ML14282A509

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Handouts to Applicants (Folder 3)
ML14282A509
Person / Time
Site: Susquehanna  Talen Energy icon.png
Issue date: 08/11/2014
From:
Operations Branch I
To: D'Antonio J
Susquehanna
References
TAC U01896
Download: ML14282A509 (55)
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Text

{{#Wiki_filter:Table R Abnormal Rad Levels/Radiological Effluents Category I General Emergency Radiological Effluents I RG1 (Pg R-22) Page 2 of 27 IC Dose at the EMERGENCY PLAN BOUNDARY Resulting from an Actual or Imminent Release of Gaseous Radioactivity Exceeds 1000 mrem TEDE or 5000 mrem Child Thyroid CDE for the Actual or Projected Duration of the Release Using Actual Meteorology. Modes: ALL EAL Threshold Value (1 or 2 or 3 or 4 or 5) Note: If dose assessment results are available at the time of declaration, the classification should be based on EAL 2 instead of EAL 1. While necessary declarations should not be delayed awaiting results, the dose assessment should be initiated I completed in order to determine if the classification should be subsequently escalated.

1.

VALID Noble Gas vent stack monitor reading(s) that exceeds or is expected to exceed a site total release rate of 6.2E+9)lCi/min for 15 minutes or longer and Dose Projections are not available. OR

2.

VALID dose assessment using actual meteorology indicates projected doses greater than 1000 mrem TEDE or 5000 mrem child thyroid CDE at or beyond the EPB. OR

3.

A VALID reading sustained for 15 minutes or longer on the RMS perimeter radiation monitoring system greater than 1000 mR!hr. (The RMS perimeter radiation monitoring system is only monitored when the TSC or EOF is activated) OR

4.

Field survey results indicate closed window dose rates exceeding 1 000 mR!hr expected to continue for more than one hour at or beyond the EPB. OR

5.

Analyses of field survey samples indicate child thyroid dose commitment at or beyond the EPB of 5000 mrem assuming one hour of inhalation. Site Area Emergency RS1 (Pg 26) IC Dose at the EMERGENCY PLAN BOUNDARY Resulting from an Actual or Imminent Release of Gaseous Radioactivity Exceeds 100 mrem TEDE or 500 mrem Child Thyroid CDE for the Actual or Projected Duration of the Release. Modes: ALL EAL Threshold Value (1 or 2 or 3 or 4 or 5) Note: If dose assessment results are available at the time of declaration, the classification should be based on EAL 2 instead of EAL 1. While necessary declarations should not be delayed awaiting results, the dose assessment should be initiated I completed in order to determine if the classification should be subsequently escalated.

1.

VALID Noble Gas vent stack monitor reading(s) that exceeds or is expected to exceed a site total release rate of 6.2E+8)lCi/min for 15 minutes or longer and Dose Projections are not available. OR

2.

VALID dose assessment using actual meteorology indicates projected doses greater than 1 00 mrem TEDE or 500 mrem child thyroid CDE at or beyond the EPB. OR

3.

A VALID reading sustained for 15 minutes or longer on the RMS perimeter radiation monitoring system greater than 100mR!hr. (The RMS perimeter radiation monitoring system is only monitored when the TSC or EOF is activated) OR

4.

Field survey results indicate closed window dose rates exceeding 100 mR!hr expected to continue for more than one hour at or beyond the EPB. OR

5.

Analyses of field survey samples indicate child thyroid dose commitment at or beyond the EPB of 500 mrem assuming one hour of inhalation. Alert RA1 (Pg 30) IC Any UNPLANNED Release of Gaseous or Liquid Radioactivity to the Environment that Exceeds 200 Times Technical Requirements Manual Limits for 15 Minutes or Longer. Modes: ALL EAL Threshold Value (1 or 2 or 3 or 4)

1.

VALID Noble Gas vent stack monitor reading(s) that exceeds a site total release rate of 2.0E+8 )lCilmin and that is sustained for 15 minutes or longer. OR

2.

Confirmed sample analyses for airborne releases indicate total site release rates for 15 minutes or longer resulting in dose rates at the SITE BOUNDARY of: A. Noble gases > 1.0E+5 mrem/year whole body, OR B. Noble gases> 6.0E+5 mrem/year skin, OR C. 1-131, 1-133, H-3, and particulates with half-lives greater than 8 days >3.0E+5 mrem/year to any organ (inhalation pathways only). OR

3.

Confirmed sample analyses for liquid releases indicate concentrations in excess of 200 times the Technical Requirements Manual liquid effluent limits for 15 minutes or longer.

4.

OR VALID reading on any liquid effluent monitor that exceeds two hundred times the alarm setpoint established by a current radioactivity discharge permit for 15 minutes or longer Attachment R EP-RM-004 Revision 1 Page 19 of 195 Notification of Unusual Event RU1 (Pg 34) IC Any UNPLANNED Release of Gaseous or Liquid Radioactivity to the Environment that Exceeds Two Times the Technical Requirements Manual Limits for 60 Minutes or Longer. Modes: ALL EAL Threshold Value (1 or 2 or 3 or 4)

1.

VALID Noble Gas vent stack monitor reading(s) that exceeds a total site release rate of 2.0E+6 )lCilmin and that is sustained for 60 minutes or longer. OR

2.

Confirmed sample analyses for airborne releases indicate total site release rates, with a release duration of 60 minutes or longer, resulting in dose rates at the SITE BOUNDARY of : A. Noble gases >1 000 mrem/year whole body, OR B. Noble gases >6000 mrem/year skin, OR C. 1-131, 1-133, H-3 and particulates with half-lives greater than 8 days > 3000 mrem/year to any organ (inhalation pathway only). OR

3.

Confirmed sample analyses for liquid releases indicate concentrations with a release duration of 60 minutes or longer in excess of two times the Technical Requirements Manual liquid effluent limits.

4.

OR VALID reading on any liquid effluent monitor that exceeds two times the alarm setpoint established by a current radioactivity discharge permit for 60 minutes or longer

Table R Abnormal Rad Levels/Radiological Effluents Abnonnal Radiation RA2 (Pg 38) Levels IC Release of Radioactive Material or Increases in Radiation Levels within the Facility that Impedes Operation of Systems Required to Maintain Safe Operations or to Establish Or Maintain Cold Shutdown. Modes: ALL EAL Threshold Value (1 or 2)

1.

VALID radiation reading> 15 mR/hr in the Main Control Room, the Radwaste Control Room or both the Security Control Center (SCC) and Alternate Security Control Center (ASCC). OR

2.

VALID radiation monitor readings> 10 Rlhr in areas requiring infrequent access to maintain plant safety functions (Table R-1). Irradiated Fuel RA3 (Pg 43) Accidents IC Damage to Irradiated Fuel or Loss of Water Level that has or will Result in the Uncovering of Irradiated Fuel Outside the Reactor Vessel. Modes: ALL EAL Threshold Value (1 or 2 or 3)

1.

UNPLANNED VALID Refuel Floor Area Radiation Monitor (Table R-2) readings greater than 1 R/hr. OR

2.

Water level < 22 feet above the RPV flange for the reactor refueling cavity that will result in irradiated fuel uncovering. OR

3.

Water level < 22 feet above seated irradiated fuel for the spent fuel pool that will result in irradiated fuel uncovering. Page 3 of 27 IC Attachment R EP-RM-004 Revision 1 Page 20 of 195 RU2 (Pg 40) Unexpected Increase in Plant Radiation. Modes: ALL EAL Threshold Value (1 or 2)

1.

A. Uncontrolled water level decrease in the reactor refueling cavity, fuel transfer canal or spent fuel pool with all irradiated fuel assemblies remaining covered by water as indicated by any of the following on either unit: Unexpected Fuel Pool Water Low Level alarm OR Skimmer Surge Tank Low Level alarm OR Visual observation of an uncontrolled water level drop below a fuel pool skimmer surge tank inlet, OR Observation of water draining down the outside wall of Primary Containment. AND B. UNPLANNED VALID Refuel Floor Area Radiation Monitor (Table R-2) readings greater than 500 mR!hr. OR

2.

UNPLANNED VALID Area Radiation Monitor readings increase by a factor of 1000 over normal* levels.

  • Normal levels can be considered as the highest reading in the past twenty-four hours excluding the current peak value.

RB Area Elevation (ft) 749 719 670 645 L_- ARM Number 14 47 (44 U-2) 49 Page 4 of 27 Table R-1 Reactor Buildina Radiation Manit ARM Number High ARM Channel Range Description 52 RWCU Recirc Pp Access 54 Fuel Pool Pump Area 50 CRD North 51 CRD South 53 Remote Shutdown Room 48 HPCI PP Turbine Room 57 RCIC PP Turbine Room 55 RHR A C PP Room 56 RHR B D PP Room Table R-2 Refuel Floor Area Radiation Monitors Description Range (mR/hr) Spent Fuel Pool Area 0.1 -1000 Spent Fuel Pool Area 0.1 -1000 Refuel Floor Area 102 - 106 Attachment R EP-RM-004 Revision 1 Page 21 of 195

ABLE F - FISSION PRODUCT BARRIER DEGRADATION I BarriP-r I 1. FuP-1 r.l::ui R:uriP-r Pn ~n tn ~~

b.

RPV Level

c.

RCS Leak Rate or Containment Isolation Failure or Breach/Bypass

e. Drywell Radiation
f. Emergency Director/Recovery Manager Judgement Page 2 of 30 A. High MSL Flow AND High Steam Tunnel Temperature annunciators.

OR B. Direct report of steam release. Drywell Pressure ~ psig. Indication of a RCS leak inside drywell. (Pg. 61) Any condition in the judgement of the Emergency Director or Recovery Manager that indicates Loss of the RCS barrier. (Pg. 64) FS 1: Site Area Emergency OR gpm inside Primary Containment. (Pg. 57)

2.

Unisolable primary system leakage outside Primary Containment as indicated by: A. Any Reactor Building area exceeds Max Normal Reactor Building Temperature Limit per Table F-1. OR B. Any Reactor Building area exceeds Max Normal Reactor Building Radiation Limit per Table F-2. Any condition in the judgement Emergency Director or Recovery Manager that indicates Potential Loss of the RCS barrier. (Pg. 64) Loss OR Potential Loss Of ANY Two Barriers. Attachment F EP-RM-004 Revision 1 Page 46 of 195 OR automatic valves in any one line penetrating Primary Containment to close AND a downstream pathway to the environment exists. (Pg. 67)

2.

Intentional venting per EP-DS-004 is performed. OR

3.

Unisolable primary system leakage outside Primary Containment as indicated by: A. Any Reactor Building areas exceed Max Safe Reactor Building Temperature Limit per Table F-3. OR B. in Drywell Pressure following initial increase in pressure. (Pg. 71) Any condition in the judgement of the Emergency Director or Recovery Manager that indicates Loss of the PRIMARY CONTAINMENT barrier. (Pg. 74) FA1 : Ale OR

2. Drywell Hydrogen or Suppression Chamber Hydrogen > 6% AND Drywell Oxygen or Suppression Chamber Any condition in the judgement of the Emergency Director or Recovery Manager that indicates Potential Loss of the PRIMARY CONTAINMENT barrier.

ANY Loss OR ANY Potential Loss Of EITHER Fuel Clad OR RCS. FU 1: Notification of Unusual Event ANY Loss OR ANY Potential Loss Of Primary Containment

Table F-1 Max Normal Reactor Building Temperature Limit RB Area Max Normal Elevation Area Temperature Temp (ft) (oF) 818 General Area 110 779 General Area 110 749 General Area 110 RWCU-Pump Room 120 RWCU-Heat Exch Room 120 RWCU-Penetration Room 120 719 General Area 110 Main Steam Line Tunnel 157 683 General Area 110 HPCI Pipe Routing Area 120 RCIC Pipe Routing Area 120 670 General Area 110 645 HPCI-Equip Area 120 HPCI-Emerg Area Cooler 120 645 RCIC-Emerg Area Cooler 120 RCIC-Equip Area 120 645 RHR Equip Area 1 110 645 RHR Equip Area 2 110 645 CS Pump Room A 110 645 CS Pump Room 8 110 645 R8 Sump Room 110 Page 3 of 30 RB Area Elevation (ft) 818 749 719 670 645 Table F-2 Max Normal Reactor Building Radiation Limit ARM Number Arm Channel Description 35 Cask Stor Area 14 Spent Fuel Grit Mon 15 Refuel Floor North (South U2) 42 Refuel Floor West 47 (44 U2) Spent Fuel Grit Mon 8 RWCU Recirc PP Access 10 Fuel Pool PP Area 5 CRD North 6 CRD South 16 Remote Shutdown Room 3 HPCI PP & Turbine Room 2 RCIC PP & Turbine Room 25 RHR A C PP Room 1 RHR 8 D PP Room Attachment F EP-RM-004 Revision 1 Page 47 of 195 Alarm Level Hi Alarm Hi Alarm Hi Alarm Hi Alarm Hi Alarm

Table F-3 Max Safe Reactor Building Temperature RB Area Max Safe Elevation Area Temperature Temp (ft) (oF) 818 General Area 120 779 General Area 120 749 General Area 120 RWCU-Pump Room 147 RWCU-Heat Exch Room 147 RWCU-Penetration Room 131 719 General Area 120 Main Steam Line Tun 177 683 General Area 120 HPCI Pipe Routing Area 167 RCIC Pipe Routing Area 167 670 General Area 120 6A5 HPCI-Equip Area 167 HPCI-Emerg Area Cooler 167 645 RCIC-Emerg Area Cooler 167 RCIC-Equip Area 167 645 RHR Equip Area 1 142 645 RHR Equip Area 2 142 I 645 CS Pump Room A 142 645 CS Pump Room B 142 645 RB Sump Room 125 Page 4 of 30 Table F-4 Max Safe Reactor Building Radiation Monitors RB Area ARM Number Arm Channel Elevation (ft) Description I 818 49 Refuel Floor Area 749 52 RWCU Recirc PP Access 54 Fuel Pool Pump Area 719 50 CRD North 51 CRD South 670 53 Remote Shutdown Room 645 48 HPCI PP & Turbine Room 57 RCIC PP & Turbine Room 55 RHR A C Pump Room 56 RHR 8 D PurDJ:> Room Max Safe Attachment F EP-RM-004 Revision 1 Page 48 of 195 Radiation Levels Per E0-000-1 04 (R/HR) 10 10 10 10 10

a ~.BLE M -SYSTEMS MALFUNCTIONS cGategory I GENERAL EMERGENCY Loss of AC Power I MG1 (Pg 79) Loss of DC Power Failure of Reactor Protection System Page 2 of 35 IC Prolonged Loss of All Offsite Power and Prolonged Loss of All Onsite AC Power to Essential Busses. Modes: 1, 2, 3 EAL Threshold Value Loss of power from Startup Transformer 10 AND 20 to either unit. AND All 4.16 kV ESS Busses on either unit are de-energized. AND A Restoration of at least two 4.16 kV ESS Busses on each unit within 4 hours is not likely. OR B. RPVWater Level<.". MG3 (Pg 87) IC Failure of the Reactor Protection System to Complete an Automatic Scram and Manual Scram was NOT Successful and there is Indication of an Extreme Challenge to the Ability to Cool the Core. Modes: 1, 2 EAL Threshold Value lndication(s) exist that indicate that Reactor Protection System setpoint was exceeded. AND RPS, ARI, and Manual Scram/ARI fail to initiate and complete a scram that reduces reactor power to < I%. AND A Reactor water level cannot be maintained >.... OR B. The combination of RPV Pressure and Suppression Pool Temperature cannot be maintained below the HCTL curve, Figure M-1. NOTE: Although the HCTL curve is not evaluated in E0-000-1 03 until the reactor is shutdown, the curve must be used to consider entry into this EAL SITE AREA EMERGENCY MS1 (Pg 81) IC Loss of All Offsite Power and Loss of All Onsite AC Power to Essential Busses. Modes: 1, 2, 3 EAL Threshold Value Loss of power from Startup Transformer 10 AND 20 to either unit. AND All 4.16 kV ESS Busses on either unit are de-energized. AND Failure to restore power to at least two 4.16KV ESS busses on each unit within fifteen minutes from the time of loss of both offsite and ON SITE AC power. MS2 (Pg 86) Loss of All Vital DC Power. Modes: 1, 2, 3 EAL Threshold Value Loss of all vital DC power to either unit based on less than 105 volts on the 125 VDC main distribution busses 1 D612 (2D612), 1 D622 (2D622), 1 D632 (2D632), AND 1 D642 (2D642) for> 15 minutes. NOTE: Busses do not trip on undervoltage condition. MS3 (Pg 91) IC Failure of Reactor Protection System to Complete or Initiate an Automatic Reactor Scram once a Reactor Protection System Setpoint has been Exceeded and Manual Scram was NOT Successful. Modes: 1, 2 EAL Threshold Value lndication(s) exist that indicate that Reactor Protection System setpoint was exceeded. AND RPS, ARI, and Manual Scram/ARI fail to initiate and complete a scram that reduces reactor power to < I%. ALERT MA1 (Pg 83) IC AC Power Capability to Essential Busses Reduced to a Single Power Source for Greater than 15 Minutes such that Any Additional Single Failure Would Result in Station Blackout. Modes: 1, 2, 3 EAL Threshold Value (1 or 2)

1.

Loss of power from Startup Transformer 10 AND 20 to either unit for> 15 minutes. OR AND Onsite AC power is reduced to a single 4.16 kV ESS Bus on either unit.

2.

Loss of power from Startup Transformer 10 OR 20 to either unit for> 15 minutes. AND On site AC _29wer is not available. MA3 (Pg 93) IC Failure of Reactor Protection System to Complete or Initiate an Automatic Reactor Scram once a Reactor Protection System Setpoint has been Exceeded and Manual Scram was Successful. Modes: 1, 2 EAL Threshold Value lndication(s) exist that indicate that Reactor Protection System setpoint was exceeded AND RPS automatic scram did not reduce reactor power to < I%. AND A Manual Scram or ARI initiates and reduces reactor power to <I%. Attachment M EP-RM-004 Revision 1 Page 76 of 195 NOTICE OF UNUSUAL EVENT MU1 (Pg 85) IC Loss of all Offsite Power to Essential Busses for Greater than 15 Minutes. Modes: 1, 2, 3 EAL Threshold Value Loss of power from Startup Transformer 10 AND 20 to either unit for> 15 minutes.

,IBLE M -SYSTEMS MALFUNCTIONS oss of Communications Technical Specifications Inadvertent Criticality Figure M-1 HEAT CAPACilYTEMPERATURE LIMIT 260

25Q,
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I -24"' I I I u... v r r I I l, ~ ~~ r -~ f I ] 1 I 1 ! j. ~ I : ~, -210 1 I 0 ~ 200 to ~ -1~ c-* ( t __, : -** f r-*. r.,. t-w 190

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--~ ) ~ ---~ "'"~ -T. -*....-;;;~ I y-"'""" ..J 180 c I I I T--.---r-- 0 170 ~ 160 z 150 0 140 ~ 130 w 120 0:::

a. 110

~ 100 Cl) 90 80: 70~~~~~~-.-.~~~~.-.-.-~~~~-.-.~~~~ 1.Z 13 14 15 16 17 18 19 20 21 2Z 23 24 25 26 27 2.8 29 30 31 32 33 34 35 36 37 38 SUPPRESSION POOl LEVEL (f1') Page 4 of 35 RPVPRESS (PSIG) TABLE M-1 ON SITE/OFFSITE COMMUNICATIONS CAPABILITY SYSTEM ON SITE OFFSITE UHF Radio X Commercial Telephone X X Systems Loss of dedicated conference X lines to offsite agencies FTS-2001 (ENS) X Plant PA System X Plant Cellular Telephone X Telecopy Transmittal X Sound Powered Phones X Attachment M EP-RM-004 Revision 1 Page 78 of 195 MUS (Pg 106) IC UNPLANNED Loss of all ON SITE or Offsite Communications Capabilities. Modes: 1, 2, 3 EAL Threshold Value (1 or 2)

1.

UNPLANNED loss of all ON SITE communications capability per Table M-1 affecting the ability to perform routine operations. OR

2.

.UNPLANNED loss of all offsite communications capability per Table M-1. MU9 (Pg 108) IC Inability to Reach Required Shutdown within Technical Specification Limits. Modes: 1, 2, 3 EAL Threshold Value Plant is not brought to required operating mode within Technical Specifications LCO Action Statement Time. MU10 (Pg 109) IC Inadvertent Criticality. Modes: 3 EAL Threshold Value An UNPLANNED extended positive period observed on nuclear instrumentation. Table M-2 Significant transients SCRAM Reactor Recirculation System Runback ( >25% thermal power changes) ECCS Initiations UNPLANNED Thermal Power Changes > 10% Load Reiect > 25% electrical load Table M-3 Safety System Annunciators/Indicators ECCS Containment Isolation Reactor Trip Process or Effluent Radiation Monitors Electrical Distribution/Diesel Generators Table M-4 Safety Function Indicators Reactor Power-Nuclear Instrumentation Displays, Full Core Display, SIP panel displays Decay Heat Removal-valve and pump indications for RHR, Core Spray, HPCI, RCIC, Suppression Pool level and TemJJ_erature Containment Safety Functions-Pressure indication, Hydrogen/Oxygen concentrations, radiation levels, Coolant System integrity-RPV level, RPV pressure, Containment pressure, Containment Radiation level

Table C Cold Shutdown/Refueling System Malfunctions Category GENERAL EMERGENCY SITE AREA EMERGENCY Loss of AC Power Loss of DC Power Decay Heat Removal Page 2 of 34 ALERT CA1 (Pg 162) IC Loss of All Offsite Power and Loss of All ON SITE AC Power to Essential Busses. Modes: 4, 5, D EAL Threshold Value Loss of power from Startup Transformer 10 AND 20 to either unit. AND All 4.16 kV ESS Busses on either unit are de-energized. AND Failure to restore power to at least two 4.16KV ESS busses on each unit within fifteen minutes from the time of loss of both offsite and ON SITE AC power. CA3 (Pg 165) IC Inability to Maintain Plant in Cold Shutdown with Irradiated Fuel in the RPV. Modes: 4, 5 EAL Threshold Value (1 or 2 or 3)

1.

With Secondary Containment and RCS integrity1 not established, an UNPLANNED event results in RCS temperature >. °F. OR

2.

With Secondary Containment established and RCS integrity 1 not established, an UNPLANNED event results in RCS temperature >.oF for> 20 minutes2. OR

3.

An UNPLANNED event results in RCS temperature > 200°F for> 60 minutes2 or results in an RCS pressure increase of greater than 20 psig. 1NOTE: By definition, in Mode 5 RCS integrity is not established. 2NOTE: If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced then this EAL is not applicable. Attachment C EP-RM-004 Revision 1 Page 157 of 195 NOTICE OF UNUSUAL EVENT CU1 (Pg 163) IC Loss of All Offsite Power to Essential Busses for Greater than 15 Minutes. Modes: 4, 5 EAL Threshold Value Loss of power from Startup Transformer 10 AND 20 to either unit for > 15 minutes. CU2 (Pg 164) IC UNPLANNED Loss of Required DC Power for Greater than 15 Minutes. Modes: 4, 5 EAL Threshold Value UNPLANNED loss of all vital DC power to either unit based on less than 105 volts on the 125 VDC main distribution busses 1 D612 (2D612), 10622 (2D622), 10632 (2D632), AND 10642 (2D642) AND Failure to restore power to one required DC bus within 15 minutes from the time of loss. Note: Busses do not trip on undervoltage condition. CU3 (Pg 167) IC UNPLANNED Loss of Decay Heat Removal Capability with Irradiated Fuel in the RPV. Modes: 4, 5 EAL Threshold Value (1 or 2)

1.

An UNPLANNED event results in RCS temperature> 200°F, the Technical Specification cold shutdown temperature limit. OR

2.

Loss of all RCS temperature and RPV level indication for > 15 minutes.

Table C Cold Shutdown/Refueling System Malfunctions Category GENERAL EMERGENCY RCS Leakage/RCS CG4 (Pg 169) Draindown IC Loss of RPV Inventory Affecting Fuel Clad Integrity with Containment Challenged with Irradiated Fuel in the RPV. Modes: 4, 5 EAL Threshold Value (1 and 2 and 3)

1.

Loss of RPV inventory as indicated by unexplained Drywell or Reactor Building Sump or Drywell Equipment Drain Tank level increase, or Suppression Pool level increase, or other indications of loss of RPV inventory, AND

2.

RPV Level: A. <... (TAF) for > 30 minutes OR B. RPV level cannot be monitored concurrent with indication of core uncovery for > 30 minutes as evidenced by one of the following: Containment High Range Rad Monitor reading greater than or equal to 1 OR/hr, Erratic Source Range Monitor Indication Visual indication. AND

3.

Indication of Containment challenged as indicated by one or more of the following: Drywell Pressure > 53 psig and increasing OR Drywell Hydrogen or Suppression Chamber Hydrogen > 6% AND Drywell Oxygen or Suppression Chamber Oxygen > 5% OR Secondary Containment not established OR Any Reactor Building area exceeds Max Safe Radiation Levels per Table C-1 Page 3 of 34 SITE AREA EMERGENCY CS4 (Pg 173) IC Loss of RPV Inventory Affecting Core Decay Heat Removal Capability. Mode: 4 EAL Threshold Value (1 or 2)

1.

With secondary Containment NOT established: (a or b)

a. Loss of RPV inventory as indicated by RPV level

<- 135' OR

b. RPV level cannot be monitored for> 30 minutes with a loss of RPV inventory as indicated by:

Unexplained Drywell or Reactor Building Sump or Drywell Equipment Drain Tank level increase, or Unexplained Suppression Pool level increase, or Other unexplained indications of loss of RPV inventory OR 2. With secondary Containment established: (a or b) a. Loss of RPV inventory as indicated by RPV level-" (TAF) OR

b. RPV level cannot be monitored for> 30 minutes with a loss of RPV inventory as indicated by either:

Unexplained Drywell or Reactor Building Sump or Drywell Equipment Drain Tank level increase, or Unexplained Suppression Pool level increase, or Other unexplained indications of loss of RPV inventory Erratic Source Ran e Monitor Indication ALERT CA4 (Pg 175) IC Loss of RCS Inventory. Mode: 4 EAL Threshold Value (1 or 2)

1.

Loss of RCS inventory as indicated by RPV level< -129" OR

2.

Loss of RCS inventory as indicated by unexplained Drywell or Reactor Building Sump or Drywell Equipment Drain Tank level increase, or Suppression Pool level increase, or other indications of loss of RPV inventory, AND RCS level cannot be monitored for> 15 minutes. Attachment C EP-RM-004 Revision 1 Page 158 of 195 NOTICE OF UNUSUAL EVENT CU4 (Pg 177) IC RCS Leakage. Mode: 4 EAL Threshold Value (1 or 2)

1.

Unidentified or pressure boundary leakage > 1 0 gpm. OR

2.

Identified leakage > 25 gpm.

Table C Cold Shutdown/Refueling System Malfunctions Category GENERAL EMERGENCY SITE AREA EMERGENCY ALERT Fuel Clad Degradation Loss of Communications Inadvertent Criticality Page 5 of 34 Attachment C EP-RM-004 Revision 1 Page 160 of 195 NOTICE OF UNUSUAL EVENT CU6 (Pg 185) IC Fuel Clad Degradation. Modes: 4, 5 EAL Threshold Value (1 or 2)

1.

UNPLANNED increase in the following radiation monitor readings: Refuel Floor radiation Monitor> 750 mr/hr Refuel Platform ARM > 750 mr/hr Refuel Floor Continuous Air Monitor > 300 DAC OR

2.

Reactor coolant activity, determined by sample analysis ~4 1-1Ci/gm of 1-131 dose equivalent. CU7 (Pg 187) IC UNPLANNED Loss of all ON SITE or Offsite Communications Capabilities. Modes:. 4, 5 EAL Threshold Value (1 or 2)

1.

UNPLANNED loss of all ON SITE communications capability per Table C-2 affecting the ability to perform routine operations. OR

2.

UNPLANNED loss of all offsite communications capability per Table C-2. CUB (Pg 189) IC Inadvertent Criticality. Modes: 4, 5 EAL Threshold Value An UNPLANNED extended positive period observed on nuclear instrumentation. I I I ' I

Table C-1 Reactor Building Radiation Monitors ARM Number Max Safe Radiation RB Area High ARM Channel Levels Elevation (ft) Range Description Per E0-000-1 04 (R/HR) 818 49 Refuel Floor Area 10 749 52 RWCU Recirc PP 10 54 Fuel Pool PP Area 719 50 CRD North 10 51 CRD South 670 53 Remote Shutdown Room 10 645 48 HPCI PP & Turbine Room 10 57 RCIC PP & Turbine Room 55 RHR A C PP Room 56 RHR B D PP Room Table C-2 ON SITE/Offsite Communications Capability SYSTEM ON SITE OFF SITE UHF Radio X Commercial telephone systems X X Loss of dedicated conference lines to the offsite X agencies FTS-2001 (ENS) X Plant PA System X Plant cellular telephones X Telecopy Transmittal X Sound powered phones X Page 6 of 34 Attachment C EP-RM-004 Revision 1 Page 161 of 195

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NRLA NRLB Input to: 1.) Backup Setpoint Setdown Logic 2.) Selected Level Input to Backup 1 E Controller I.J_ RFPT.!WIIin Twblne L.wi B Trip logic Turbin. Lewll Trip Logic I Attachment A ON-145-001 Revision 35 REACTOR VESSEL LEVEL INPUTS Upset Level NRLB Total stnm Flow :o-14.0 t.lbmlhr (a~ s tM nm. Filter) I NRLC -- ~---t~T~ 'NRLBB' NRLB Biased_, NRLA (3): Unbla,ed NRLB l.evei ii'IP'A orlyby M-..iS.Iectlon +/-51nches

  • +/-10 Inches
  • State Alarm
  • State Alarm NRLBB NRLA
  • Input marked Deviant
  • r

-~Median Levell Validation J i?_ AUTO/MANUAL FWLC INPUT LEVEL SELECTION RFPTIMIIin r..wn. ~....,., a.,.._ TripL.ogk LIC-C32-1 R606C - ..@_ NRLC

(5) Upset leVel Biased

~ Uput Level Input onty by Mlinu* Selection -~ . f Selected Level Auto Bias Preference Order: 1)AVG LEVEL

2) NRLBB
3) NRLA
4) NRLC
5) UPSET
6) Default t,

~-- 'UPSETLB' M:"::"a~n~u~a..-:1 S::o-e~le~c~te::-:dl"":L-:e~ve~l Preference Order: 1)AVG LEVEL

2) NRLBB
3) NRLA
4) NRLC
5) NRLB
6) UPSETLB
7) UPSETL LI-C32-1 R606A LI-C32-1 R606B

_i_ XR-10602, Point# 6 TMS PICSY Point TRA142 Page 1 of 1 13" Normal Setpoint Set down Logic Selected Level Indication on XR-10602 Point# 5 LT ~. 35" DEFAULT TO 35" the Manually If ALL Level Input~;: Declared BAD Selected Levellnpu RoDingAver111ge I r--t -~~-J ....,.. ~ A & B Rx Recirc Pumps Limiter Inputs: 13"- # 1 Limiter A & B Rx Recirc Pumps Limiter Inputs: 30" - # 2 Limiter Selected Level Input to FWLC 1E or3E Control RXWATER HI-LO LEVEL Selected Level Selected Level Input for RWCU Input for FW Lo Letdown Load Valve Controller Controller HIC-G33-1 R606 LIC-C32-1 R602

3.1 REACTIVITY CONTROL SYSTEMS 3.1.4 Control Rod Scram Times PPL Rev. 2 Control Rod Scram Times 3.1.4 LCO 3.1.4

a.

No more than

  • OPERABLE control rods shall be "slow," in accordance with Table 3.1.4-1; and
b.

No more than I OPERABLE control rods that are "slow" shall occupy adjacent locations. APPLICABILITY: MODES 1 AND 2. ACTIONS CONDITION A. Requirements of the LCO not met. SUSQUEHANNA-UNIT 1 REQUIRED ACTION A.1 Be in MODE 3. TS/3.1-12 COMPLETION TIME 12 hours Amendment ya. 237

SURVEILLANCE REQUIREMENTS (continued) SR 3.1.4.2 SR 3.1.4.3 SR 3.1.4.4 SURVEILLANCE Verify, for a representative sample, each tested control rod scram time is within the limits of Table 3.1.4-1 with reactor steam dome pressure ~ 800 psig. Verify each affected control rod scram time is within the limits of Table 3.1.4-1 with any reactor steam dome pressure. Verify each affected control rod scram time is within the limits of Table 3.1.4-1 with reactor steam dome pressure~ 800 psig. SUSQUEHANNA-UNIT 1 TS /3.1-13 PPL Rev. 2 Control Rod Scram Times 3.1.4 FREQUENCY 200 days cumulative operation in MODE 1 Prior to declaring control rod OPERABLE after work on control rod or CRD System that could affect scram time Prior to exceeding 40% RTP after fuel movement within the affected core cell Prior to exceeding 40% RTP after work on control rod or CRD System that could affect scram time Amendment.:t-78-, ~. 249

Table 3.1.4-1 (page 1 of 1) Control Rod Scram Times PPL Rev. 2 Control Rod Scram Times 3.1.4

~()TE:S--------------------------------------------------------

1.

()PE:RABLE: control rods with scram times not within the limits of this Table are considered "slow."

2.

E:nter applicable Conditions and Required Actions of LC() 3.1.3, "Control Rod ()PE:RABILITY," for control rods with scram times> 7 seconds to notch position 05. These control rods are inoperable, in accordance with SR 3.1.3.4, and are not considered "slow." ~()TCH P()SITI()~ 45 39 25 05 SCRAM TIME:s(a)(b) (seconds) when RE:ACT()R STE:AM DOME: PRE:SSURE: 2 800 psig 0.52 0.86 1.91 3.44 (a) Maximum scram time from fully withdrawn position, based on de-energization of scram pilot valve solenoids at time zero. (b) Scram times as a function of reactor steam dome pressure, when <800 psig are within established limits. SUSQUE:HA~~A-U~IT 1 3.1-14 Amendment 178

PPL Rev. 0 Control Rod Scram Accumulators 3.1.5 3.1 REACTIVITY CONTROL SYSTEMS 3.1.5 Control Rod Scram Accumulators LCO 3.1.5 Each control rod scram accumulator shall be OPERABLE. APPLICABILITY: MODES 1 and 2. ACTIONS

NOTE------------------------------------------------------------

Separate Condition entry is allowed for each control rod scram accumulator. CONDITION REQUIRED ACTION COMPLETION TIME A One control rod scram A.1

NOTE---------------

accumulator inoperable Only applicable if the with reactor steam dome associated control rod scram pressure 2::

  • psig.

time was within the limits of Table 3.1.4-1 during the last scram time Surveillance. Declare the associated control 8 hours rod scram time "slow." OR A.2 Declare the associated control 8 hours rod inoperable. (continued) SUSQUEHANNA-UNIT 2 3.1-15 Amendment 151

ACTIONS CONDITION B. Two or more control rod scram accumulators inoperable with reactor steam dome pressure 0 ps1g. SUSQUE:HANNA-UNIT 2 PPL Rev. 0 Control Rod Scram Accumulators 3.1.5 RE:QUIRE:D ACTION 8.1 Restore charging water header pressure to 2:

  • psig.

NOTE:---------------

8.2.1 Only applicable if the associated control rod scram time was within the limits of Table 3.1.4-1 during the last scram time Surveillance. COMPLE:TION TIME: discovery of Condition B concurrent with charging water header pressure <.psig Declare the associated control 1 hour rod scram time "slow." 8.2.2 Declare the associated control 1 hour rod inoperable. (continued) 3.1-16 Amendment 151

CONDITION C. D. SUSQUEHANNA-UNIT 2 C. 1 AND C.2 PPL Rev. 0 Control Rod Scram Accumulators 3.1.5 REQUIRED ACTION COMPLETION TIME 3.1-17 Amendment 151

PPL Rev. 5 Primary Containment Isolation Instrumentation 3.3.6.1 3.3 INSTRUMENTATION 3.3.6.1 Primary Containment Isolation Instrumentation LCO 3.3.6.1 The primary containment isolation instrumentation for each Function in Table 3.3.6.1-1 shall be OPERABLE. APPLICABILITY: According to Table 3.3.6.1-1. ACTIONS

N 0 T E S-----------------------------------------------------

1. Penetration flow paths may be unisolated intermittently under administrative controls.
2. Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Place channel in trip. 12 hours for Functions channels inoperable. 2.a, 2.d, 6.b, ?.a, and ?.b AND 24 hours for Functions other than Functions 2.a, 2.d, 6.b, ?.a and ?.b B. One or more automatic B.1 Restore isolation 1 hour Functions with isolation capability. capability not maintained. (continued) SUSQUEHANNA-UNIT 1 TS I 3.3-52 Amendment 17;8, 213

ACTIONS (continued) CONDITION C. Required Action and C.1 associated Completion Time of Condition A or B not met. D. As required by Required D. 1 Action C.1 and referenced in Table 3.3.6.1-1. OR D.2.1 D.2.2 E. As required by Required E.1 Action C.1 and referenced in Table 3.3.6.1-1. F. As required by Required F.1 Action C.1 and referenced in Table 3.3.6.1-1. G. As required by Required G.1 Action C.1 and referenced in Table 3.3.6.1-1. SUSQUEHANNA-UNIT 1 PPL Rev. 5 Primary Containment Isolation Instrumentation 3.3.6.1 REQUIRED ACTION COMPLETION TIME Enter the Condition Immediately referenced in Table 3.3.6.1-1 for the channel. Isolate associated main 12 hours steam line (MSL). Be in MODE 3. 12 hours AND Be in MODE 4 36 hours Be in MODE 2. 6 hours Isolate the affected 1 hour penetration flow path(s). Isolate the affected 24 hours penetration flow path(s). (continued) 3.3-53 Amendment 178

ACTIONS (continued) CONDITION H. As required by Required H.1 Action C.1 and referenced in Table 3.3.6.1-1. AND H.2 OR Required Action and associated Completion Time for Condition F or G not met. I. As required by Required 1.1 Action C.1 and referenced in Table 3.3.6.1-1 OR 1.2 J. As required by Required J.1 Action C.1 and referenced in Table 3.3.6.1-1. OR J.2 SUSQUEHANNA-UNIT 1 PPL Rev. 5 Primary Containment Isolation Instrumentation 3.3.6.1 REQUIRED ACTION COMPLETION TIME Be in MODE 3. 12 hours Be in MODE 4. 36 hours Declare associated 1 hour standby liquid control subsystem (SLC) inoperable. Isolate the Reactor 1 hour Water Cleanup System. Initiate action to restore Immediately channel to OPERABLE status. Initiate action to isolate Immediately the Residual Heat Removal (RHR) Shutdown Cooling System. 3.3-54 Amendment 178

PPL Rev. 5 Primary Containment Isolation Instrumentation 3.3.6.1 SURVEILLANCE REQUIREMENTS

N()TES----------------------------------------------------

1.

Refer to Table 3.3.6.1-1 to determine which SRs apply for each Primary Containment Isolation Function.

2.

When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains isolation capability. SR 3.3.6.1.1 SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.4 SR 3.3.6.1.5 SURVEILLANCE Perform CHANNEL CHECK.

1.

A test of all required contacts does not have to be performed

2.

For Functions 2.e, 3.a, and 4.a, a test of all required relays does not have to be performed Perform CHANNEL FUNCTI()NAL TEST. Perform CHANNEL CALIBRATI()N. Perform CHANNEL CALIBRATI()N. Perform L()GIC SYSTEM FUNCTI()NAL TEST. SUSQUEHANNA-UNIT 1 3.3-55 FREQUENCY 12 hours 92 days 92 days 24 months 24 months Amendment 178

SR 3.3.6.1.6 PPL Rev. 5 Primary Containment Isolation Instrumentation 3.3.6.1

N()TE-----------------------------

1.

For Function 1.b. channel sensors are excluded.

2.

Response time testing of isolating relays is not required for Function 5.a. Verify the IS() LA TI()N SYSTEM RESP()NSE TIME is within limits. FREQUENCY 24 months on a STAGGERED TEST BASIS SUSQUEHANNA-UNIT 1 TS /3.3-56 Amendment 191

FUNCTION

1.

Main Steam Line Isolation

a.

Reactor Vessel Water Level - Low Low Low, Level 1

b.

Main Steam Line Pressure - Low

c.

Main Steam Line Flow-High

d.

Condenser Vacuum -Low

e.

Reactor Building Main Steam Tunnel Temperature-High

f.

Manual Initiation PPL Rev. 5 Primary Containment Isolation Instrumentation 3.3.6.1 Table 3.3.6.1-1 {page 1 of 6) Primary Containment Isolation Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE CONDITIONS SYSTEM ACTION C.1 REQUIREMENTS VALUE 1,2,3 2 D SR 3.3.6.1.1 ~ -136 inches SR 3.3.6.1.2 SR 3.3.6.1.4 SR 3.3.6.1.5 2 E SR 3.3.6.1.2 ~ 841 psig SR 3.3.6.1.3 SR 3.3.6.1.5 SR 3.3.6.1.6 1,2,3 2 per MSL D SR 3.3.6.1.1 $ 179 psid SR 3.3.6.1.2 SR 3.3.6.1.4 SR 3.3.6.1.5 1 2 D SR 3.3.6.1.2 ~ 8.8 inches 2(a)' 3(a) SR 3.3.6.1.3 Hg vacuum SR 3.3.6.1.5 1,2,3 2 D SR 3.3.6.1.2 $ 184°F SR 3.3.6.1.3 SR 3.3.6.1.5 1,2,3 G SR 3.3.6.1.5 NA (continued) (a) With any main turbine stop valve not closed. SUSQUEHANNA-UNIT 1 TS /3.3-57 Amendment 1}tf. 246

FUNCTION

2.

Primary Containment Isolation

a.

Reactor Vessel Water Level - Low, Level 3

b.

Reactor Vessel Water Level - Low Low, Level2

c.

Reactor Vessel Water Level - Low Low Low, Level1

d.

Drywell Pressure - High

e.

SGTS Exhaust Radiation - High

f.

Manual Initiation SUSQUEHANNA-UNIT 1 PPL Rev. 5 Primary Containment Isolation Instrumentation 3.3.6.1 Table 3.3.6.1-1 (page 2 of 6) Primary Containment Isolation Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE CONDITIONS SYSTEM ACTION C.1 REQUIREMENTS VALUE 1,2,3 2 H SR 3.3.6.1.1 <! 11.5 inches SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.5 1,2,3 2 H SR 3.3.6.1.1 <! -45 inches SR 3.3.6.1.2 SR 3.3.6.1.4 SR 3.3.6.1.5 1,2,3 2 H SR 3.3.6.1.1 <!-136 inches SR 3.3.6.1.2 SR 3.3.6.1.4 SR 3.3.6.1.5 1,2,3 2 H SR 3.3.6.1.2 ~ 1.88 psig SR 3.3.6.1.4 SR 3.3.6.1.5 1,2,3 H SR 3.3.6.1.1 ~ 31 mR!hr SR 3.3.6.1.2 SR 3.3.6.1.4 SR 3.3.6.1.5 1,2,3 G SR 3.3.6.1.5 NA (continued) 3.3-58 Amendment 178

FUNCTION

3.

High Pressure Coolant Injection (HPCI) System Isolation

a.

HPCI Steam Line tJ. Pressure - High

b.

HPCI Steam Supply Line Pressure - Low

c.

HPCI Turbine Exhaust Diaphragm Pressure-High

d.

Drywell Pressure - High

e.

HPCI Pipe Routing Area Temperature-High

f.

HPCI Equipment Room Temperature-High

g.

HPCI Emergency Area Cooler Temperature - High

h.

Manual Initiation SUSQUEHANNA-UNIT 1 PPL Rev. 5 Primary Containment Isolation Instrumentation 3.3.6.1 Table 3.3.6.1-1 (page 3 of6) Primary Containment Isolation Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE CONDITIONS SYSTEM ACTION C.1 REQUIREMENTS VALUE 1,2,3 F SR 3.3.6.1.2

!> 383 inches SR 3.3.6.1.3 H20 SR 3.3.6.1.5 1,2,3 2

F SR 3.3.6.1.2

e
90 psig SR 3.3.6.1.3 SR 3.3.6.1.5 1,2,3 2

F SR 3.3.6.1.2

!> 20 psig SR 3.3.6.1.3 SR 3.3.6.1.5 1,2,3 2

F SR 3.3.6.1.2

!>1.88 psig SR 3.3.6.1.3 SR 3.3.6.1.5 1,2,3 F

SR 3.3.6.1.2

!> 174°F SR 3.3.6.1.3 SR 3.3.6.1.5 1,2,3 F

SR 3.3.6.1.2

!> 174°F SR 3.3.6.1.3 SR 3.3.6.1.5 1,2,3 F

SR 3.3.6.1.2

!> 174°F SR 3.3.6.1.3 SR 3.3.6.1.5 1,2,3 G

SR 3.3.6.1.5 NA (continued) 3.3-59 Amendment 178

FUNCTION

4.

Reactor Core Isolation Cooling (RCIC) System Isolation

a.

RCIC Steam Line tJ. Pressure - High

b.

RCIC Steam Supply Line Pressure-Low

c.

RCIC Turbine Exhaust Diaphragm Pressure - High

d.

Drywell Pressure - High

e.

RCIC Pipe Routing Area Temperature-High

f.

RCIC Equipment Room Temperature-High

g.

RCIC Emergency Area Cooler Temperature-High

h.

Manual Initiation SUSQUEHANNA-UNIT 1 PPL Rev. 5 Primary Containment Isolation Instrumentation 3.3.6.1 Table 3.3.6.1-1 (page 4 of6) Primary Containment Isolation Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE CONDITIONS SYSTEM ACTION C.1 REQUIREMENTS VALUE 1,2,3 F SR 3.3.6.1.2 ~ 193inches SR 3.3.6.1.3 H20 SR 3.3.6.1.5 1,2,3 2 F SR 3.3.6.1.2

53 psig SR 3.3.6.1.3 SR 3.3.6.1.5 1,2,3 2

F SR 3.3.6.1.2 ~ 20 psig SR 3.3.6.1.3 SR 3.3.6.1.5 1,2,3 2 F SR 3.3.6.1.2 ~ 1.88 psig SR 3.3.6.1.3 SR 3.3.6.1.5 1,2,3 F SR 3.3.6.1.2 ~ 174°F SR 3.3.6.1.3 SR 3.3.6.1.5 1,2,3 F SR 3.3.6.1.2 ~ 174°F SR 3.3.6.1.3 SR 3.3.6.1.5 1,2,3 F SR 3.3.6.1.2 ~ 174°F SR 3.3.6.1.3 SR 3.3.6.1.5 1,2,3 G SR 3.3.6.1.5 NA (continued) 3.3-60 Amendment 178

FUNCTION

5.

Reactor Water Cleanup (RWCU) System Isolation

a.

RWCU Differential !!. Flow-High

b.

RWCU Penetration Area Temperature-High

c.

RWCU Pump Area Temperature-High

d.

RWCU Heat Exchanger Area Temperature-High

e.

SLC System Initiation

f.

Reactor Vessel Water Level -Low Low, Level2

g.

RWCU Flow - High

h.

Manual initiation PPL Rev. 5 Primary Containment Isolation Instrumentation 3.3.6.1 Table 3.3.6.1-1 (page 5 of 6) Primary Containment Isolation Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE CONDITIONS SYSTEM ACTION C.1 REQUIREMENTS VALUE 1,2,3 F SR 3.3.6.1.1 s 67 gpm SR 3.3.6.1.2 SR 3.3.6.1.4 SR 3.3.6.1.5 SR 3.3.6.1.6 1,2,3 F SR 3.3.6.1.2 s 137°F SR 3.3.6.1.3 SR 3.3.6.1.5 1,2,3 F SR 3.3.6.1.2 s 154°F SR 3.3.6.1.3 SR 3.3.6.1.5 1,2,3 F SR 3.3.6.1.2 s 154°F SR 3.3.6.1.3 SR 3.3.6.1.5 1,2,3 2(b) SR 3.3.6.1.5 NA 1,2,3 2 F SR 3.3.6.1.1 ~ -45 inches SR 3.3.6.1.2 SR 3.3.6.1.4 SR 3.3.6.1.5 1,2,3 F SR 3.3.6.1.1 s 472 gpm SR 3.3.6.1.2 SR 3.3.6.1.4 SR 3.3.6.1.5 1,2,3 G SR 3.3.6.1.5 NA (continued) (b) SLC System Initiation only inputs into one of the two trip systems. SUSQUEHANNA-UNIT 1 3.3-61 Amendment 17ft, 239

FUNCTION

6.

Shutdown Cooling System Isolation

a.

Reactor Steam Dome Pressure-High

b.

Reactor Vessel Water Level-Low, Level3

c.

Manual Initiation

7.

Traversing lncore Probe Isolation

a.

Reactor Vessel Water Level - Low, Level 3

b.

Drywell Pressure-High PPL Rev. 5 Primary Containment Isolation Instrumentation 3.3.6.1 Table 3.3.6.1-1 {page 6 of 6) Primary Containment Isolation Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE CONDITIONS SYSTEM ACTION C.1 REQUIREMENTS VALUE 1,2,3 F SR 3.3.6.1.2 ~ 108 psig SR 3.3.6.1.3 SR 3.3.6.1.5 3,4,5 2(c) J SR 3.3.6.1.1

.
11.5 inches SR 3.3.6.1.2 SR 3.3.6.1.3 SR 3.3.6.1.5 3,4,5 1 (c)

G SR 3.3.6.1.5 NA 1,2,3 2 G SR 3.3.6.1.1 SR 3.3.6.1.2

.
11.5 inches SR 3.3.6.1.3 SR 3.3.6.1.5 1,2,3 2

G SR 3.3.6.1.2 ~ 1.88 psig SR 3.3.6.1.4 SR 3.3.6.1.5 (c) Only one trip system required in MODES 4 and 5 when RHR Shutdown Cooling System integrity maintained. SUSQUEHANNA-UNIT 1 TS I 3.3-62 Amendment~. ~. 259

3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.10 RCS Pressure and Temperature (PIT) Limits PPL Rev. 2 RCS P/T Limits 3.4.1 0 LCO 3.4.10 RCS pressure, RCS temperature, RCS heatup and cooldown rates, and the recirculation pump starting temperature requirements shall be maintained within limits. APPLICABILITY: At all times. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

NOTE-----------

A.1 Restore parameter( s) 30 minutes Required Action A.2 to within limits. shall be completed if this Condition is AND entered. A.2 Determine RCS is 72 hours acceptable for Requirements of the continued operation. LCO not met in MODES 1, 2, and 3. B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time of Condition A not AND met. B.2 Be in MODE 4. 36 hours (continued) SUSQUEHANNA - l)NIT 1 3.4-24 Amendment 178

ACTIONS (continued) CONDITION C.

NOTE-----------

Required Action C.2 shall be completed if this Condition is entered. Requirements of the LCO not met in other than MODES 1, 2, and 3. SUSQUEHANNA-UNIT 1 C.1 AND C.2 REQUIRED ACTION Initiate action to restore parameter(s) to within limits. Determine RCS is acceptable for operation. 3.4-25 PPL Rev. 2 RCS P/T Limits 3.4.1 0 COMPLETION TIME Immediately Prior to entering MODE 2 or 3. Amendment 178

SURVEILLANCE REQUIREMENTS SR 3.4.10.1 SURVEILLANCE

N 0 T E -----------------------------

Only required to be performed during RCS heatup and cooldown operations and RCS inservice leak and hydrostatic testing. Verify: PPL Rev. 2 RCS PIT Limits 3.4.1 0 FREQUENCY

a.

RCS pressure and RCS temperature are to 30 minutes the right of the most limiting curve specified SR 3.4.10.2 in Figures 3.4.1 0-1 through 3.4.1 0-3; and

b.

NOTE-----------------------------

Only applicable when governed by Figure 3.4.1 0-2, Curve B, and Figure 3.4.1 0-3, Curve C. RCS heatup and cooldown rates are

1 00°F in any one hour period; and
c.

N 0 TE -----------------------------

Only applicable when governed by Figure 3.4.10-1, Curve A RCS heatup and cooldown rates are ::::; 20 OF in any one hour period. Verify RCS pressure and RCS temperature are to the right of the criticality limit (Curve C) specified in Figure 3.4.1 0-3. SUSQUEHANNA-UNIT 1 TS /3.4-26 Once within 15 minutes prior to control rod withdrawal for the purpose of achieving criticality (continued) Amendment 200

PPL Rev. 2 RCS PIT Limits 3.4.1 0 SURVEILLANCE REQUIREMENTS (continued) SR 3.4.10.3 SR 3.4.10.4 SR 3.4.10.5 SURVEILLANCE

NOTE----------------------------

Only required to be met in MODES 1 I 21 31 and 4 during recirculation pump start. FREQUENCY Verify the difference between the bottom head Once within 15 minutes coolant temperature and the reactor pressure vessel prior to each startup of a (RPV) coolant temperature is ~ 145°F. recirculation pump

NOTE----------------------------

Only required to be met in MODES 1 I 21 31 and 4 during recirculation pump start. Verify the difference between the reactor coolant temperature in the recirculation loop to be started and the RPV coolant temperature is ~ 50°F.

N 0 T E ---------------------------

Only required to be met in single loop operation when:

a.

THERMAL POWER ~ 27% RTP; or

b.

The operating recirculation loop flow ~ 211320 gpm. Verify the difference between the bottom head coolant temperature and the RPV coolant temperature is ~ 145°F. Once within 15 minutes prior to each startup of a recirculation pump Once within 15 minutes prior to an increase in THERMAL POWER or an increase in loop flow (continued) SUSQUEHANNA-UNIT 1 3.4-27 Amendment 9'8~ 246

SURVEILLANCE REQUIREMENTS (continued) SR 3.4.10.6 SR 3.4.10.7 SR 3.4.10.8 SURVEILLANCE

N 0 TE----------------------------

Only required to be met in single loop operation when the idle recirculation loop is not isolated from the RPV, and:

a.

THERMAL POWER ~ 27% RTP; or

b.

The operating recirculation loop flow ~ 21,320 gpm. Verify the difference between the reactor coolant temperature in the recirculation loop not in operation and the RPV coolant temperature is ~ 50°F.

N 0 T E ----------------------------

Only required to be performed when tensioning the reactor vessel head bolting studs. Verify reactor vessel flange and head flange temperatures are ~ 70°F.

NOTE----------------------------

Not required to be performed until 30 minutes after RCS temperature ~ 80°F in MODE 4. Verify reactor vessel flange and head flange temperatures are ~ 70°F. SUSQUEHANNA-UNIT 1 3.4-28 PPL Rev. 2 RCS PIT Limits 3.4.10 FREQUENCY Once within 15 minutes prior to an increase in THERMAL POWER or an increase in loop flow. 30 minutes 30 minutes (continued) Amendment JJS, 246

SR 3.4.10.9 SURVEILLANCE

NOTE----------------------------

Not required to be performed until 12 hours after RCS temperature ::::; 1 00°F in MODE 4. Verify reactor vessel flange and head flange temperatures are ~ 70°F. SUSQUEHANNA-UNIT 1 3.4-29 PPL Rev. 2 RCS P/T Limits 3.4.10 FREQUENCY 12 hours Amendment 178

r--... (J) en Q_ '--.-/ -o 0 Q) I Q_ 0 f-- Q) en en Q) > L 0 0 0 Q) 0::: c E

..:::J Q)

L

J en en Q)

L Q_ 1300 1200 1100 1000 900 800 700 600 500 400 300 200 100 0 60 II ~l I {i! IJI / J 'll v',l ~~-* ,, ~ I i J,I I I ~... i ,I I I ~~* I ~ I ..~ 80 100 120 --Upper Vessel -Beltline -*-Bottom Head 140 160 180 PPL Rev. 2 RCS PIT Limits 3.4.10 200 Minimum Reactor Vessel Metal Temperature (degrees F) FIGURE 3.4.1 0-1 System Hydrotest Limit with Fuel in Vessel for 35.7 EFPY (Curve A) SUSQUEHANNA-UNIT 1 TS /3.4-30 Amendment 200", 232

1300 1200 1100 h I I I // I I II I I I CJ) 'iii 1000 f J r 0.. u 0 Q) I 0.. 0 f-Q) (/) (/) Q) > 0 u 0 Q) cr: c E

.=i Q)
J

(/) (/) Q) Q_ 900 800 700 600 500 /) I I I I v I I I .I I I I f I ,~*4 I / I Beltline I 400 I --- Upper Vessel I Bottom Head I 300 ~ ~ 200 / ~, 100 0 60 80 100 120 140 160 180 200 Minimum Reactor Vessel Metal Temperature (degrees F) FIGURE 3.4.1 0-2 Non-Nuclear Heating Limit for 35.7 EFPY (Curve B) PPL Rev. 2 RCS PIT Limits 3.4.10 SUSQUEHANNA-UNIT 1 TS /3.4-30a Amendment 200, 232

Ol .iii Q_ -o 0 Q) I Q_ 0 1-Q) en en Q) > L 0 4J () 0 Q) 0::: c 4J .E

.J Q)

L

l en en Q)

L 0... 1300 1200 1100 1000 1 I I I 900 ~ 800 700 I I 600 500 400 300 v 200 100 0 _,/ v r 60 80 1 00 120 140 160 180 200 Minimum Reactor Vessel Metal Temperature (degrees F) FIGURE 3.4. 1 0-3 Nuclear (Core Critical) Limit for 35.7 EFPY (Curve C) PPL Rev. 2 RCS PIT Limits 3.4.1 0 SUSQUEHANNA-UNIT 1 TS I 3.4-30b Amendment 200, 232

3.6 CONTAINMENT SYSTEMS 3.6.1.3 Primary Containment Isolation Valves (PCIVs) LCO 3.6.1.3 Each PCIV shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3, PPL Rev. 3 PCIVs 3.6.1.3 When associated instrumentation is required to be OPERABLE per LCO 3.3.6.1, "Primary Containment Isolation Instrumentation." ACTIONS

NOTES--------------------------------------------------------

1.

Penetration flow paths may be unisolated intermittently under administrative controls.

2.

Separate Condition entry is allowed for each penetration flow path.

3.

Enter applicable Conditions and Required Actions for systems made inoperable by PCIVs.

4.

Enter applicable Conditions and Required Actions of LCO 3.6.1.1, "Primary Containment," when PCIV leakage results in exceeding overall containment leakage rate acceptance criteria in MODES 1, 2, and 3. CONDITION REQUIRED ACTION COMPLETION TIME A.

NOTE---------------

A.1 Isolate the affected 4 hours except for Only applicable to penetration flow path by main steam line penetration flow paths with use of at least one closed two PCIVs except for the and de-activated automatic AND HzOz Analyzer valve, closed manual penetrations. valve, blind flange, or 8 hours for main check valve with flow steam line through the valve secured. One or more penetration flow paths with one PCIV inoperable except for purge valve leakage not within limit. AND (continued) SUSQUEHANNA-UNIT 1 TS I 3.6-8 Amendment 195

ACTIONS CONDITION A (continued) A.2 SUSQUEHANNA-UNIT 1 REQUIRED ACTION

NOTE-------------

Isolation devices in high radiation areas may be verified by use of administrative means. Verify the affected penetration flow path is isolated. TS /3.6-9 PPL Rev. 3 PCIVs 3.6.1.3 COMPLETION TIME Once per 31 days for isolation devices outside primary containment Prior to entering MODE 2 or 3 from MODE 4, if primary containment was de-inerted while in MODE 4, if not performed within the previous 92 days, for isolation devices inside primary containment (continued) Amendment 195

ACTIONS (continued) CONDITION B.

NOTE-------------

8.1 Only applicable to penetration flow paths with two PCIVs except for the H20 2 Analyzer penetrations. One or more penetration flow paths with two PCIVs inoperable except for purge valve leakage not within limit. C.

NOTE-------------

C.1 Only applicable to penetration flow paths with only one PCIV. One or more penetration flow paths with one PC IV inoperable. AND C.2 SUSQUEHANNA-UNIT 1 REQUIRED ACTION Isolate the affected penetration flow path by use of at least one closed and de-activated automatic valve, closed manual valve, or blind flange. Isolate the affected penetration flow path by use of at least one closed and de-activated automatic valve, closed manual valve, or blind flange.

NOTE-------------

Isolation devices in high radiation areas may be verified by use of administrative means. Verify the affected penetration flow path is isolated. TS I 3.6-10 PPL Rev. 3 PCIVs 3.6.1.3 COMPLETION TIME 1 hour 72 hours except for excess flow check valves (EFCVs) AND 12 hours for EFCVs Once per 31 days (continued) Amendment 195

ACTIONS (continued) CONDITION D.

NOTE--------------

Only applicable to the H20 2 Analyzer penetrations. One or more H202 Analyzer penetrations with one or two PCIVs inoperable. E. Secondary containment bypass leakage rate not within limit. F. One or more penetration flow paths with one or more containment purge valves not within purge valve leakage limit. G. Required Action and associated Completion Time of Condition A, B, C, D, E, or F not met in MODE 1, 2, or 3. H. Required Action and associated Completion Time of Condition A, B, C, D, E, or F not met for PCIV(s) required to be OPERABLE during MODE 4, 5 or Operations with the potential for draining the reactor vessel (OPDRVs). SUSQUEHANNA-UNIT 1 REQUIRED ACTION D.1 Isolate the affected penetration flow path by the use of at least one closed and de-activated automatic valve, closed manual valve, or blind flange. AND D.2 Verify the affected penetration flow path is isolated. E.1 Restore leakage rate to within limit. F.1 Restore the valve leakage to within valve leakage limit. G.1 Be in MODE 3. AND G.2 Be in MODE 4. H.1 Initiate action to suspend OPDRVs. OR H.2 Initiate action to restore valve(s) to OPERABLE status. TS I 3.6-11 PPL Rev. 3 PCIVs 3.6.1.3 COMPLETION TIME 72 hours Once per 31 days 4 hours 24 hours 12 hours 36 hours Immediately Immediately Amendment 195

SURVEILLANCE REQUIREMENTS SR 3.6.1.3.1 SR 3.6.1.3.2 SURVEILLANCE

NOTES----------------------------

1.

Only required to be met in MODES 1, 2, and3.

2.

Not required to be met when the 18 and 24 inch primary containment purge valves are open for inerting, de-inerting, pressure control, ALARA or air quality considerations for personnel entry, or Surveillances that require the valves to be open. Verify each 18 and 24 inch primary containment purge valve is closed.

NOTES----------------------------

1.

Valves and blind flanges in high radiation areas may be verified by use of administrative means.

2.

Not required to be met for PCIVs that are open under administrative controls. Verify each primary containment isolation manual valve and blind flange that is located outside primary containment and not locked, sealed, or otherwise secured and is required to be closed during accident conditions is closed. SUSQUEHANNA-UNIT 1 3.6-12 PPL Rev. 3 PCIVs 3.6.1.3 FREQUENCY 31 days 31 days (continued) Amendment 178

SURVEILLANCE REQUIREMENTS (continued) SR 3.6.1.3.3 SR 3.6.1.3.4 SR 3.6.1.3.5 SURVEILLANCE

NC>TES-----------------------------

1.

Valves and blind flanges in high radiation areas may be verified by use of administrative means.

2.

Not required to be met for PCIVs that are open under administrative controls. Verify each primary containment manual isolation valve and blind flange that is located inside primary containment and not locked, sealed, or otherwise secured and is required to be closed during accident conditions is closed. Verify continuity for each of the traversing in core probe (TIP) shear isolation valve explosive charge. Verify the isolation time of each power operated and each automatic PCIV, except for MSIVs, is within limits. SUSQUEHANNA-UNIT 1 TS /3.6-13 PPL Rev. 3 PCIVs 3.6.1.3 FREQUENCY Prior to entering MC>DE 2 or 3 from MC>DE 4 if primary containment was de-inerted while in MC>DE 4, if not performed within the previous 92 days 31 days In accordance with the lnservice Testing Program (continued) Amendment 178

SURVEILLANCE REQUIREMENTS (continued) SR 3.6.1.3.6 SR 3.6.1.3.7 SR 3.6.1.3.8 SR 3.6.1.3.9 SURVEILLANCE

N()TE------------------------------

()nly required to be met in M()DES 1, 2 and 3. Perform leakage rate testing for each primary containment purge valve with resilient seals. Verify the isolation time of each MSIV is 2': 3 seconds and ~ 5 seconds. Verify each automatic PCIV actuates to the isolation position on an actual or simulated isolation signal. Verify a representative sample of reactor instrumentation line EFCVs actuate to check flow on a simulated instrument line break. SUSQUEHANNA-UNIT 1 TS /3.6-14 PPL Rev. 3 PCIVs 3.6.1.3 FREQUENCY 24 months In accordance with the I nservice Testing Program 24 months 24 months (continued) Amendment..:H13, 223

SURVEILLANCE REQUIREMENTS (continued) SR 3.6.1.3.10 SR 3.6.1.3.11 SR 3.6.1.3.12 SURVEILLANCE Remove and test the explosive squib from each shear isolation valve of the TIP System.

N()TES----------------------------

()nly required to be met in M()DES 1, 2, and 3. Verify the combined leakage rate for all secondary containment bypass leakage paths is ::;; 15 scfh when pressurized to 2::: Pa.

N()TES----------------------------

()nly required to be met in M()DES 1, 2, and 3. Verify leakage rate through each MSIV is

100 scfh and
;300 scfh for the combined leakage including the leakage from the MS Line Drains, when the MSIVs are tested at :::::: 24.3 psig or Pa and the MS Line Drains are tested at Pa.

PPL Rev. 3 PCIVs 3.6.1.3 FREQUENCY 24 months on a STAGGERED TEST BASIS In accordance with the Primary Containment Leakage Rate Testing Program. In accordance with the Primary Containment Leakage Rate Testing Program. (continued) SUSQUEHANNA - UNIT 1 TS I 3.6-15 Amendment -i-00, 24-e, 251

SURVEILLANCE REQUIREMENTS (continued) SR 3.6.1.3.13 SURVEILLANCE

N()TE------------------------------

()nly required to be met in M()DES 1, 2, and 3. Verify combined leakage rate through hydrostatically tested lines that penetrate the primary containment is within limits. SUSQUEHANNA-UNIT 1 TS I 3.6-16 PPL Rev. 3 PCIVs 3.6.1.3 FREQUENCY In accordance with the Primary Containment Leakage Rate Testing Program. Amendment 178

PPL Rev. 1 3.4 Reactor Coolant System Reactor Coolant System Chemistry 3.4.1 3.4.1 Reactor Coolant System Chemistry TRO 3.4.1 The chemistry of the reactor coolant system shall be maintained within the limits specified in Table 3.4.1-1 and the conductivity recorder shall be OPERABLE. APPLICABILITY: At all times ACTIONS

NOTE -----------------------------------------------------------

1. The provisions of TRO 3.0.4 are not applicable.

CONDITION A Conductivity not within the limits specified in Table 3.4.1-1. B. Conductivity or Chlorides not within the limits specified in Table 3.4.1-1 in MODE 1 but with conductivity less than 10 J..Lmho/cm at 25°C and chloride concentration less than 0.5 ppm. C. pH not within limits of Table 3.4.1-1 in MODE 1. D. Required Actions and associated Completion Times of Conditions B or C not met. SUSQUEHANNA-UNIT 1 REQUIRED ACTION A.1 Perform TRS 3.4.1.1 and TRS 3.4.1.3. AND A.2 Perform TRS 3.4.1.4 B.1 Restore parameter within limits C.1 Restore pH within limits. D.1 Be in MODE 2 COMPLETION TIME Once per 8 hours Once per 24 hours 72 hours

336 hours/year cumulative time exceeding the limit 72 hours 6 hours (continued)

TRM /3.4-1 EFFECTIVE DATE 03/31/2006

PPL Rev. 1 ACTIONS (continued) CONDITION

NOTE----------------

Required Actions shall be completed if this Condition is entered. E. Conductivity greater than or equal to 10 J..tmho/cm at 25°C or chloride concentration greater than or equal to than 0.5 ppm. F. Reactor Coolant Chemistry not within limits specified in Table 3.4.1-1 for MODE 2or 3 G. Required Actions and Completion Time of Condition F not met. H.

NOTE-------------

Only applicable when in a MODE other than MODES 1, 2, or 3. Reactor Coolant Chemistry not within limits specified in Table 3.4.1-1. SUSQUEHANNA-UNIT 1 E.1 AND E.2 F.1 G.1 AND G.2 H.1 Reactor Coolant System Chemistry 3.4.1 REQUIRED ACTION COMPLETION TIME Be in MODE 3. 12 hours

N 0 TE ----------

Cooldown should be performed as rapidly Be in MODE 4. as possible not to exceed the cooldown rate 36 hours Restore parameter to within 48 hours limits. Be in MODE 3 12 hours Be in MODE 4 36 hours Restore parameter to 72 hours within limits (continued) TRM /3.4-2 EFFECTIVE DATE 10/23/1998

PPL Rev. 1 ACTIONS (continued) CONDITION

NOTE ---------------

1.1 Required Actions shall be completed if this Condition is entered. I. Chloride concentration greater than 0.5 ppm for > 24 hours J. Conductivity recorder J.1 inoperable. SUSQUEHANNA-UNIT 1 Reactor Coolant System Chemistry 3.4.1 REQUIRED ACTION COMPLETION TIME Perform Engineering Prior to proceeding to Evaluation of structural MODE3 integrity Obtain in-line conductivity

N 0 T E ----------

measurement or grab sample Applicable in MODES 1, 2, or

3.

Once per 4 hours AND

N 0 TE ----------

Applicable in other than MODES 1, 2, or 3. Once per 24 hours TRM I 3.4-3 EFFECTIVE DATE 10/23/1998

PPL Rev. 1 Reactor Coolant System Chemistry 3.4.1 TECHNICAL REQUIREMENT SURVEILLANCE TRS 3.4.1.1 TRS 3.4.1.2 TRS 3.4.1.3 TRS 3.4.1.4 SURVEILLANCE Analyze a sample of reactor coolant for chlorides Analyze a sample of reactor coolant for conductivity Analyze a sample of reactor coolant for pH FREQUENCY 72 hours 72 hours

N()TE---------------

()nly required to be performed if reactor conductivity is greater than 1.0 flmho/cm at 25°C 72 hours Perform CHANNEL CHECK of the continuous 7 days conductivity monitor SUSQUEHANNA-UNIT 1 TRM /3.4-4 EFFECTIVE DATE 10/23/1998

PPL Rev. 1 TABLE 3.4.1-1 Reactor Coolant System Chemistry 3.4.1 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS CONDUCTIVITY CHLORIDE MODE 1-1mho/cm @ 25°C CONCENTRATION ppm pH 1

.:; 1.0
.:; 0.2 5.6 ::.:; pH ::.:; 8.6 2 and 3
.:; 2.0
.:; 0.1 5.6 ::.:; pH ::.:; 8.6 At all times other
.:; 10.0
.:; 0.5 5.3 ::.:; pH ::.:; 8.6 than MODE 1, 2, or 3 SUSQUEHANNA-UNIT 1 TRM /3.4-5 EFFECTIVE DATE 10/23/1998}}
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