Figure 3.4.12-1, "LTOP Setpoint Limit." Having the setpoints of both valves within the limits of the LCO ensures the PIT limits will not be exceeded in any analyzed event.

When a PORV is opened in an increasing pressure transient, the release of coolant causes the pressure increase to slow and reverse. As the

. PORV releases coolant, tile system pressure decreases until a reset pressure is reached and the valve closed. The pressure continues to decrease below the reset pressure as the valve closes.

Palisades Nuclear Plant B 3.4.12-2 Revised 10/29/2009

LTOP System B3.4.12 BASES BACKGROUND PCS Vent Requirements (continued)

Once the PCS is depressurized, a vent exposed to the containment atmosphere will maintain the PCS at containment ambient pressure in an PCS overpressure transient if the relieving requirements of the transient do not exceed the capabilities of the vent. Thus, the vent path must be capable of relieving the flow resulting from the limiting LTOP mass injection or heatup transient and maintaining pressure below the PfT limits. The required vent capacity may be provided by one or more vent paths.

Reference 3 has determined that any vent path capable of relieving 167 gpm at a PCS pressure of 315 psia is acceptable. The 167 gpm flow rate is based on an assumed charging imbalance due to interruption of letdown flow with three charging pumps operating, a 40°F per hour PCS heatup rate, a 60°F per hour pressurizer heatup rate, and an initially depressurized and vented PCS. Neither HPSI pump nor Primary Coolant Pump (PCP) starts need to be assumed with the PCS initially depressurized, because LCO 3.4.12 requires both HPSI pumps to be incapable of injection into the PCS and LCO 3.4.7, "PCS Loops-MODE 5, Loops Filled," places restrictions on starting a PCP. ,

The pressure relieving ability of a vent path depends not only upon the area of the vent opening, but also upon the configuration of the piping connecting the vent opening to the PCS. A long, or restrictive piping connection may prevent a larger vent opening from providing adequate flow, while a smaller opening immediately adjacent to the PCS could be adequate. The areas of multiple vent paths cannot simply be added to determine the necessary vent area.

The following vent path examples are acceptable:

1. Fiemoval of a steam generator primary rnanway;

2. Removal of the pressurizer manway;

3. Removal of a PORV or pressurizer safety valve;

4. Both PORVs and associated block valves open; and

5. Opening of both PCS vent valves MV-PC514 and MV-PC515.

Palisades Nuclear Plant B 3.4.12-3 Revised 10/29/2009

LTOP System B 3.4.12 BASES BACKGROUND Reference 4 determined that venting the PCS through MV-PC514 and (continued) MV-PC515 provided adequate flow area. The other listed examples provide greater flow areas with less piping restriction and are ther~fore acceptable. Other vent paths shown to provide adequate capacity could also be used. The vent path(s) must be above the level of reactor coolant, to prevent draining the PCS.

One open PORV provides sufficient flow area to prevent excessive PCS pressure. However, if the PORVs are elected as the vent path, both valves must be used to meet the single failure criterion, since the PORVs are held open against spring pressure by energizing the operating solenoid.

When the shutdown cooling system is in service with MO-3015 and MO-3016 open, additional overpressure protection is provided by the relief valves on the shutdown cooling system. References 5 and 6 show that this relief capacity will prevent the PCS pressure from exceeding its pressure limits during any of the above mentioned events.

APPLICABLE Safety analyses (Ref. 7) demonstrate that the reactor vessel is SAFETY ANALYSES adequately protected against exceeding the Referenc;:e 1 PIT limits during shutdown. In MODES 1 and 2, and in MODE 3 with all PCS cold leg temperature at or exceeding430°F" the pressurizer safety valves preve.nt PCS pressure from exceeding the Reference 1 limits. Below 430°F, overpressure prevention falls to the, OPERABLE PORVs or to a depressurized PCS and a sufficiently sized PCS vent. Each of these means has a limited overpressure relief capability.

The actual temperature at which the pressure in the PIT limit curve falls below the pressurizer safety valve setpoint increases as the reactor vessel material toughness decreases due to neutron embrittlement. Each lilne the PIT limit curves are revised, the LTOP System Stlould LJe re-evaluated to ensure its functional requirements can still be satisfied using the PORV method or the depressurized and vented PCS congition.

Reference 3 contains the acceptance limits that satisfy the LTOP requirements. Any change to the PCS must be evaluated against these analyses to determine the impact of the change on the LTOP acceptaf)ce limits.

Palisades Nuclear Plant B 3.4.12-4 Rev'ised 10/29/2009

LTOP System B 3.4.12 BASES APPLICABLE Transients that are capable of overpressurizing the PCS are SAFETY ANALYSES categorized as either mass injection or heatup transients (continued)

Mass Injection Type Transients

a. Inadvertent safety injection; or

b. Charging/letdown flow mismatch.

Heatup Type Transients

a. Inadvertent actuation of pressurizer heaters;

b. Loss of Shutdown Cooling (SOC); or

c. PCP startup with temperature asymmetry within the PCS or between the PCS and steam generators.

Rendering both HPSI pumps incapable of injection is required during the LTOP MODES to ensure that mass injection transients beyon{:lJbe capability of the LTOP overpressure protection system, do not occur. The Reference 3 analyses demonstrate that either one PORV or the pes vent can maintain PCS pressure below limits when three charging pump are actuated. Thus, the LCO prohibits the operation of both HPSI pumps and does not place any restrictions on charging pump operation.

Fracture mechanics analyses were used to establish the applicable temperature range for the LTOP LCO as below 430°F. At and above this temperature, the pressurizer safety valves provide the reactor vessel pressure protection. The vessel materials were assumed to have a neutron irradiation accumulation equal to 2.192 E19 nvt.

Palisades Nuclear Plant B 3.4.12-5 Revised 10/29/2009

LTOP System B 3.4.12 BASES APPLICABLE PORV Performance SAFETY ANALYSES (continued) The fracture mechanics analyses show that the vessel is protecteo when the PORVs are set to open at or below the setpoint curve specified in Figure 3.4.12-1 of the accompanying LCO. The setpoint is derived by modeling the performance of the LTOP System, assuming the limiting allowed LTOP transient. The valve qualification process considered pressure overshoot and undershoot beyond the PORV opening and closing setpoints, resulting from signal processing and valve stroke times.

The PORV setpoints at or below the derived limit ensure the Reference 1 limits will be met.

The PORV setpoints will be re-evaluated for compliance when the PrT limits are revised. The PrT limits are periodically modified as the reactor vessel material toughness decreases due to embrittlement caused by neutron irradiation. Revised PrT limits are determined using neutron fluence projections and the results of examinations of the reactor vessel material irradiation surveillance specimens. The Bases for LCO 3.4.3 discuss these examinations.

The PORVs are considered active components. Thus, the failure of one PORV represents the worst case, single active failure. '

........

PCS Vent Performance With the PCS depressurized, analyses show the required vent size is capable of mitigating the limiting allowed LTOP overpressure transient. In that event, this size vent maintains PCS pressure less than the maximum PCS pressure on the PrT limit curve.

The PCS vent is passive and is not subject to active failure.

LTOP System satisfies Criterion 2 of 10 CFR 50.36(c)(2).

Palisades Nuclear Plant B 3.4.12-6 Revised 10/29/2009

LTOP System B 3.4.12 BASES LCO This LCO is required to ensure that the LTOP System is OPERABLE.

The LTOP System is OPERABLE when both HPSI pumps are incapable of injecting into the PCS and pressure relief capabilities are OPERABLE.

Violation of this LCO could lead to the loss of low temperature overpressure mitigation and violation of the Reference 1 limits as a result of an operational transient.

To limit the coolant injection capability, LCO 3.4.12.a require'S both HPSI

. pumps be incapable of injecting into the PCS. LCO 3.4. ~.a is modified by two Notes. Note 1 only requires both HPSI pumps to be incapable of injecting into the PCS when any PCS cold leg temperature is < 300°F.

When all PCS cold leg temperatures are 2 300°F, a start of both HPSI pumps in conjunction with a charging/letdown imbalance will not cause the PCS pressure to exceed the 10 CFR 50 Appendix G limits. Thus, a restriction on HPSI pump operation when all PCS cold leg temperatures are 2 300°F is not required. Note 2 is provided to assure that this LCO does not cause hesitation in the use of a HPSI pump for PCS makeup if it is needed due to a loss of shutdown cooling or a loss of PCS inventory.

The elements of the LCO that provide overpressure mitigation through pressure relief are:

a. Two OPERABLE PORVs; or

b. The PCS depressurized and vented.

A PORV is OPERABLE for LTOP when its block valve is open, its lift setpoint is set consistent with Figure 3.4.12-1 in the accompanying LCO and testing has proven its ability to open at that setpoint, and motive power is available to the valve and its control circuit.

A PCS vent is OPERABLE when open with an area capable ofrelie,-:ing 2:: -167 gpm at a PCS pressure of 3-15 psia. , *.

Each of these methods of overpressure prevention is capable of mitigating the limiting LTOP transient.

Palisades Nuclear Plant B 3.4_12-7 Revised 10/29/2009

LTOP System B 3.4.12 BASES APPLICABILITY This LCO is applicable in MODE 3 when the temperature of any PCS cold leg is < 430°F, in MODES 4 and 5, and in MODE 6 when the reactor vessel head is on. The pressurizer safety valves provide overpressure protection that meets the Reference 1 PIT limits at and above 430°F.

When the reactor vessel head is off, overpressurization cannot occur.

LCO 3.4.3 provides the operational PIT limits for all MODES.

LCO 3.4.10, "Pressurizer Safety Valves," requires the OPERABILITY of the pressurIZer safety valves t~at provide overpressure protection during MODES 1 and 2, and MODE 3. with all PCS cold leg temperatures

430°F.

Low temperature overpressure prevention is most critical during shutdown when the PCS is water solid, and a mass addition or a heatup transient can cause a very rapid increase in PCS pressure with little or no time available for operator action to mitigate the event.

ACTIONS A Note prohibits the application of LCO 3.0.4.b to inoperable PORVs used for LTOP. There is an increased risk associated with entering MODE 4 from MODE 5 with PORVs used for LTOP inoperable and the.

provisions of LCO 3.0.4.b, which allow entry into a MODE or other specified condition in the Applicability with the LCO not met after performance of a risk assessment addressing inoperable systems and components, should not be applied in this circumstance.

With one or two HPSI pumps capable of injecting into the PCS, overpressurization is possible.

Tile immediate Completion Time to initiate actions to restore restricted coolant injection capability to the PCS reflects the importance of maintaining overpressure protection of the PCS.

• A Palisades Nuclear Plant B 3.4.12-8 Revised 10/29/2009

LTOP System B 3.4.12 BASES ACTIONS B.1 (continued)

With one required PORV inoperable and pressurizer water level S 57%,

the required PORV must be restored to OPERABLE status within a Completion Time of 7 days. Two valves are required to meet the LCO requirement and to provide low temperature overpressure mitigation while withstanding a single failure of an active component.

The Completion Time is based on only one PORV being required to mitigate an overpressure transient, the likelihood of an active failure of the remaining valve path during this time period being very low, and that a steam bubble exists in the pressurizer. Since the pressure response to a transient is greater if the pressurizer steam space is small or if the PCS is solid, the Completion Time for restoration of a PORV flow path to service is shorter. The maximum pressurizer level at which credit can be taken for having a bubble (57%, which provides about 700 cubic feet of steam space) is based on judgment rather than by analysis. This level provides the same steam volume to dampen pressure transients as would be available at full power. This steam volume provides time for operator action (if the PORVs failed to operate) in the interval between an inadvertent SIS and PCS pressure reaching the 10 CFR 50, Appendix G pressure limit. The time available for action would depend upon the existing pressure and temperature when the inadvertent SIS occurred.

The consequences of operational events that will overpressurize the PCS are more severe at lower temperature (Ref. 8). With the pressurizer water level> 57%, less steam volume is available to dampen pressure increases resulting from an inadvertent mass injection or heatup transients. Thus, with one required PORV inoperable and the pressurizer water level> 57%, tile Completion Time to restore tile required PORV to OPERABLE status is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Completion Time to restore the required PORV to OPERABLE status when the pressurizer water level is > 57%, which usually occurs in MODE 5 or in MODE 6 when the vessel head is on, is a reasonable amount of time to investigate and repair PORV failures without a lengthy period with only one PORV OPERABLE to protect against overpressure events.

Palisades Nuclear Plant B 3.4.12-9 Revised 10/29/2009

LTOP System B 3.4.12 BASES ACTIONS 0.1 (continued)

If two required PORVs are inoperable, or if the Required Actions and the associated Completion Times are not met, or if the LTOP System is inoperable for any reason other than Condition A, B, or C, the PCS must be depressurized and a vent established within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The vent must be sized to provide a relieving capability of ;:: 167 gpm at a pressure of 315 psia which ensures the flow capacity is greater than that required for the worst case mass injection transient reasonable during the applicable MODES. This action protects the PCPB from a low temperature overpressure event and a possible brittle failure of the reactor vessel.

The Completion Time of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to depressurize and vent the PCS is based on the time required to place the plant in this condition and the relatively low probability of an overpressure event during this time period due to operator attention and administrative requirements.

SURVEILLANCE SR 3.4.12.1 REQUIREMENTS To minimize the potential for a low temperature overpressure event by limiting the mass injection capability, both HPSI pumps are verified to be incapable of injecting into the PCS. The HPSI pumps are rendered incapable of injecting into the PCS by means that assure that a single event cannot cause overpressurization of the PCS due to operation of the pump. Typical methods for accomplishing this are by pulling the HPSI pump breaker control power fuses, racking out the HPSI pump motor circuit breaker, or closing the manual discharge valve.

SR 3.4.12.1 is modified by a Note which only requires the SR to be met when complying with LCO 3.4.12.a. When all pes cold leg temperature are;:: 300°F, a start of both HPSI pumps in conjunction with a charging/letdown imbalance will not cause tile PCS ~ressure to exceed the 10 CFR 50 Appendix G limits. Thus, this SR is only required when any PCS cold leg temperature is reduced to less than 300°F.

The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> interval considers operating practice to regularly assess potential degradation and to verify operation within the safety analysis.

Palisades Nuclear Plant B 3.4.12-10 Revised 10/29/2009

LTOP System B 3.4.12 BASES SURVEILLANCE SR 3.4.12.2 REOUIREMENTS (continued) SR 3.4.12.2 requires a verification that the required PCS vent, capable of relieving;::: 167 gpm at a PCS pressure of 315 psia, is OPERABLE by verifying its open condition either:

a. Once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for a valve that is not locked open; or b.

-

Once every 31 days for a valve that is locked open.

The passive vent arrangement must only be open to be OPERABLE.

This Surveillance need only be performed if vent valves are being used to satisfy the requirements of this LCO. This Surveillance does not need to be performed for vent paths relying on the removal of a steam generator primary manway cover, pressurizer manway cover, safety valve or PORV since their position is adequately addressed using administrative controls and the inadvertent reinstallation of these components is unlikely. The Frequencies consider operating experience with mispositioning of unlocked and locked vent valves, respectively.

SR 3.4.12.3 The PORV block valve must be verified open every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to provide the flow path for each required PORV to perform its function when actuated. The valve can be remotely verified open in the main control room.

The block valve is a remotely controlled, motor operated valve. The power to the valve motor operator is not required to be removed, and the manual actuator is not required locked in the inactive position. Thus, the block valve can be closed in the event the PORV develops excessive 10akage or does not close ,sticks open) after relieving all overpressure event.

The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Frequency considers operating experience with accidental movement of valves having remote control and position indication capabilities available where easily monitored. These considerations include the administrative controls over main control room access and equipment control.

Palisades Nuclear Plant B3.4.12-11 Revised *10/29/2009

LTOP System B 3.4.12 BASES SURVEILLANCE SR 3.4.12.4 REQUIREMENTS (continued) Performance of a CHANNEL FUNCTIONAL TEST is required every 31 days. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay This is acceptable because all of the Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions. PORV actuation could depressurize the PCS and is not required. The 31 day Frequency considers experience with equipment reliability.

A Note has been added indicating this SR is required to be performed 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after decreasing any PCS cold leg temperature to < 430°F. This Note allows a discrete period of time to perform the required test without delaying entry into the MODE of Applicability for LTOP. This option may be exercised in cases where an unplanned shutdown below 430°F is necessary as a result of a Required Action specifying a plant shutdown, or other plant evolutions requiring an expedited cooldown of the plant.

The test must be performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after entering the LTOP MODES.

SR 3.4.12.5 Performance of a CHANNEL CALIBRATION on each required PORV actuation channel is required every 18 months to adjust the entire channel so that it responds and the valve opens within the required LTOP range and with accuracy to known input.

The 18 month Frequency considers operating experience with equipment reliability and is consistent with the typical refueling outage schedule.

Palisades Nuclear Plant B 3.4.12-12 Revised 10/29/2009

LTOP System B 3.4.12 BASES REFERENCES 1. 10 CFR 50, Appendix G

2. Generic Letter 88-11

3. CPC Engineering Analysis, EA-A-PAL-92-095-01

4. CPC Engineering Analysis, EA-TCD-90-01

5. CPC Engineering Analysis, EA-E-PAL-89-040-1

6. CPC Corrective Action Document, A-PAL-91-011

7. FSAR, Section 7.4

8. Generic Letter 90-06 Palisades Nuclear Plant B 3.4.12-13 Revised 10/29/2009

SWS B 3.7.8 B 3.7 PLANT SYSTEMS B 3.7.8 Service Water System (SWS)

BASES BACKGROUND The SWS provides a heat sink for the removal of process and operating heat from safety related components during a Design Basis Accident (DBA) or transient. During normal operation or a normal shutdown, the SWS also provides this function for various safety related and nonsafety related components. The safety related function is covered by this LCO.

The isolation of the SWS to components or systems may render those components inoperable but does not affect the OPERABILITY of the SWS System.

The SWS consists of three pumps connected in parallel taking suction from a common intake structure supplied by Lake Michigan. The discharge of the pumps flow into a common header before splitting into three headers (two critical headers for safety-related equipment and a single non-critical header for non safety-related equipment). The return piping from the three headers join into a common line and discharge to the cooling tower makeup basin. A train of SWS shall be that equipment electrically connected to a common safety bus necessary to remove heat from the various heat loads. There are two SWS trains, each associated with a Safeguards Electrical Train which are described in Specification 3.8.9, "Distribution Systems - Operating." The SWS train associated with the Left Safeguards Train consists of one SWS pump (P-7B), associated piping, valves, and controls for the equipment to perform their safety function. The SWS train associated with the Right Safeguards Train consists of two SWS pumps (P-7A, P-7C), associated piping, valves, and controls for the equipment to perform their safety function. The pumps and valves are remote manually aligned, except in the unlikely event of a Loss Of Coolant Accident (LOCA).

SWS components receive three automatic actuation signals, a Safety Injection Signal (SIS), a Recirculation Actuation Signal (RAS), or a Diesel Generator (DG) start signal:

1. SIS starts the SWS pumps, isolates the non-critical service water header, and realigns the Containment Air Cooler (CAC) service water valves to the post accident cooling configuration.

Palisades Nuclear Plant B 3.7.8-1 Revised 10/29/2009

SWS B 3.7.8 BASES BACKGROUND 2. RAS realigns the CCW heat exchanger service water outlet valves (continued) for maximum cooling.

3. A DG start signal opens the DG lube oil and jacket water cooler inlet valves.

The DG which powers two SWS pumps (P-7A, P-7C), also powers the fans associated witb VHX-1, VHX-2, and VHX-3 (V-1A, V-2A and V-3A).

This is necessary because if reliance tor containment cooling is placed on CACs, at least two service water pumps must be OPERABLE to provide the necessary service water flow to assure OPERABILITY of the CACs.

The Service Water System cools three groups of loads. The SWS loads are described in the FSAR (Ref. 1), the major loads are:

1. Critical loads inside the Containment, Containment Air Coolers VHX-1, VHX-2, VHX-3, (and VHX-4)

2. Critical loads outside the Containment, and Diesel Generators 1-1 and 1-2 Component Cooling Heat Exchangers E-54A and E-54B Engineered Safeguards Room Coolers VHX-27 A and VHX-27B Control Room HVAC Coolers VC-1 0 and VC 11 Instrument Air Compressor C-2A and C-2C After Coolers

3. Non-critical loads in the Turbine Building* .

Each of these groups of loads can be cooled by the flow from one SWS pump. During normal operation, when SWS flow from the CACs and CCW heat exchangers is throttled by temperature control valves, two SWS pumps can provide the required flow for all three groups of loads.

During post accident conditions, with all other SWS and related system components OPERABLE, one hundred percent of the mquired SWS post accident cooling capability can be provided by anyone SWS pUll1p. If SWS or related systems have components out of serVice, additional SWS pumps may be required to provide the required cooling capability.

For post accident cooling, the Engineered Safety Features signals reposition several valves to maximize containment cooling and conserve SWS flow. Initially, a safety injection signal will start the SWS pumps, realign the SWS valves for the CACs (which cool the containment atmosphere), and close the non-critical SWS header isolation valve.

Palisades Nuclear Plant B 3.7.8-2 Revised 10/29/2009

SWS B 3.7.8 BASES BACKGROUND Subsequently, if the Safety Injection Refueling Water Tank has been (continued) emptied, a RAS will realign the SWS outlet valves on the CCW heat exchangers (CCW cools the Shutdown Cooling Heat Exchangers, which cool the containment spray flow). The occurrence of these automatic actions will provide the one hundred percent of the required post accident SWS cooling capability while limiting the SWS flow requirement to that which can be

• provided by two SWS pumps.

If the- Containment Air Coolers are not needed for post accident containment cooling. SWS flow to the containment may then be isolated, further reducing the required SWS post accident cooling capability to that which can be provided by one SWS pump.

One hundred percent of the required SWS post accident cooling capability can be provided by anyone SWS pump if SWS flow both to the non-critical header and to the critical loads inside the containment are capable of being isolated.

1. The capability to isolate SWS flow to the non-critical SWS header requires its isolation valve, CV-1359, to be OPERABLE.

2. The allowance to isolate SWS flow to the containment requires the ability to provide post accident containment cooling without reliance on CACs.

The capability to isolate SWS flow to the containment requires one SWS Containment Isolation Valve, CV-0824 or CV-0847, to be OPERABLE.

One hundred percent of the required SWS post accident cooling capability can be provided by any two SWS pumps if SWS.flow either to the non-critical header or to the critical loads inside the containment are capable of being isolated.

Qne hunqred percent of the required SWS post accident cooling capability can be provided by three E?WS'pumps even with SWS flow being provided to both the CACs and the Non-critical SWS header.

Additional information about th.e design and operation of the SWS, along with a list of the components served, is presented in the FSAR, Section 9.1 (Ref ..1). The principal safety related functions of the SWS is the removal of decay heat from the reactor via the Component Cooling Water (CCW)

System and the removal of heat from the containment atmosphere via the CACs.

Palisades Nuclear Plant B 3.7.8-3 Revised 10/29/2009

SWS B 3.7.8 BASES APPLICABLE The design basis of the SWS is for one SWS train, in conjunction with SAFETY ANALYSES the CCW System and a 100% capacity containment cooling system (containment spray, CACs, or a combination), removing core decay heat between 20 to 40 minutes following a design basis LOCA. This prevents the containment sump fluid from increasing in temperature during the recirculation phase following a LOCA and provides for a gradual reduction in the temperature of this fluid as it is supplied to the Primary Coolant System by the safety injection pumps. The SWS is designed to perform its function with a single failure of any active component, assuming the loss of offsite power.

The SWS, in conjunction with the CCW System, also cools the plant from Shutdown Cooling (SOC) entry Condition, as discussed in the FSAR, Section 6.1 (Ref. 2) to MODE 5 during normal and post accident operations.

The time required for this evolution is a function of the number of CCW and SOC System trains that are operating. This assumes that the maximum Lake Michigan water temperature of LCO 3.7.9, "Ultimate Heat Sink (UHS),"

occurs simultaneously with maximum heat loads on the system.

The SWS satisfies Criterion 3 of 10 CFR 50.36(c)(2).

LCO Two SWS trains are required to be OPERABLE to provide the required redundancy to ensure that the system functions to remove post accident heat loads, assuming the worst single active failure occurs coincident with the loss of offsite power.

The SWS train associated with the Left Safeguard Electrical Distribution Train is considered OPERABLE when:

a. SWS pump P-7B is OPERABLE; and

b. The associated piping, valves, and instrumentation and controls required to perform the safety related function are OPERABLE.

The SWS train associated with the Right Safeguards Electrical Distribution Train is OPERABLE when:

a. SWS pumps P-7A and P-7C are OPERABLE; and

b. The associated piping, valves, and instrumentation and controls required to perform the safety related function are OPERABLE.

The isolation of SWS from other components or systems not required for safety may render those components or systems inoperable but does not affect the OPERABILITY of the SWS System.

Palisades Nuclear Plant B 3.7.8-4 Revised 10/29/2009

SWS B 3.7.8 BASES APPLICABILITY In MODES 1, 2, 3, and 4, the SWS System is a normally operating system, which is required to support the OPERABILITY of the equipment serviced by the SWS and required to be OPERABLE in these MODES. In MODES 5 and 6, the OPERABILITY requirements of the SWS are determined by the systems it supports.

ACTIONS

~'"

Condition A is applicable whenever one or more SWS trains is inoperable.

Action A.i requires restoration of both trains to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time is based on the assumption that at least 100% of the required SWS post accident cooling capability (that assumed in the safety analyses) is available. (If, however, less than 100% of the SWS post accident cooling is available, Condition C must also be entered.)

Mechanical system LCOs typically provide a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time under conditions when a required system can perform its required safety function, but may not be able to do so assuming an additional failure. When operating in accordance with the Required Actions of an LCO Condition, it is not necessary to be able to cope with an additional single failure.

The SWS system can provide one hundred percent of the required post accident cooling capability following the occurrence of ariy'slngle"activ~

failure. Therefore, the SWS function can be met during conditions when those components which could be deactivated by a single active failure are known to be inoperable. Under that condition, however, the ability to provide the function after the occurrence of an additional failure cannot be guaranteed. Therefore, continued operation with one or more trains inoperable is allowed only for a limited time.

B.1 and B.2 Condition B is applicable when the Required Actions of CondiHonAcannot be completed within the required Completion Time. Condition A is applicable whenever one or more trains is inoperable. Therefore, when Condition B is applicable, Condition A is also applicable. (If less than 100% of the post accident SWS cooling capability is available, Condition C must be entered as well.) Being in Conditions A and B concurrently maintains both Completion ~_

Time clocks for instances where equipment repair allows exit fmm Coneition B while the plant is still within the applicable conditions of the LCO.

Palisades Nuclear Plant B 3.7.8-5 Revised 10/29/2009

SWS B 3.7.8 BASES ACTIONS B.1 and B.2 (continued)

If the inoperable SWS trains cannot be restored to OPERABLE status within the associated required Completion Time of Condition A, the plant must be placed in a MODE in which the LCO does not apply. To achieve this status, the plant must be placed in at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions iR-an orderly manner and without challenging plant systems.

Condition C is applicable with one or more trains inoperable when there is less than 100% of the required SWS post accident cooling capability available. Condition A is applicable whenever one or more trains is inoperable. Therefore, when this Condition is applicable, Condition A is also applicable. Being in Conditions A and C concurrently maintains both Completion Time clocks for instances where equipment repair restores 100%

of the required SWS post accident cooling capability while the LCO is still applicable, allowing exit from Condition C (and LCO 3.0.3).

The Service Water System cools three groups of loads:

1. Critical loads inside the Containment,

2. . Criti'cal loads outside the Containment, and

3. Non-critical loads in the Turbine Building.

As discussed in the Background section of these bases, each of these groups of loads can be cooled by the flow from one SWS pump.

One hundred percent of the required SWS post accident cooling capability can be provided by anyone SWS pump if:

1. The non-criticClI SINS header isolation valve, CV1859, is OPER~BLE, and

.,' .

2.. 'pla'nt conditions allow adequate containment cooling to be provided without reliance on CACs and one SWS Containment Isolation Valve, CV-0824 or CV-0847, is OPERABLE.

One hundred p~Jcent of.!he required SWS post accident cooling capability can be provided by any two SWS pumps if:

Palisades Nuclear Plant B 3.7.8-6 Revised 10/29/2009

SWS B 3.7.8 BASES ACTIONS (continued)

1. The non-critical SWS header isolation valve, CV-1359, is .

OPERABLE, or

2. Plant conditions allow adequate containment cooling to be provided without reliance on CACs and one SWS Containment Isolation Valve, CV-0824 or CV-0847, is OPERABLE.

One hundred percent of the required SWS post accident cooling capability can be provided by three SWS pumps even with SWS flow being provided to both the CACs and the Non-critical SWS header.

With less than 100% of the required SWS post accident cooling capability available, the plant is in a condition outside the assumptions of the safety analyses. Therefore, LCO 3.0.3 must be entered immediately.

SURVEILLANCE SR 3.7.8.1 REQUIREMENTS Verifying the correct alignment for manual, power operated, and automatic valves in the SWS flow path ensures that the proper flow paths

. exist for SWS operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since they are verified to be in the correct position prior to locking, sealing, or securing. This SR also does not apply to valves that cannot be inadvertently misaligned, such as check valves. This Surveillance does not require any testing or valve manipulation; rather, it involves verification that those valves capable of potentially being mispOSitioned are in the correct position.

This SR is modified by a Note indicating that the isolation of SWS to components or systems may render those components inoperable but uoes not affect the OPI::RA81L1TY of the SWS.

The 31 day Frequency is based on engineering judgment, is consistent with the procedural controls governing valve operation, and ensures correct valve positions.

Palisades Nuclear Plant B 3.7.8-7 Revised 10/29/2009

SWS B 3.7.8 BASES SURVEILLANCE SR 3.7.8.2 REQUIREMENTS This SR verifies proper automatic operation of the SWS valves on an actual or simulated actuation signal. Specific signals (e.g., safety injection) are tested under Section 3.3, "Instrumentation." This Surveillance is not required .for valves that are locked, sealed, or otherwise secured in the required position under administrative controls.

This SR is m.Qdified by a Note which states this SR is only required to be met in MODES 1, 2, and 3. The instrumentation providing the input signal"'*

is not required in MODE 4, therefore, to keep consistency with Section 3.3, "Instrumentation," the SR is not required to be met in this MODE.

Operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency. Therefore, the Frequency is acceptable from a reliability standpoint.

SR 3.7.8.3 The SR verifies proper automatic operation of the SWS pumps on an actual or simulated actuation signal in the "with standby power available" mode which tests the starting of the pumps by the SIS-X relays. The starting of the pumps by the sequencer is performed in Section 3.8, "Electrical Power Systems." This SR is modified by a Note which states this SR is not required to be met in MODE 4. The instrumentation

_ providing the. input signal is not required in MODE 4, therefore, to keep consistency with Section 3.3, "Instrumentation," the SR is not required to be met in this MODE. Operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency. Therefore, the Frequency is acceptable from a reliability standpoint.

[\EFEIl:'::NC:'::S 1. FSAR, Section 9.1

2. FSAR, Section 6.1 Palisades Nuclear Plant B 3.7.8-8 Revised 10/29/2009

SFP Boron Concentration B3.7.15 B 3.7 PLANT SYSTEMS B 3.7.15 Spent Fuel Pool (SFP) Boron Concentration BASES BACKGROUND As described in LCO 3.7.16, "Spent Fuel Pool Storage," fuel assemblies are stored in the fuel storage racks in accordance with criteria based on initial enrichment, discharge burn up, and decay time.

The criteria were based on the assumption that 850 ppm of soluble boron was present in the spent fuel pool. The pool is required to be maintained at a boron concentration of;::: 1720 ppm. Criterion 2 of 10 CFR 50.36 (c) (2) requires that criticality control be achieved without credit for soluble boron. However, in 1998 the NRC documented requirements that could be established to maintain criticality below 0.95. This is documented in "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants", Laurence I. Kopp, U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Reactor Systems Branch, February 1998. The precedent of taking credit for soluble boron in spent fuel pool water to provide criticality control has also been established. Soluble boron credit was used in the Westinghouse Spent Fuel Rack Criticality Analysis Methodology described in WCAP-14416-NP-A and that methodology was approved for use by an NRC Safety Evaluation dated October 25, 1996. The criteria discussed above was developed using a method that closely followed the Westinghouse methodology. Additionally the requirements specified by the NRC guidance are in place at Palisades.

APPLICABLE A fuel assembly could be inadvertently loaded into a fuel storage rack SAFETY ANALYSES location not allowed by LCO 3.7.16 (e.g., an insufficiently depleted or insufficiently decayed fuel assembly). Another type of postulated accident is associated with a fuel assembly that is dropped onto the fully loaded fuel pool

.storage rack. Either incident could have a positive reactivity effect, decreasing the margin to criticality. However, the negative reactivity effect of the soluble boron compensates for the increased reactivity caused by either one of the two postulated accident scenarios.

The concentration of dissolved boron in the SFP satisfies Criterion 2 of 10 CFR 50.36(c)(2).

LCO The specified concentration of dissolved boron in the SFP preserves the assumptions used in the analyses of the potential accident scenarios described above. This concentration of dissolved boron is the minimum required concentration for fuel assembly storage and movement within the SFP.

APPLICABILITY This LCO applies whenever fuel assemblies are stored in the spent fuel pool.

Palisades Nuclear Plant B3.7.15-1 Amendment No.1-GG, 2G+, 236

SFP Boron Concentration B 3.7.15 BASES ACTIONS The ACTIONS are modified by a Note indicating that LCO 3.0.3 does not apply.

If moving irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the fuel movement is independent of reactor operation. Therefore, inability to suspend movement of fuel assembUes is not sufficient reason to require a reactor shutdown ..

A.i. and A.2 When the concentration of boron in the spent fuel pool is less than required, immediate action must be taken to preclude an accident from happening or to mitigate the consequences of an accident in progress.

This is most efficiently achieved by immediately suspending the movement of fuel assemblies. This does not preclude the movement of fuel assemblies to a safe position. In addition, action must be immediately initiated to restore boron concentration to within limit.

SURVEILLANCE SR 3.7.15.1 REQUIREMENTS This SR verifies that the concentration of boron in the spent fuel pool is within the required limit. As long as this SR is met, the analyzed incidents are fully addressed. The 7 day Frequency is appropriate because no major replenishment of pool water is expected to take place over a short period of time.

REFERENCES None Palisades Nuclear Plant B 3.7.15-2 Amendment No. -+-&9, aG-7, 236

Spent Fuel Pool Storage B 3.7.16 B 3.7 PLANT SYSTEMS B 3.7.16 Spent Fuel Pool Storage BASES BACKGROUND The fuel storage facility is designed to store either new (nonirradiated) nuclear fuel assemblies, or used (irradiated) fuel assemblies in a vertical configuration underwater. The storage pool is sized to store 892 fuel assemblies, which includes storage for failed fuel canisters.

The fuel storage racks are grouped into two regions, Region I and Region II per Figure B 3.7.16-1. The racks are designed as a Seismic Category I structure able to withstand seismic events. Region I contains racks in the spent fuel pool having a 10.25 inch center-to-center spacing and a single rack in the north tilt pit having an 11.25 inch by 10.69 inch center-to-center spacing. Region II contains racks in both the spent fuel pool and the north tilt pit having a 9.17 inch center-to-center spacing. Region I has restrictive loading patterns to address degradation of neutron absorbing material in the Region I racks. The loading patterns accommodate some face-adjacent fuel assemblies. Region I also has provisions for storing non-fissile bearing components. Because of the smaller spacing and poison concentration, Region II has limitations for fuel storage. Further information on limitations can be found in Section 4.0, "Design Features." These limitations (e.g., enrichment, burnup, loading patterns) are sufficient to maintain a kelt of s 0.95 when flooded with borated water and keff < 1.0 when flooded with unborated water.

APPLICABLE The fuel storage facility was originally designed for noncriticality by use SAFETY ANALYSES of adequate spacing, and "flux trap" construction, whereby the fuel assemblies are inserted into neutron absorbing stainless steel cans.

The current criticality calculations also take credit for soluble boron to prevent criticality.

The spent fuel pool storage meets the requirements specified in "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants", Laurence I. Kopp, U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Reactor Systems Branch, February 1998. This document established the requirements for use of soluble boron to maintain k8ff sO.95.

The spent fuel pool storage satisfies Criterion 2 of 10 CFR 50.36(c)(2).

LCO The restrictions for Region I in Specification 4.3.1.1 on fuel assembly enrichment and the storage pool loading pattern, and on the placement of non-fissile bearing components, ensure that the keff of the spent fuel Palisades Nuclear Plant B 3.7.16-1 Amendment No.tW, aw, 236

Spent Fuel Pool Storage B3.7.16 BASES pool will always remain::;; 0.95, assuming the pool to be flooded with water borated to 850 ppm. Non-fissile bearing components shall be stored in accordance with Specification 4.3.1.1 j. The restrictions are consistent with the criticality safety analyses performed for the spent fuel pool.

The restrictions for Region II in Table 3.7.16-1, in the accompanying LCO, on fuel assembly enrichment and minimum burnup combinations, ensure that the keff of the spent fuel pool will always remain::;; 0.95, assuming the pool to be flooded with water borated to 850 ppm. The restrictions are consistent with the criticality safety analyses performed for the spent fuel pool according to Table 3.7.16-1, in the accompanying LCO. Fuel assemblies not meeting the criteria of Table 3.7.16-1 shall be stored in accordance with Specification 4.3.1.1.

Specification 4.3.1.1 describes U-235 enrichment restrictions for fuel assemblies stored in Region I based on maximum nominal planar average. The term "nominal" describes the design enrichment specified for an assembly. The criticality calculations that support the Region I storage requirements include a manufacturer's fuel enrichment tolerance of +/-0.05 weight percent U-235. Specification 4.3.1.1 does not include the manufacturer's fuel enrichment tolerance.

The term "maximum" refers to an assembly's limiting nominal planar average U-235 enrichment. Palisades' fuel assembly design~yhave several distinct axial planar regions, and each region may have a different nominal planar average U-235 enrichment. Additionally, fuel assembly enrichments may vary from pin to pin within a given axial planar region. The criticality analysis conservatively assumes each pin is loaded with the nominal enrichment for that planar region. The highest nominal planar average enrichment of the distinct axial planar regions is considered to be the maximum nominal planar average enrichment for that assembly. This value is used to verify that storage requirements have been met. The manufacturer's fuel enrichment tolerance of +/-0.05 weight percent is excluded from this value.

APPLICABILITY This LCO applies whenever any fuel assembly or non-fissile bearing component is stored in the spent fuel pool or the north tilt pit.

ACTIONS The ACTIONS are modified by a Note indicating that LCO 3.0.3 does not apply.

If moving irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the fuel movement is independent of reactor Palisades Nuclear Plant B 3.7.16-2 Amendment No . .:t-gg, aw, 236

Spent Fuel Pool Storage 83.7.16 BASES operation. Therefore, in either case, inability to move fuel assemblies is not sufficient reason to require a reactor shutdown.

When the configuration of fuel assemblies or non-fissile bearing components stored in the spent fuel pool is not in accordance with the storage requirements, immediate action must be taken to make the necessary movement(s) to bring the configuration into compliance with

.-

the requirements.

SURVEILLANCE SR 3.7.16.1 REQUIREMENTS This SR verifies by administrative means that the combination of fuel assembly maximum nominal planar average enrichment and proposed fuel assembly placement is in accordance with Specification 4.3.1.1 prior to placing the assembly in a Region I storage location. This SR also verifies by administrative means that non-fissile bearing component storage will be in accordance with Specification 4.3.1.1 j. prior to placing the component in a Region I storage location.

This SR also verifies by administrative means that the combination of initial enrichment, burnup and decay time of the fuel assembly is in accordance with Table 3.7.16-1 in the accompanying LCO prior to placing the fuel assembly in a Region II storage location.

REFERENCES None Palisades Nuclear Plant 83.7.16-3 Amendment No. :t-gg, aw, 236

Spent Fuel Pool Storage 83.7.16 8ASES

---+--.~--~--~.---~-+~-.---------- ~--- -- - -- --_. -._- ---

---

.-~,--- .. ---~-----, ".- - '"""7.~--/--"-- ------ ---- - " - --I

-~

Ii ""'J":'" Ii , /

!l / i1 11 , 11 ,l / /

'/,.. /,1

\/-l /

,,,I, I:

I~======'====='=' ===="=",'==="==="','

I, iii '/

/

I

/

__ ,-

[I II , / .( ,/ /

i!/ ,', ',','

7 , Ii /, II I il

"

I; iIi , '

/  ;'

Region I of the main pool is comprised of Region IA and Region lB.

,; -', Region I of the north tilt

+---. "

-~---.- ~~~--.-.

pit is designated as

,Region IE. These sub-regions are defined in Specification 4.3.1.

Figure 8 3.7.16-1 (page 1 of 1)

Spent Fuel Pool Arrangement Palisades Nuclear Plant 83.7.16-4 . Amendment No. +W, 2-G+, 236