Text

Event Classification Guide,Technical Basis, for Hope Creek Generating Station & Salem
	ML18102A794
Person / Time
Site:	Salem, Hope Creek
Issue date:	01/21/1997
From:	; Public Service Enterprise Group
To:	;
Shared Package
ML18102A792	List: ML18102A791; ML18102A793; ML18102A794;
References
	PROC-970121-01, PROC-970121-1, NUDOCS 9701310344
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January 21, 1997 SECTION TITLE REV# PAGES DATE T.0.C. Table of Contents/Signature Page 00 2 01/21/97 Introduction and Usage 00 3 01/21/97 II Glossary of Acronyms & Abbreviations 00 5 01/21/97 1.0 Fuel Clad Challenge 00 9 01/21/97 2.0 RCS Challenge 00 8 01/21/97 3.0 Fission Product Barriers (Table)

3. 1 Fuel Clad Barrier 00 13 01/21/97 3 .2 RCS Barrier 00 18 01/21/97 3.3 Containment Barrier 00 17 01/21/97 4.0 EC Discretion 00 8 01/21/97 5.0 Failure to SCRAM 00 10 01/21/97 6.0 Radiological Releases/Occurrences 6.1 Gaseous Effiuent Release 00 44 01/21/97 6.2 Liquid Effiuent Release 00 4 01/21/97 6.3 In - Plant Radiation Occurrences 00 6 01/21/97 6.4 Irradiated Fuel Event 00 8 01/21/97 7.0 Electrical Power 7 .1 Loss of AC Power Capabilities 00 18 01/21/97 7.2 Loss of DC Power Capabilities 00 5 01121/97 8.0 System Malfunctions 8.1 Loss of Heat Removal Capability 00 8 01/21/97 8.2 Loss of Overhead Annunciators 00 8 01/21/97 8.3 Loss of Communications Capability 00 4 01/21/97 8.4 Control Room Evacuation 00 4 01121/97 8.5 Technical Specifications 00 2 01/21/97 9.0 Hazards - Internal/External 9.1 Security Threats 00 8 01121/97 9.2 Fire 00 6 01/21/97 9.3 Explosion 00 4 01121197 9.4 Toxic/Flammable Gases 00 11 01121197 9.5 Seismic Event 00 4 01121/97

9. 6 High Winds 00 7 01/21/97

9. 7 Flooding 00 4 01/21/97 9.8 Turbine FailureNehicle Crash/ 00 6 01121197 Missile Impact 9.9 River Level 00 4 01/21/97 HCGS Rev. 00

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HC EAL Technical Basis T.O.C. Pg. 2 of 2 HOPE CREEK ECG TECHNICAL BASIS TABLE OF CONTENTS/SIGNATURE PAGE January 21, 1997 SECTION TITLE REV# PAGES DATE 10.0 Reserved for future use 11.0 Reportable Action Levels (RALs)* 11.1 Technical Specifications 00 7 01/21/97 11.2 Design Basis/ Unanalyzed Condition 00 6 01/21/97 11.3 Engineered Safety Features (ESF) 00 4 01/21/97 11.4 Personnel Safety/Overexposure 00 8 01/21/97 11.5 Environmental 00 3 01/21/97 11.6 After-the-Fact 00 1 01/21/97 11.7 Security/Emergency Response 00 5 01121/97 Capabilities 11.8 Public Interest 00 3 01121/97 11.9 Accidental Criticality/ 00 9 01121/97 Special Nuclear Material I Rad Material Shipments - Releases 11.10 Voluntary Notifications 00 2 01/21/97 HCGS Rev. 00

I HCGS EAL/RAL Technical Basis Section i

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v_,.......,- ~ INTRODUCTION & USAGE Section i NOTE This document may be referenced for clarification of an EAL or RAL prior to declaration within the Assessment Time period, as appropriate to ensure the correct classification. CAUTION DO NOT delay classification of an Emergency by referring to this document unless significant doubt exists about the intent or meaning of an EAL. LO PURPOSE OF THE ECG TECHNICAL BASIS (TB) I. I Classification

Reference:

Provides a reference document which assists the Emergency Coordinator or SNSS in classifying emergency or non-emergency events by presenting;

detailed basis information on each EAL and RAL
analysis of the effect of the event on fission product barriers
escalation criteria guidance 1.2 Training and Communication

Reference:

Provides the basis and source of the action levels within the ECG to include in training and to reference when communicating with outside agencies the methodology and rationale of a classification. 2.0 TECHNICAL BASIS (TB) STRUCTURE

2. I Basis Sections Format:

Items 2. 1.1 thru 2.1.4 are taken directly from the Event Classification Guide. They are repeated in this document for the sake of convenience in cross referencing the information in items 2.1.5 thru 2.1. l 0.

2. I. I INITIATING CONDITION (IC): A generic nuclear power plant condition or event where either the potential exists for a radiological emergency or HCGS REV. 00

HCGS EALIRAL Technical Basis Section i Pg 2 of3 non-emergency reportable event OR such an emergency or non-emergency reportable event has occurred. 2.1.2 EAL NUMBER (EAL#): Each Emergency Action Level (EAL) has been assigned a unique alpha numeric identifier. Each digit of the EAL# has a specific meaning that is not important to the users, but is important to the personnel who develop and maintain the ECGs. The digit and EAL # are defined in the ECG Introduction Section i. 2.1.3 EMERGENCY ACTION LEVEL (EAL) OR REPORTABLE ACTION LEVEL (RAL): A predetermined, site-specific, observable threshold used to define when the generic initiating condition has been met, placing the plant in a given emergency class or non-emergency report.

2.1.4 OPCON

Refers to the Operational Condition at Hope Creek during which a particular IC/EAL is applicable. The OPCON that the plant was in when the event started, prior to any protection system or operator actions, should be utilized when classifying events. 2.1. 5 BASIS: Provides an explanation of terms and expressions used in the action levels for better understanding of their meaning and, when appropriate, their derivation. Words contained in an EAL or RAL that are bold face are either threshold values associated with that action level or are words that are defined in the basis for that specific EAL/RAL. 2.1.6 BARRIER ANALYSIS: Provides a short statement about any of the three fission product barriers that may be affect'ed by this event. 2.1. 7 ESCALATION CRITERIA: Provides a brief description of any additional conditions or events that, should they occur, would require escalation of the emergency level. Other EAL #s may be included for reference.

2.1.8 DISCUSSION

Provides additional background information on the action level and concerns for plant safety. Basis calculations are included for some specific EAL thresholds where appropriate to aid in communicating the derivation and assumptions that were used in development.

2.1.9 DEVIATION

Provides a brief explanation of any differences between the Salem EAL and the NUMARC based EAL examples given in NESP-007. 2.1.10

REFERENCES:

Provides a short list of the pertinent documents that are the basis for information included in this technical basis document. HCGS REV. 00

HCGS EAL/RAL Technical Basis Section i Pg 3 of3 3.0 TECHNICAL BASIS (TB) USAGE Event classification should always be made with direct reference to the ECG. If any numbering inconsistency or error should be discovered between the ECG and TB, the EAL #s from the ECG are to be used. 3 .1 Classification: USE the Hope Creek ECG to first of all; 3.1.1 ASSESS the event and/or plant conditions and DETERMINE which ECG._ section(s) is most appropriate. 3.1.2 REFER to Section EAL/RAL Flowchart diagram(s), review and identify the Initiating Conditions that are related to the event/condition that has occurred or is ongoing. 3 .1.3 REVIEW the associated EALs or RALs as compared to the event and SELECT the highest appropriate emergency or reportable action level lfthere is any doubt with regard to assessment of a particular EAL or RAL, the ECG Technical Basis should be reviewed.

3. 1.4 REFER to this document after initial notifications are begun to review escalation criteria and the technical basis in order to gain a broader understanding of the reasons for taking action at this time.

3.2 Training and Communication

Reference:

3.2.1 This document is used in training Emergency Coordinators and those tasked with analyzing events and advising the EC on classifications. 3.2.2 Offsite Agencies that require further explanation of the EAL or RAL in effect may reference a copy of this document or Offsite Reference Manual _(layperson' s guide) if available. HCGS REV. 00

ECG Section ii Pg 1 of5 HOPE CREEK EVENT CLASSIFICATION GUIDE Glossary of Acronyms & Abbreviations Section ii ---*:~ ... -~

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I AAAG Accident Assessment Advisory Group (Del~ware) . / i AC Alternating Current 1 (o~ ADS Automatic Depressuri:zation System i ALARA As Low As Reasonably Achievable .... ~---,**~--=""..............,....,.,.__~-.: APRM Average Power Range Monitor ARI Alternate Rod Insertion ARM Area Radiation Monitor ASAP As Soon As Possible ASM Administrative Support Manager AS Administrative Supervisor ATWS Anticipated Transient Without Scram BKGD Background BKR Breaker (electrical circuit) BNE Bureau of Nuclear Engineering (NJDEPE) CACS Containment Atmosphere Control System CAS Central Alarm Station CCPM Corrected Counts per Minute CEDE Committed Effective Dose Equivalent CDE Committed Dose Equivalent CFR Code of Federal Regulations CIS Containment Isolation System CNTMT Containment (Barrier) CP Control Point CPM Counts Per Minute CR Control Room CREF Control Room Emergency Filter System CRIDS Control Room Integrated Display System CRD Control Rod Drive css Core Spray System DC Direct Current DAPA Drywell Atmosphere Post Accident (Radiation monitor) DDE Deep Dose Equivalent DEi Dose Equivalent Iodine DEMA Delaware Emergency Management Agency DEP Department of Environmental Protection (NJ) HCGS Rev. 00

ECG Section ii Pg 2 of5 DID Direct Inward Dial (phone system) DLD Drywell Leak Detection DOE Department of Energy DOT Department of Transportation DPCC/DCR - Discharge Prevention, Containment, & Countermeasures/ Discharge Cleanup & Removal Plan DPM Disintegrations per Minute DRCF Dose Rate Conversion Factor EACS ESF Equipment Area Cooling System EAL Emergency Action Level EAS Emergency Alert System (Broadcast) EC Emergency Coordinator ECCS Emergency Core Cooling Systems ECG Emergency Classification Guide EDG Emergency Diesel Generator EDO Emergency Duty Officer EMRAD Emergency Radio (NJ) ENC Emergency News Center ENS Emergency Notification System (NRC) EOC Emergency Operations Center (NJ & DE) EOF Emergency Operations Facility

EOP Emergency Operating Procedure EPA Emergency Preparedness Advisor EPA Environmental Protection Agency EPC Emergency Preparedness Coordinator EPIP Emergency Plan Implementing Procedure EPZ Emergency Planning Zone EQPT Equipment ERDS Emergency Response Data System ERM Emergency Response Manager ERO Emergency Response Organization ESF Engineered Safety Feature ESSX Electronic Switch System Exchange (Centrex) FC Fuel Clad (Barrier) FFD Fitness For Duty FRVS Filtration, Recirculation, and Ventilation System FTS Federal Telecommunications System (NRC) GE General Emergency HCLL Heat Capacity Level Limit HCGS Hope Creek Generating Station HCGS Rev. 00 L

ECG Section ii Pg 3 of5 HCTL Heat Capacity Temperature Limit HEPA High Efficiency Particulate Absorbers HPCI High Pressure Coolant Injection HTV Hardened Torus Vent HVAC Heating, Ventilation & Air Conditioning HWCI Hydrogen Water Chemical Injection HX Heat Exchanger JAW In Accordance With IC Initiating Condition ICMF Initial Contact Message Form

1DLH Immediately Dangerous to Life and Health IRM Intermediate Range Monitor I/S In Service KI Potassium Iodide KV Kilovolt LAC Lower Alloways Creek LCO Limiting Condition for Operation LDE Lens Dose Equivalent LEL Lower Explosive Limit LLD Lowest Level Detectable LOCA Loss of Coolant Accident LOP Loss of Offsite Power LPCI Low Pressure Coolant Injection LPZ Low Population Zone MCR Main Control Room MDA Minimum Detectable Amount MEA Minimum Exclusion Area MEES Major Equipment & Electrical Status (Form)

MET Meteorological M.0.U. Memorandum of Understanding MRO Medical Review Officer MSIV Main Steam Isolation Valve MS IV SS Main Steam Isolation Valve Sealing System MSL Main Steam Line NA WAS National Attack Warning Alert System NCO Nuclear Control Operator NDAB Nuclear Department Administration Building (TB2) NEO Nuclear Equipment Operator NETS Nuclear Emergency Telecommunications System HCGS Rev. 00

ECG Section ii Pg 4 of5 NFE Nuclear Fuels Engineer NFPB Normal Full Power Background NJSP New Jersey State Police NOAA National Oceanographic and Atmospheric Administration NPV North Plant Vent NRC Nuclear Regulatory Commission NSS Nuclear Shift Supervisor NS SSS Nuclear Steam Supply Shutoff System NSTA Nuclear Shift Technical Advisor NUMARC Nuclear Management and Resources Council NWS National Weather Service OBE Operating Basis Earthquake OCA Owner Controlled Area ODCM Offsite Dose Calculation Manual OEM Office of Emergency Management (NJ) OHA Overhead Annunciator OPCON Operating Condition OSB Operational Status Board (Form) osc Operations Support Center PAG Protective Action Guideline PAR Protective Action Recommendation PASS Post Accident Sample System PC Primary Containment (Barrier) PCIG Primary Containment Instrument Gas System PCIS Primary Containment Isolation System PSIG Pounds Square Inch Gauge RAD Radiation RAL Reportable Action Level RC Reactor Coolant RCA Radiologically Controlled Area RCAM Repair and Corrective Action Mission RCIC Reactor Core Isolation Cooling RCS Reactor Coolant System (Barrier) RHR Residual Heat Removal (Containment Heat Removal) RM Recovery Manager RMO Recovery Management Organization RMS Radiation Monitoring System RPS Radiation Protection Supervisor RPS Reactor Protection System RPV Reactor Pressure Vessel RRCS Redundant Reactivity Control System HCGS Rev.00

ECG Section ii Pg 5 of5 RSM Radiological Support Manager RWCU Reactor Water Cleanup (System) SACS Safety Auxiliaries Cooling System SAE Site Area Emergency SAM Severe Accident Management SAS Secondary Alarm Station (Security) SBO Station Blackout SCBA Self Contained Breathing_ Apparatus SCP Security Contingency Procedure SDE Shallow Dose Equivalent SDM Shutdown Margin SLC Standby Liquid Control SJAE Steam Jet Air Ejector SNM Special Nuclear Material SNSS Senior Nuclear Shift Supervisor sos Systems Operations Supervisor (Security) SPDS Safety Parameter Display System SPV South Plant Vent SRM Source Range Monitor SRPT Shift Radiation Protection Technician SRV Safety Relief Valve SSCL Station Status Checklist SSE Safe Shutdown Earthquake ssws Station Service Water System SSNM Strategic Special Nuclear Material TAF Top of Active Fuel TDR Technical Document Room TEDE Total Effective Dose Equivalent TIP Traversing Incore Probe TLV Threshold Limit Value TIS Technical Specifications TSC Technical Support Center TSS Technical Support Supervisor TSTL Technical Support Team Leader TSTM Technical Support Team Member UE Unusual Event UFSAR Updated Final Safety Analysis Report UHS Ultimate Heat Sink USCG United States Coast Guard VDC Volts Direct Current WB Whole Body HCGS Rev. 00

HCGS EALIRALTechnical Basis 1.0 Fuel Clad Challenge 1.1 RCS Activity UNUSUAL EVENT-1.1.1.a l IC Fuel Clad Degradation --~~--------.......) EAL Reactor Coolant Sample Activity > 4 µCi/gm Dose Equivalent I-131 OPERATIONAL CONDITION - 1, 2, 3, 4, 5 BASIS A Reactor Coolant sample analysis with specific activity in excess of the Technical Specification limit of 4 µCi/gm Dose Equivalent Iodine-131 (DEl-131) is indicative of a degradation of the fuel clad, and is a precursor of more serious problems. This activity level is chosen instead of the 0.2 µCi/gm DEl-131 Technical Specification limit, under which operation is allowed to continue for up to 48 hours to accommodate short duration Iodine spikes following changes in thermal power. This EAL threshold does not use the term "Valid", since Reactor Coolant Sample Activity of greater than 4 uCi/gm DEi- 131 can only occur as the result of fuel clad degradation and not as the result of a resin I chemical intrusion transient or HWCI System malfunction. Unusual Event declaration is warranted only when actual fuel clad degradation has occurred. Barrier Analysis This event does not reach the threshold for the loss of the Fuel Clad Barrier, but does affect that barrier. ESCALATION CRITERIA Emergency Classification will escalate to an Alert or higher when a sample analysis of Reactor Coolant activity exceeds 300 µCi/gm DEI-131 per ECG Section 3 .0, Fission Product Barrier Table. EAL - I. I. I.a Rev. 00 Page 1 of 2

HCGS EALIRALTechnical Basis DISCUSSION The Technical Specification limit on Reactor Coolant activity ensures that the 2 hour thyroid and whole body doses resulting from a Main Steam Line failure outside the containment during steady state operation will not exceed a small fraction of the IOCFRIOO limits. This limit accommodates Iodine Spiking, which frequently occurs following shutdowns, startups, rapid power changes and coolant clepressurization. Iodine spikes are characterized by a rapid increase in Reactor Coolant Iodine concentration by as much as three orders of magnitude followed by a return to prespike concentrations. This spiking is a temporary excursion and is not caused by a sudden fuel failure. The Technical Specification limit of> 100/E µCi/gm is excluded from this EAL because this limit does not include Iodine Activity. DEVIATION NUMARC EAL SU 4.2 suggests that the Operating Mode Applicability for this EAL is ALL. When the Reactor is defueled, the source term needed to achieve an RCS Activity of 4 uCi/gm Dose Equivalent 1-131 is not available. Hence, this EAL is applicable in Operational Conditions 1,2,3,4 and 5. REFERENCES NUMARC NESP-007, SU4.2 Technical Specification LCO 3.4.5 HC.OP-AB.ZZ-OlOO(Q), High Reactor Coolant Activity HC.OP-AB.ZZ-0203(Q), Main Steam Line High Radiation 10 CFRlOO EAL - 1.1.1.a Rev. 00 Page 2 of 2

HCGS EALIRALTechnical Basis 1.0 Fuel Clad Challenge 1.1 RCS Activity UNUSUAL EVENT-1.1.1.b IC Fuel Clad Degradation EAL Valid Offgas Pretreatment Radiation Monitor (9RX621I9RX622) High Alarm Condition ( 2: 2.2E+04 mRem/hr) OPERATIONAL CONDITION - 1, 2, 3, 4 BASIS A Valid Offgas Pretreatment Radiation Monitor High alarm is indicative of a degradation of the fuel clad, and is a precursor of a more serious problems. The alarm is set at 2.2E+04 mR/hr, which ensures that the alarm will actuate prior to exceeding the Technical Specification Offgas System Noble Gas Effluent Limit of 3.3E5 µCi/sec. Valid is defined as the Offgas Pretreatment Radiation Monitor High Alarm actuating specifically due to fuel clad degradation, thus precluding unwarranted Unusual Event declaration as the result of a resin I chemical intrusion transient, or HWCI System malfunction. Unusual Event declaration is warranted only when actual fuel clad damage has occurred. Barrier Analysis This event does not reach the threshold for the loss of the Fuel Clad Barrier, but does affect that barrier. ESCALATION CRITERIA Emergency Classification will escalate to an Alert or higher when a sample an<!lysis of Reactor Coolant activity exceeds 300 µCi/gm DEi- I 3 I per ECG Section 3 .0, Fission Product Barrier Table. EAL - 1.1.1.b Rev. 00 Page 1 of 2

HCGS EAL/RALTechnical Basis DISCUSSION The Offgas Pretreatment Radiation Monitors (9RX621I9R.X622) monitor gamma radiation levels attributable to the non-condensable fission product gases produced in the reactor and transported with steam through the turbine to the condenser. This instrument takes a sample from the sample tap between the fourth and fifth holdup pipe of the Offgas system. Restricting the gross radioactivity from the Main Condenser provides reasonable assurance that the Total Body exposure to an individual at the exclusion area boundary will not exceed a small fraction of the limits of I 0 CFR Part 100 in the event this effluent is inadvertently discharged directly to the environment without treatment. Operating Experience at HCGS has demonstrated that Reactor coolant activity changes for reasons other than fuel clad degradation can result in temporarily increasing Main Steam Line Radiation Monitors and Offgas Pretreatment Radiation Monitors. Such events (e.g. - resin intrusion) do not require classification under this EAL. DEVIATION NUMARC EAL SU 4.1 suggests that the Operating Mode Applicability for this EAL is ALL. In Operational Condition 5 and Defueled, the MSIV s will be closed, thus rendering the Offgas Pretreatment Radiation Monitors unavailable for detection of increased RCS Activity. Hence, this EAL is applicable in Operational Conditions 1, 2, 3 and 4. REFERENCES NUMARC NESP-007, SU4. l Technical Specifications; Table 3.3.7.1 (5); LCO 3.11.2.7 HC.OP-AB.ZZ-OlOO(Q), High Reactor Coolant Activity HC.OP-AB.ZZ-0203(Q), Main Steam Line High Radiation HC.RP-AR.SP-OOOl(Q), Radiation Monitoring System Alarm Response OE-6144, Resin Intrusion 10 CFRlOO EAL - 1.1.1.b Rev. 00 Page 2 of 2

HCGS EAL/RALTechnical Basis 1.0 Fuel Clad Challenge 1.1 RCS Activity UNUSUAL EVENT-1.1.1.c IC Fuel Clad Deg. adation EAL Valid Main Steam Line Radiation Monitor High High Alarm Condition

   ~3 times Normal Full Power Background)

OPERATIONAL CONDITION - I, 2, 3, 4 BASIS A Valid Main Steam Line Radiation Monitor High High alarm (::: 3 times Normal Full Power Background) is indicative of degradation of the fuel clad and may be a precursor of more serious problems. Valid is defined as the Main Steam Line Radiation Monitor High High Alarm actuating specifically due to fuel clad degradation, thus precluding unwarranted Unusual Event declaration as the result of a resin I chemical intrusion transient, or HWCI System malfunction. Unusual Event declaration is warranted only when actual fuel clad degradation has occurred. Reaching the High High Alarm on ANY of the 4 Main Steam Line Radiation Monitor channels, as determined by receipt of ANY one of the following, due to fuel clad degradation, warrants Unusual Event declaration. Overhead Annunciator C6-B2, MN STM LINE RAD HI HI OR INOP CRIDS Point D2121, MN STM LINE HI HI RAD I INOP - W CRIDS Point D2122, MN STM LINE HI HI RAD I INOP - X CRIDS Point D2 I 23, MN STM LINE HI HI RAD I INOP - Y CRIDS Point D2 l 24, MN STM LINE HI HI RAD I INOP - Z Barrier Analysis This event does not reach the threshold for the loss of the Fuel Clad Barrier, but does affect that barrier. EAL - 1.1.1.c Rev.00 Page 1 of 3

HCGS EALIRALTechnical Basis ESCALATION CRITERIA Emergency Classification will escalate to an Alert or higher when a sample analysis of Reactor Coolant activity exceeds 300 µCi/gm DEI-131 per ECG Section 3.0, Fission Product Barrier Table. DISCUSSION The Main Steam Line Radiation Monitor Channels (9RX509, 9RX5 l 0, 9RX5 l l, 9RX512) monitor gamma radiation levels at the Main Steam Lines. A High High alarm is indicative of a release of gap activity to the coolant, but may not be indication of a major failure of the fuel clad. A Valid Main Steam Line Radiation High High alarm condition requires a manual Reactor Scram and Main Steam Isolation Valve closure to reduce and isolate the potential source of the radioactivity release. The terminology used for the 3 times Normal Full Power Background threshold differs between the Main Control Room Overhead Annunciators and the Radiation Monitoring System (RM-11 ). As a result, specific monitor channels are not included in the EAL. Overhead Annunciators use the terminology of "High High" for this threshold, where the RM-11 uses the terminology of "High" for the same threshold. For the purpose of this EAL, the High High setpoint terminology used by the Overhead Annunciators is used, though the same indications are available on the following RM-11 Channels: Main Steam Line "Channel" A (Grid 114; 9RX509) Main Steam Line "Channel" B (Grid 1/4; 9RX5 l 0) Main Steam Line "Channel" C (Grid 1/4; 9RX51 I) Main Steam Line "Channel" D (Grid 1/4; 9RX512) In addition, the Main Steam Line Radiation Monitor Numac Drawers can be used to trend changes in Main Steam Line Radiation Levels. A rapid power reduction from full power may cause the Main Steam line Radiation Monitors to momentarily increase to 1.5 times normal full background readings. This is due to the response time of the HWCI Hydrogen Flow Controller and the transport time from the Hydrogen Injection point (Secondary Condensate Pumps). Operating Experience at HCGS has demonstrated that Reactor coolant activity changes for reasons other than fuel clad degradation can result in temporarily increasing Main Steam Line Radiation Monitors and Offgas Pretreatment Radiation Monitors. Such events (e.g. - resin intrusion) do not require classification under this EAL. EAL - 1.1.1.c Rev. 00 Page 2 of 3

HCGS EALIRALTechnical Basis DEVIATION NUMARC EAL SU 4.1 suggests that the Operating Mode Applicability for this EAL is ALL. In Operational Condition 5 and Defueled, the MSIV s will be closed, thus rendering the Main Steam Line Rad Monitors unavailable for detection of increased RCS Activity. Hence, this EAL is applicable in Operational Conditions I, 2, 3 and 4. REFERENCES NUMARC NESP-007, SU4. l HC.RP-AR.SP-OOOl(Q), Radiation Monitoring System Alarm Response HC.OP-AR.ZZ-001 l(Q), Annunciator Response Procedures, Window C6-B2 HC.OP-AB.ZZ-0203(Q), Main Steam Line High Radiation HCGS Technical Specifications 3/4.3, Instrumentation Technical Specifications, LCO 3.11.2. 7 OE-6144, Resin Intrusion 10CFRIOO EAL - 1.1.1.c Rev.00 Page 3 of 3

HCGS EALIRALTechnical Basis 1.0 Fuel Clad Challenge 1.1 RCS Activity ALERT - 1.1.2 IC Fuel Clad Degradation EAL ANY one of the following:

Reactor Coolant Sample Activity > 4 µCi/gm Dose Equivalent I-131
Valid Offgas Pretreatment Radiation Monitor (9RX621I9RX622) High Alarm Condition

( ~ 2.2E+04 mRem/hr)

Valid Main Steam Line Radiation Monitor High High Alarm ConditionQ:3 times Normal Full Power Background)

ANY SR V is determined to be Stuck Open OPERATIONAL CONDITION - 1,2, 3 BASIS Indication ofFuel Clad Degradation coincident with ANY SRV determined to be Stuck Open is indicative of a Loss of the RCS Barrier, as fission products are being transported directly to the Suppression Pool, thus compromising the integrity of the RCS Barrier. Hence, Alert declaration is warranted. In the event c.n SRV is Stuck Open with NO indications of Fuel Clad degradation, an emergency declaration is NOT warranted, since an open SR V is within the analyzed design envelope of the plant and does not, by itself, represent a degradation in the level of plant safety. An SRV is considered to be Stuck Open when the SRV can not be reclosed by operator action within 2 minutes of ANY spurious, automatic or manual actuation. A Stuck Open SRV SHOULD NOT be considered as an Unisolable RCS Leak> 50 GPM, as the consequences of a Stuck Open SRV discharging to the Suppression Pool are different than an Unisolable RCS Leak exceeding 50 GPM that is discharging into the Drywell Air Space. EAL- 1.1.2 Rev. 00 Page I of2

HCGS EAL/RALTechnical Basis Barrier Analysis RCS Barrier has been lost. ESCALATION CRITERIA Emergency Classification will escalate based upon the Potential Loss or Loss of additional Fission Prnduct Barriers per EAL Section 3.0 DISCUSSION A Stuck Open SRV by itself requires a 1 Hour Report if a Unit Shutdown (Manual Reactor Scram) is initiated to comply with Technical Specification or a 4 Hour Report if the SRV is reclosed within the Technical Specification limits, due to the ESF actuation. A Stuck Open SRV discharging Reactor Coolant to the Suppression Pool does not represent the same challenge to the RCS and Primary Containment as an Unisolable RCS Leak discharging into the Drywell. The consequences of a Stuck Open SRV do not represent a significant precursor to further plant degradation, as plant design (Pressure Suppression ability of the Torus) and the Abnormal Operating Procedure for a Stuck Open SRV (directing a Manual Reactor Scram within 2 minutes ifthe SRV can not be closed), minimize the consequences of the event. In contrast, an Unisolable RCS Leak (refer to 3.2.2.a Basis) represents a situation where there is concern for "break propagation", which could lead to a significantly larger uncontrolled loss of RCS inventory. Hence, a Stuck Open SRV must be coincident with Fuel Clad Degradation for the RCS Barrier to be considered lost. DEVIATION None REFERENCES NUMARC Questions and Answers, June 1993, "Fission Product Barrier Question #7" HC.OP-AB.ZZ-0121(Q), Failed Open Safety/Relief Valve EAL - 1.1.2 Rev.00 Page 2 of2

I I HCGS EALIRALTechnical Basis 2.0 RCS Challenge 2.1 RCS Leakage l I I UNUSUAL EVENT- 2.1.1.a / 2.1.1.b ! IC RCS Leakage EAL EITHER one of the following:

Reactor Coolant System Pressure Boundary Leakage > 10 gpm (Using 10 minute average)
Reactor Coolant System Unidentified Leakage> IO gpm (Using I 0 minute average)

OPERATIONAL CONDITION - 1, 2, 3 BASIS RCS Pressure Boundary and Unidentified Leakage exceeding 10 gpm is indicative of possible degradation of the RCS and may be a precursor of a more serious condition. RCS Operational Leakage addressed by these 2 EALs is specifically RCS leakage into the Drywell. Leakage into the Drywell that is confirmed to not be RCS Leakage, i.e. a leaking Drywell Cooling Coil, does not warrant classification under this EAL. These types of RCS Operational Leakage, exceeding their respective EAL thresholds, should be classified as an Unusual Event, regardless of whether or not the leak has been isolated, since the EAL thresholds exceed the Technical Specification (T/S) limits. Classification should be based on the I 0 minute average and not an instantaneous value, to assure accurate event classification. The value of 10 gpm for RCS Pressure Boundary and Unidentified Leakage was set higher than the TIS limits of 0 and 5 gpm respectively, to allow time to implement corrective actions (including plant shutdown) prior to exceeding the threshold. Only operating conditions in which there is fuel in the reactor coolant system and the system is pressurized are specified. EAL - 2.1.1.a / 2.1.1.b Rev. 00 Page 1 of 3

HCGS EALIRALTechnical Basis Barrier Analysis This event does not reach the threshold for the loss of the RCS Barrier, but does affect that barrier. ESCALATION CRITERIA Emergency Classification will escalate to an Alert when either UHisolable RCS Leak Rate exceeds 50 GPM or Drywell Pressure exceeds 1.68 psig per EAL Section 3.2.2 DISCUSSION Allowable leakage rates from the RCS are based on predicted and experimentally observed behavior ofcracks in pipes. Utilizing the leak before break methodology, it is anticipated that there will be indication(s) of minor RCS boundary leakage prior to a fault escalating to a major leak or a system rupture. Detection of low levels of leakage* while pressurized allows for implementation of mitigative actions and permits monitoring for catastrophic failure or rupture precursors. The limit for Unidentified and Pressure Boundary Leakage is set to a lower value, than Identified Leakage due to concern over "break propagation" resulting from an Unidentified or Pressure Boundary Leak (Small Break), that could potentially lead to a significantly larger loss of inventory. Identified Leakage occurs when there is degradation or failure of a mechanical joint. Pipe "break propagation" is thus not an issue. Drywell Leak Detection (OLD) Instrumentation available via the Radiation Monitoring System (RM-11) to determine RCS Leakage into the Drywell includes:

(9AX313) Drywell Equipment Drain Sump (OLD EQPT) Monitor
(9AX314) Drywell Floor Drain Sump (DLD FLR) Monitor
(9AX3 l 7) Lower Drywell Air Condensate Coolers (OLD CCM LOW) Monitor
(9AX3 l 8) Upper Drywell Air Condensate Coolers (DLD CCM UP) Monitor
(9AX3 l 9) Drywell Sumps (DLD SMS) Monitor
(9AXJ20) Drywell Air Condensate Coolers Summation (OLD CCM SUM) Monitor Redundant Instrumentation for OLD is available on Panel I O-C-604 located in the back of the Main Control Room.

TIS required actions based on this leak rate may require a plant shutdown and subsequent depressurization, unless the source of the leak can be located, identified, and/or stopped. DEVIATION None EAL - 2.1.1.a / 2.1.1.b Rev. 00 Page 2 of 3

HCGS EAL/RALTechnical Basis REFERENCES NUMARC NESP-007, SUS NUMARC Questions and Answers, June 1993, "General Question #12" NUMARC Questions and Answers, June 1993, "Fission Product Barrier Question #11" HC.OP-SO.SM-OOOI(Q), Isolation Systems Operation HC.OP-AB.ZZ-0116 (Q), Containment Isolation and Recovery From An Isolation HC.OP-AB.ZZ-0201 (Q), Drywell High Pressure/Loss ofDrywell Cooling HC.OP-EO.ZZ-0101 (Q)-FC, Reactor Pressure Vessel (RPV) Control HC.OP-EO.ZZ-0102 (Q)-FC, Primary Containment Control HC.OP-GP.ZZ-0005 (Q), Drywell Leakage Source Detection HCGS Technical Specifications, LCO 3.4.3.2 EAL - 2.1.1.a I 2.1.1.b Rev. 00 Page 3 of 3

HCGS EAL/RALTechnical Basis 2.0 RCS Challenge 2.1 RCS Leakage UNUSUAL EVENT- 2.1.1.c IC RCS Leakage EAL Reactor Coolant System Identified Leakage > 25 gpm averaged over any 24 hour period OPERATIONAL CONDITION - 1, 2, 3 BASIS RCS Identified Leakage exceeding 25 gpm is indicative of possible degradation of the RCS and may be a precursor of a more serious condition. RCS Operational Leakage addressed by this EAL is specifically RCS leakage into the Drywell. Leakage into the Drywell that is confirmed to not be RCS Leakage, i.e. a leaking Drywell Cooling Coil, does not warrant classification under this EAL. Identified Leakage should ONLY be classified as an Unusual Event, when the leak rate exceeds 25 gpm when averaged over any 24 hour period, regardless of whether or not the leak has been isolated. The 24 hour average is included as part of the EAL threshold to provide consistency with the Technical Specification (T/S) limit for Identified Leakage. Only operating conditions in which there is fuel in the reactor coolant system and the system is pressurized are specified. Barrier Analysis This event does not reach the threshold for the loss of the RCS Barrier, but does affect that barrier. ESCALATION CRITERIA Emergency Classification will escalate to an Alert when either Unisolable RCS Leak Rate exceeds 50 GPM or Drywell Pressure exceeds 1.68 psig per EAL Section 3.2.2 EAL - 2.1.1.c Rev. 00 Pagel of 3

HCGS EAL/RALTechnical Basis DISCUSSION Allowable leakage rates from the RCS are based on predicted and experimentally observed behavior of cracks in pipes. Utilizing the leak before break methodology, it is anticipated that there will be indication(s) of minor RCS boundary leakage prior to a fault escalating to a major leak or a system rupture. Detection of low levels of leakage while pressurized allows for implementation of mitigative actions and permits monitoring for catastrophic failure or rupture precursors. The limit for Unidentified and Pressure Boundary Leakage is set to a lower value, then Identified Leakage due to concern over "break propagation" resulting from an Unidentified or Pressure Boundary Leak (Small Break), that could potentially lead to a significantly larger loss of inventory. Identified Leakage occurs when there is degradation or failure of a mechanical joint. Pipe "break propagation" is thus not an issue. Drywell Leak Detection (DLD) Instrumentation available via the Radiation Monitoring System (RM-11) to determine RCS Leakage into the Drywell includes:

(9AX313) Drywell Equipment Drain Sump (DLD EQPT) Monitor
(9AX314) Drywell Floor Drain Sump (DLD FLR) Monitor
(9AX3 l 7) Lower Drywell Air Condensate Coolers (DLD CCM LOW) Monitor
(9AX318) Upper Drywell Air Condensate Coolers (DLD CCM UP) Monitor
(9AX319) Drywell Sumps (DLD SMS) Monitor
(9AX320) Drywell Air Condensate Coolers Summation (DLD CCM SUM) Monitor Redundant Instrumentation for DLD is available on Panel 10-C-604 located in the back of the Main Control Room.

TIS required actions based on this leak rate may require a plant shutdown and subsequent depressurization, unless the source of the leak can be located, identified, and/or stopped. DEVIATION NUMARC EAL SUS suggests that exceeding an RCS Identified Leakage limit of25 gpm warrants the declaration of an Unusual Event because it may be a precursor to a more serious condition. The Hope Creek Technical Specification limit for RCS Identified Leakage is 25 GPM averaged over any 24 hour period. The plant is within Technical Specification as long as this limit is not exceeded and hence an Unusual Event is not warranted until the limit is exceeded. This philosophy is consistent with that contained in NUMARC EAL SU2, which only requires declaration of an Unusual Event when the plant is outside the Technical Specification. RCS Pressure Boundary and Unidentified Leakage that exceed the NUMARC EAL threshold will be classified as an Unusual Event, as this leakage exceeds the Technical Specification limit. EAL - 2.1. i .c Rev.00 Page 2 of 3

HCGS EAL/RALTechnical Basis In addition, NVMARC EAL SUS appears to apply specifically to those plants that do not allow for averaging of RCS Identified Leakage over a 24 hour period. Furthermore, NUMARC Questions and Answers Document, June 1993, "General Question #12", addresses those cases where the Technical Specification LCO has been exceeded and the required Action section has been entered (i.e. 4 Hours to identify and reduce the leakage below the limit). The EAL threshold for RCS Identified Leakage does not consider this time for Unusual Event declaration. The Q&A also states that the EAL for RCS Identified Leakage has been significantly raised from 10 to 2S gpm at some plants. Since the Hope Creek Technical Specification limit is already set at 2S gpm averaged over any 24 hour period, the EAL should not be more limiting than the Technical Specifications. REFERENCES NUMARC NESP-007, SUS NUMARC Questions and Answers, June 1993, "General Question #12" NUMARC Questions and Answers, June 1993, "Fission Product Barrier Question # 11" HC.OP-SO.SM-000 l (Q), Isolation Systems Operation HC.OP-AB.ZZ-0116 (Q), Containment Isolation and Recovery From An Isolation HC.OP-AB.ZZ-0201 (Q), Drywell High Pressure/Loss ofDrywell Cooling HC.OP-EO.ZZ-0101 (Q)-FC, Reactor Pressure Vessel (RPV) Control HC.OP-EO.ZZ-0102 (Q)-FC, Primary Containment Control HC.OP-GP.ZZ-OOOS (Q), Drywell Leakage Source Detection HCGS Technical Specifications, LCO 3.4.3.2 EAL - 2.1.1.c Rev. 00 Page 3 of 3

HCGS EALIRALTechnical Basis 2.0 RCS Challenge 2.1 RCS Leakage UNUSUAL EVENT- 2.1.1.d IC RCS Leakage EAL Successful Isolation of a Reactor Recirc Pump Dual Seal Failure within 10 minutes of recognition OPERATIONAL CONDITION - I, 2, 3 BASIS Successful Isolation of a Reactor Recirc Pump Dual Seal Failure within IO minutes of recognition is classified as an Unusual Event, due to the significance of the event. Even though the consequences of a successfully isolated Recirc Pump Dual Seal failure are minor, with no possibility for "break propagation", an Unusual Event is warranted due to the multiple failures of mechanical joints that allowed the discharge of a significant quantity of Reactor Coolant (>50 GPM) directly into the Drywell Air Space. Successful is defined as indication of ALL of the following within 10 minutes of recognition of the Recirc Pump Dual Seal failure.

Recirc Pump Suction and Discharge Valves have closed
RWCU Suction Valve from the Recirc Loop has closed
Recirc Pump Seal Purge Water Valves have closed
Drywell Pressure and Temperature have begun to decrease
RCS Leakage has begun to decrease IO minutes was determined to be a reasonable amount of time to isolate the pump and monitor for the effectiveness of the actions.

Only operating conditions in which there is fuel in the reactor coolant system and the system is pressurized are specified. EAL - 2.1.1.d Rev. 00 Page 1 of 2

HCGS EALIRALTechnical Basis Barrier Analysis This event does not reach the threshold for the loss of the RCS Barrier, but does affect that barrier. ESCALATION CRITERIA Emergency Classificatio-n will escalate to an Alert iften minutes elapse prior to successful isolation or Drywell Pressure exceeds 1.68 psig per EAL Section 3.2.2 DISCUSSION Prompt recognition of a Recirc Pump Dual Seal failure by the operating crew will allow for implementation of actions to isolate the leakage source in accordance with Abnormal Operating Procedures. The design of the Recirc Pump Seal limits the magnitude of the identified leakage for this event to 60 gpm due to the presence of a breakdown bushing. As a result, RCS inventory will not be significantly effected. The ability to monitor the leak rate is limited to 50 gpm, the upper limit of the Drywell Leak Detection Instrumentation. Drywell Pressure is not expected to reach the High Drywell Pressure Scram setpoint for this event, provided that the isolation was successfully completed within 10 minutes. DEVIATION None REFERENCES NUMARC NESP-007, SU5 NUMARC Questions and Answers, June 1993, "General Question #12" NUMARC Questions and Answers, June 1993, "Fission Product Barrier Question # 11" HC.OP-AB.ZZ-0112 (Q), Recirculation Pump Trip HC.OP-AB.ZZ-020 I (Q), Drywell High Pressure/Loss of Drywell Cooling HC.OP-EO.ZZ-0101 (Q)-FC, Reactor Pressure Vessel (RPV) Control HC.OP-EO.ZZ-0102 (Q)-FC, Primary Containment Control HC.OP-GP.ZZ-0005 (Q), Drywell Leakage Source Detection HCGS Technical Specifications, LCO 3.4.3.2 EAL - 2.1.1.d Rev. 00 Page 2 of 2

HCGS EAL/RALTechnical Basis 3.0 Fission Product Barriers 3.1 Fuel Clad Barrier

                                                                                      -~-,,_ ,,..;."<~,
                                                                     ~ .,...-~~~:~*:\     #*           ~

3.1.1 REACTOR WATER LEVEL ........... 3.1.1.a IC Potential Loss of Fuel Clad Barrier= 3 POINTS EAL Reactor Water Level REACHES - 161" (Top of Active Fuel), EXCLUDING intentional lowering of Reactor Water Level during an ATWS OPERATIONAL CONDITION - 1, 2, 3 BASIS Reactor Water Level reaching -161" (Top of Active Fuel - T AF), excluding intentional lowering of Reactor Water Level during an ATWS, results in an inability to maintain adequate core cooling by core submergence, causing a Potential Loss of the Fuel Clad Barrier. Without core submergence, the integrity of the fuel clad barrier is in jeopardy. Appropriate classification under this EAL is based on reaching Reactor Water Level of

-161" (instead of being able to restore and maintain above -161 ") due to the potentially severe consequences of a loss of core submergence. Reactor Water Level reaching this threshold results from either a LOCA exceeding available makeup capacity or a Total Loss of High Pressure injection capability.

In addition, during an Anticipated Transient Without Scram (ATWS), it is possible that operator actions will be taken to intentionally lower Reactor Water Level to between -161" and -190", for Reactor Power Control purposes. For this event, classification must be made in accordance with EAL Section 5.0 Barrier Analysis Fuel Clad Barrier has been potentially lost ESCALATION CRITERIA Emergency Classification will escalate based upon the Potential Loss or Loss of additional Fission Product Barriers per EAL Section 3.0. EAL - 3.1.1.a Rev.00 Page 1 of 3

HCGS EALIRALTechnical Basis DISCUSSION Core Submergence is the preferred method of maintaining adequate core cooling. When Reactor Water Level decreases to below T AF, the ability to effectively remove decay heat is being challenged, and as such the Fuel Clad fission product barrier can no longer be considered intact. While the Emergency Operating Procedures provide contingencies to establish adequate core cooling when Reactor Water Level drops below T AF (Steam Cooling with or without injection), these actions are designed to be an alternative method of providing adequate core coolin~ while actions are taken to reestablish core submergence. Sustained partial or total core uncovery can result in fuel clad damage and a significant release of fission products to the Reactor coolant. Sustained core uncovery can also result in a breach of the Reactor Vessel due to core melt material interaction with the RPV. A Loss of Core Submergence will occur when the rate of inventory loss is greater than the rate of inventory makeup from High Pressure injection sources. This condition can occur as the result of the following events/sequences (excluding intentional lowering of Reactor Water Level during an ATWS). A LOCA will cause Reactor Water Level to reach the Top of Active Fuel when the LOCA is the result of a large break (momentary core uncovery is expected to occur under this condition) or when the LOCA is due to a small or intermediate break in combination with an inability of High Pressure injection sources to keep up with the leakrate. A Loss of High Pressure injection sources without the presence of a LOCA will also result in Reactor Water Level decreasing to T AF, due to continued Reactor Steam Flow without makeup. Either of these events/sequences results in a challenge to the Fuel Clad Barrier when Reactor Water Level reaches T AF due to core uncovery, hence classification at this threshold is appropriate. However, for* both these sequences, Low Pressure ECCS are designed to inject to the Reactor as Reactor Pressure decreases below the shutoff head of the pumps. Reactor Depressurization will occur either due to the LOCA or Manual initiation of Emergency Depressurization when Reactor Water Level reaches -161 ", provided injection systems are available. This will allow for restoration* of Reactor Water Level and re-establishment of Core Submergence. Failure of these systems to restore and maintain Reactor Water Level above -200" will require escalation. DEVIATION None EAL - 3.1.1.a Rev. 00 Page 2 of 3

HCGS EALIRALTechnical Basis REFERENCES NUMARC NESP-007, FC2 HC.OP-EO.ZZ-0100 (Q)-FC, Reactor Scram HC.OP-EO.ZZ-0101 (Q)-FC, Reactor Pressure Vessel (RPV) Control HC.OP-EO.ZZ-0201 (Q)-FC, Alternate Level Control HC.OP-EO.ZZ-0207 (Q)-FC, Level/Power Control BWR Owner's Group Emerger.cy Procedure Guidelines, Rev. 4 EAL - 3.1.1.a Rev. 00 P;1ge 3 of 3

HCGS EAL/RALTechnical Basis 3.0 Fission Product Barriers 3.1 Fuel Clad Barrier 3.1.1 REACTOR WATER LEVEL 3.1.1.b IC Loss of Fuel Clad Barrier= 4 POINTS EAL Reactor Water Level CANNOT BE RESTORED AND MAINTAINED above -200" (Minimum Zero Injection RPV Water Level) OPERATIONAL CONDITION - 1, 2, 3 BASIS Inability to restore and maintain Reactor Water Level above - 200" (Minimum Zero Injection RPV Water Level), results in a loss of adequate core cooling by all mechanisms, causing a Loss of the Fuel Clad Barrier. Without adequate core cooling, the integrity of the fuel clad barrier can no longer be assured. Appropriate classification under this EAL is based on the failure of injection systems to restore and maintain Reactor Water Level above -200", following a condition that causes level to decrease below the threshold. For example, a large break LOCA is expected to cause Reactor Water Level to momentarily decrease below -200", due to the response time of Low Pressure ECCS. As these systems initiate and commence injection to the Reactor, water level will begin to increase and should be able to be maintained above -200". In this case, classification under this EAL is not appropriate as plant systems have performed their intended design function and will eventually restore adequate core cooling by core submergence. However, in the event that Low Pressure ECCS and alternate injection system, as defined in the EOPs are in a degraded condition (i.e., Station Blackout, ECCS Suction Strainer plugging, etc.) and Reactor Water Level can not be restored and maintained above -200", then classification under this EAL should occur due to the potential for release of energy to the containment from imminent fuel failure. Barrier Analysis Fuel Clad Barrier has been lost. EAL - 3.1.1.b Rev. 00 Page 1 of 3

HCGS EAL/RALTechnical Basis ESCALATION CRITERIA Emergency Classification will escalate based upon the Potential Loss or Loss of additional Fission Product Barriers per EAL Section 3.0. DISCUSSION Core submergence is the preferred method for maintaining adequate core cooling. The failure to reestablish Reactor Water Level above -161 ",the Top of Active Fuel (TAF), for an extended period ohime could lead to a significant of fuel damage. With Reactor Water Level below TAF, but above the Minimum Zero Injection RPV Water Level (-200"), adequate core cooling occurs due to the cooling effects of steam generated in the covered portion of the core flowing through the uncovered portion (Steam Cooling). The Minimum Zero Injection RPV Water Level is defined in the Emergency Operating Procedures. This method of cooling precludes any fuel clad temperature in the uncovered portion of the core from exceeding 1800°F. As Reactor Water Level drops below -200" with no injection available, this method of cooling becomes inadequate. Prolonged lack of cooling may result in severe overheating of the fuel clad, additional release of energy from accelerated clad oxidation, and eventual fuel melting. For events starting from full power operation, the failure to promptly reflood could result in some fuel melting. Even under these conditions vessel failure and containment failure with resultant release to the public would not be expected for some tim~. Reactor Water Level remaining below T AF for an extended amount of time represents an early indicator that significant core damage is in progress while providing sufficient time to initiate public protective actions. Ample time should be allowed for Low Pressure ECCS and alternate injection systems to restore Reactor Water Level prior to entry into this classification. The time basis for deciding whether or not Reactor Water can be maintained> -200" should be based on the rate of reactor depressurization, the availability of low pressure injection sources, (ECCS and alternate injection systems), and the rate of Reactor coolant inventory loss. Indications such as Reactor Water Level trend, injection flow rates, containment parameter trends, and low pressure injection system operability should also be considered. In the event, Reactor Water Level can not be restored> -200", containment flooding will be required by the EOPs. This will attempt to flood the containment as a means of flooding the RPY, and use a flooded containment as a heat sink for the nuclear fuel. DEVIATION None EAL - 3.1.1.b Rev.00 Page 2 of 3

HCGS EALIRALTechnical Basis REFERENCES NUMARC NESP-0007, FC2 HC.OP-EO.ZZ-0100 (Q)-FC, Reactor Scram HC.OP-EO.ZZ-0101 (Q)-FC, Reactor Pressure Vessel (RPV) Control HC.OP-EO.ZZ-0201 (Q)-FC, Alternate Level Control HC.OP-EO.ZZ-0207 (Q)-FC, Level/Power Control HC.OP-EO.ZZ-0208 (Q)-FC, Primary Containment Flooding BWR Owners Group Emergency Procedure Guidelines, Revision 4 EAL - 3.1. l.b Rev. 00 Page 3 of 3

HCGS EAL/RALTechnical Basis 3.0 Fission Product Barriers 3.1 Fuel Clad Barrier 3.1.2 DRYWELL ATMOSPHERE POST ACCIDENT (DAPA) RADIATION LEVEL IC Loss of Fuel Clad Barrier= 4 POINTS EAL OAP A Radiation Monitor reading ,:: 5000 R/hr OPERATIONAL CONDITION - 1, 2, 3 BASIS Drywell Atmosphere Post Accident (DAP A) Radiation monitors indicating 5000 R/hr or greater corresponds to an instantaneous release of Reactor Coolant with a concentration of 300 µCi/gm Dose Equivalent Iodine-13 I (DEl-131) into the Primary Containment . This value of Reactor Coolant Activity is well above the threshold that could occur as the result oflodine Spiking, resin/chemical intrusion transients or a HWCI System malfunction. This activity level corresponds to fuel clad damage of approximately 3. 8%. In addition, there are other events that could cause Drywell Atmosphere radiation levels to increase to this threshold, without a LOCA in the Drywell. These events involve shine from the reactor core if it is uncovered. While such events would not necessarily involve the calculated fuel clad damage percentage, they would be classifiable under other EALs at a Site Area Emergency level or higher. Barrier Analysis Fuel Clad Barrier has been lost. ESCALATION CRITERIA Emergency Classification will escalate based on the Potential Loss or Loss of additional Fission Product Barriers per EAL Section 3.0. EAL - 3.1.2 Rev. 00 Page 1 of 2

HCGS EALIRALTechnical Basis DISCUSSION EAL 3.1.3 provides a core damage analysis showing that a Reactor Coolant activity of300 µCi/gm Dose Equivalent Iodine-13 l(DEI) is indicative of3.8% clad damage. Using Attachment 2 ofEPIP 205H, 1% clad damage is indicated by a DAPA reading of l.4E3 R/hr at 0.1 hrs after shutdown (the most conservative). This is shown on the Attachment as the 0.1 % TID line. Extrapolating to the 3.8% clad damage point gives 5.32E3 R/hr. This is rounded to 5.0E3 R/hr. Hence, the Fuel Clad Barrier is lost. NUMARC EAL RC3 addresses the use ofDAPA to assess the status of the RCS Barrier, based on the release of Reactor Coolant into the Drywell. This EAL threshold is calculated assuming the instantaneous release and dispersal of the Reactor Coolant noble gas and iodine inventory associated with normal operating concentrations (within TS limits) into the Drywell Atmosphere. The reading would be lower than the threshold for EAL 3 .1.2, thus being indicative of an RCS leak only. However, due to the inability of the DAPA radiation monitors to distinguish between a cloud of released RCS gases and shine from the Reactor Vessel and adjacent piping and components, this EAL is being omitted, as permitted by the NUMARC EALs, and other indications of RCS Leakage are being used. It should be recognized that DAPA exceeding 5000 R/hr would most likely occur due to core uncovery, as Reactor Water Level decreases below the Top of Active Fuel. This condition will result in appropriate escalation to a Site Area Emergency in the Fission Product Barrier Table, and hence use ofDAPA exceeding 5000 R/hr is not needed to detect a Loss of the RCS Barrier. DEVIATION None REFERENCES NUMARC NESP-007, FC3 NUMARC NESP-007, RC3 EPIP 205H, TSC - Post Accident Core Damage Assessment HCOP-AR. SP-000 I (Q), Radiation Monitoring System Alarm Response EAL - 3.1.2 Rev. 00 Page 2 of 2

HCGS EAL/RALTechnical Basis 3.0 Fission Product Barriers 3.1 Fuel Clad Barrier 3.1.3 RCS IODINE CONCENTRATION IC Loss of Fuel Clad Barrier= 4 POINTS EAL Reactor Coolant Sample Activity ~ 300 µCi/gm Dose Equivalent 1-131 OPERATIONAL CONDITION - 1, 2, 3 BASIS Reactor Coolant sample analysis with specific activity greater than or equal to 300 µCi/gm Dose Equivalent I-131 (DEI-131) indicates fuel clad damage due to significant clad heating or mechanical stress, causing a Loss of the Fuel Clad Barrier. This threshold is well above the activity level that could occur as the result oflodine spiking. The use of the term "Valid" as a qualifier for event classification is not required, since Reactor Coolant Activity of this magnitude can only occur as the result of fuel clad damage. This activity level corresponds to approximately 3.8% fuel clad damage. Barrier Analysis Fuel Clad Barrier has been lost. ESCALATION CRITERIA Emergency Classification will escalate based on the Potential Loss or Loss of additional Fission Product Barriers per EAL Section 3.0. EAL - 3.1.3 Rev. 00 Page 1 of 3

HCGS EALIRALTechnical Basis DISCUSSION The percentage of Fuel Damage that corresponds to an RCS Activity of 300 µCi/gm DEl-131 is calculated as follows (for purposes of this calculation, cc and gm are considered equivalent): Dose Factors (RG-1. l 09) 1-131=4.39E-3 1-132 = 5.23E-5 1-133 = l.04E-3 1-134 = 1.37E-5 I-135 = 2.14E-4 Total core inventory (HCGS-UFSAR, table 12.2-135). This table gives 50% inventory, so table values are multiplied by 2.0. 1-131=8.64E7 Ci 1-132 = l.29E8 Ci 1-133 = l.99E8 Ci 1-134 = 2.32E8 Ci I-135 = 1.81E8 Ci Reactor Water Volume= 13000 cubic feet (HCGS-UFSAR, table 12.3-2) Clad Release Fraction for iodines= 0.02 (Table 4.1, NUREG-1228) The activity of each isotope in the clad would then be: 1-131=8.64E7(0.02)=1.73E6 Ci I-132 = 1.29E8(0.02) = 2.58E6 Ci I-133 = 1.99E8(0.02) = 3.98E6 Ci 1-134 = 2.32E8(0.02) = 4.64E6 Ci 1-135 = 1.81E8(0.02) = 3.62E6 Ci These activities are equivalent to 2.89E6 Ci DEI-131 DEJ-l3l= 4.39E-3(1.73E6) + 5.23E-5(2.58E6) + l.04E-3(3.98E6) + l.37E-5(4.64E6) + 2.14E-4(3.62E6) 4.93£-3 EAL - 3.1.3 Rev. 00 Page 2 of 3

HCGS EAL/RALTechnical Basis Calculating the equivalent concentration: Cone= 2.89E6 Ci(1E6µCi I Ci) = 7 _94 EJµCi/cc 13000 cf(2.8E4 cc I cf) which represents the 100% clad damage concentration. 300 µCi/cc DEI-131 is then equivalent to: 300 µCi I cc

                          = 3.78%

7.94E3 µCi I cc This is rounded to 3.8%. DEVIATION None REFERENCES NUMARC NESP-007, FCI HC.OP-AB.ZZ-OIOO(Q), High Reactor Coolant Activity HC.OP-AB.ZZ-0203(Q), Main Steam Line High Radiation HCGS Technical Specification LCO 3.4.5 NUREG 1228 - Source Term Estimation During Incident Response to Severe Nuclear Power Plant Accidents, Table 4.1 Reg. Guide 1.109, Table E-9 HCGS-UFSAR, Table 12.2-135 and Table 12.3-2 IO CFRlOO EAL- 3.1.3 Rev. 00 Page 3 of 3

HCGS EALIRALTechnical Basis 3.0 Fission Product Barriers 3.1 Fuel Clad Barrier 3.1.4 EMERGENCY COORDINATOR JUDGMENT 3.1.4.a/ 3.1.4.b IC Potential Loss(= 3 POINTS) or Loss of Fuel Clad Barrier(= 4 POINTS) EAL ANY condition, in the opinion of the EC, that indicates EITHER a Potential Loss OR Loss of the Fuel Clad Barrier OPERA TI ON AL CONDITION - 1, 2, 3 BASIS This EAL allows the Emergency Coordinator (EC) to address any condition that effects the integrity of the Fuel Clad Barrier that is not already covered elsewhere in the Fission Product Barrier Table. A complete loss of the ability to monitor the Fuel Clad Barrier should be considered as a "Potential Loss" of that barrier. Barrier Analysis Fuel Clad Barrier has been potentially lost or lost. ESCALATION CRITERIA Emergency Classification will escalate based on the potential loss or loss of additional Fission Product Barriers per EAL Section 3.0. DISCUSSION None DEVIATION None EAL - 3 .1.4.a/ 3. l .4.b Rev. 00 Page 1 of 2

HCGS EALIRALTechnical Basis REFERENCES NUMARC NESP-007, FCS EAL - 3.1.4.a/ 3.1.4.b Rev. 00 Page 2 of 2

HCGS EALIRALTechnical Basis 3.0 Fission Product Barriers 3.2 RCS Barrier 3.2.1 REACTOR WATER LEVEL 3.2.1.a IC Potential Loss of RCS Barrier= 3 POINTS EAL Reactor Water Level REACHES -129", EXCLUDING intentional lowering of Reactor Water Level during an ATWS OPERATIONAL CONDITION - 1, 2, 3 BASIS Reactor Water Level reaching -129", excluding intentional lowering of Reactor Water Level during an ATWS, indicates that the inventory loss from the RCS exceeds the capacity of available High Pressure injection sources. Below this threshold, a challenge to maintaining Adequate Core Cooling by core submergence exists, based on Reactor Water Level continuing to decrease, thus a Potential Loss of the RCS Barrier exists. Without core submergence, the integrity of the Fuel Clad would be in jeopardy. Appropriate classification under this EAL is based on reaching Reactor Water Level of -129" (instead of being able to restore and maintain above -129"), due to the challenge that exists to core submergence. Reactor Water Level reaching this threshold results from either a LOCA exceeding available makeup capacity or a Total Loss of High Pressure injection capability. In addition, during an Anticipated Transient Without Scram (ATWS), it is possible that operator action will be taken to intentionally lower Reactor Water Level to below -129". for Reactor Power Control purposes. For this event, classification must be made in accordance with EAL Section 5.0. Barrier Analysis RCS Barrier has been potentially lost. EAL - 3 .2. l.a Rev. 00 Page 1 of 3

HCGS EALIRALTechnical Basis ESCALATION CRITERIA Emergency Classification will escalate based on the Potential Loss or Loss of additional Fission Product Barriers per EAL Section 3.0. DISCUSSION Core Submergence is the preferred method of maintaining adequate core cooling. When Reactor Water Level decreases to -129", a significant challenge to continued core submergence exists. The threshold for this EAL corresponds to the initiation setpoint for the low pressure Emergency Core Cooling Systems (ECCS). Reactor Water Level reaching -129" occurs when the rate of inventory loss is greater than the rate of inventory makeup from High Pressure injection sources. This condition can occur as the result of the following events/sequences (excluding intentional lowering of Reactor Water level during an ATWS). A LOCA will cause Reactor Water Level to reach -129" when the LOCA is the result of a large break (momentary core uncovery is expected to occur under this condition) or when the LOCA is due to a small or intermediate break in combination with an inability of High Pressure injection sources to keep up with the leak rate. A Loss of High Pressure injection sources without the presence of a LOCA will also result in Reactor Water Level I decreasing to -129", due to continued Reactor Steam Flow without makeup. Either of these events/sequences results in a potential challenge to the RCS Barrier when Reactor Water level reaches -129", hence classification at this threshold is appropriate. However, for both these sequences, low Pressure ECCS are designed to inject to the Reactor as Reactor Pressure decreases below the shutoff head of the pumps. Reactor Depressurization will occur either due to the LOCA or Manual initiation of Emergency Depressurization when Reactor Water Level reaches -161 ", provided injection systems are available. This will allow for restoration of Reactor Water Level and re-establishment of Core Submergence. DEVIATION None EAL - 3.2.1.a Rev.00 Page 2 of 3

HCGS EALIRALTechnical Basis REFERENCES NUMARC NESP-0007, RCS HC.OP-SO.SM-OOOl(Q), Isolation Systems Operation HC.OP-AB.ZZ-0116 (Q), Containment Isolation and Recovery From An Isolation HC.OP-AB.ZZ-0200 (Q), Reactor Low Water Level HC.OP.EO.ZZ-0100 (Q)-FC, Reactor Scram HC.OP.EO.ZZ-0101 (Q)-FC, Reactor Pressure Vessel (RPV) Control HCGS Technical Specifications LCO 3/4.3, Instrumentation EAL - 3.2. l.a Rev. 00 Page 3 of 3

HCGS EAL/RALTechnical Basis 3.0 Fission Product Barriers 3.2 RCS Barrier 3.2.1 REACTOR WATER LEVEL 3.2.1.b IC Loss of RCS Barrier= 4 POINTS EAL Reactor Water Level REACHES -161" (Top of Active Fuel), EXCLUDING intentional lowering of Reactor Water Level during an ATWS OPERATIONAL CONDITION - 1, 2, 3 BASIS Reactor Water Level reaching -161" (Top of Active Fuel - T AF), excluding intentional lowering of Reactor Water Level during an ATWS, results in an inability to maintain adequate core cooling by core submergence, causing a Loss of the RCS Barrier. Without core submergence, the integrity of the fuel clad barrier is in jeopardy. Appropriate classification under this EAL is based on reaching Reactor Water Level of -161" (instead of being able to restore and maintain above -161 "} due to the potentially severe consequences of a loss of core submergence. Reactor Water Level reaching this threshold results from either a LOCA exceeding available makeup capacity or a Total Loss of High Pressure injection capability. In addition, during an Anticipated Transient Without Scram (ATWS}, it is possible that operator actions will be taken to intentionally lower Reactor Water Level to between -161" and -190", for Reactor Power Control purposes. For this event, classification must be made in accordance with EAL Sec(on 5.0 Barrier Analysis RCS Barrier has been lost. ESCALATION CRITERIA Emergency Classification will escalate based upon the Potential Loss or Loss of additional Fission Product Barriers per EAL Section 3.0. EAL - 3.2.1.b Rev. 00 Page 1 of 3

HCGS EALIRALTechnical Basis DISCUSSION Core Submergence is the preferred method of maintaining adequate core cooling. When Reactor Water Level decreases to below T AF, the ability to effectively remove decay heat is being challenged, and as such the Fuel Clad barrier can no longer be considered intact. While the . Emergency Operating Procedures provide contingencies to establish adequate core cooling when Reactor Water Level drops below T AF (Steam Cooling with or without injection), these actions .ire designed to be an alternative method of providing adequate core cooling while actions Jre taken to reestablish core submergence. Sustained partial or total core uncovery can result in fuel clad damage and a significant release of fission products to the Reactor coolant. Sustained core uncovery can also result in a breach of the Reactor Vessel due to core melt material interaction with the RPV. A Loss of Core Submergence will occur when the rate of inventc')' loss is greater than the rate of inventory makeup from High Pressure injection sources. This condition can occur as the result of the following events/sequences (excluding intentional lowering of Reactor Water Level during an ATWS). A LOCA will cause Reactor Water Level to reach the Top of Active Fuel when the LOCA is the result of a large break (momentary core uncovery is expected to occur under this condition) or when the LOCA is due to a small or intermediate break in combination with an inability of High Pressure injection sources to keep up with the leak rate. A Loss of High Pressure injection sources without the presence of a LOCA will also result in Reactor Water Level decreasing to T AF, due to continued Reactor Steam Flow without makeup. Either of these events/sequences results in a challenge to the Fuel Clad Barrier when Reactor Water Level reaches T AF due to core uncovery, hence classification at this threshold is appropriate. However, for both these sequences, Low Pressure ECCS are designed to inject to the Reactor as Reactor Pressure decreases below the shutoff head of the pumps. Reactor Depressurization will occur either due to the LOCA or Manual initiation of Emergency Depressurization when Reactor Water Level reaches -161 ",provided injection systems are available. This will allow for restoration of Reactor Water Level and re-establishm~nt of Core Submergence. DEVIATION None EAL - 3.2. l.b Rev. 00 Page 2 of 3

HCGS EAL/RALTechnical Basis REFERENCES NUMARC NESP-0007, RC4 HC.OP-EO.ZZ-0100 (Q)-FC, Reactor Scram HC.OP-EO.ZZ-0101 (Q)-FC, Reactor Pressure Vessel (RPV) Control HC.OP-EO.ZZ-0201 (Q)-FC, Alternate Level Control HC.OP-EO.ZZ-0207 (Q)-FC, Level/Power Control BWR Owner's Group Emergenry Procedure Guidelines, Rev. 4 EAL - 3.2.1.b Rev. 00 Pagi:: 3 of 3

HCGS EAL/RALTechnical Basis 3.0 Fission Product Barriers 3.2 RCS Barrier 3.2.2 RCS LEAK RA TE/DRYWELL PRESSURE 3.2.2.a IC Potential Loss of RCS Barrier= 3 POINTS EAL Unisolable RCS Leak Rate~ 50 GPM INSIDE Primary Containment OPERATIONAL CONDITION - 1, 2, 3 BASIS Unisolable RCS Leak Rate exceeding 50 GPM, inside Primary Containment is indicative of a potential loss of the RCS. An unisolable leak rate of this magnitude is significant due to the potential for further break propagation, resulting in a much higher loss of inventory with an ; inability to isolate the leak source. As such, this threshold is considered a Potential Loss of the / RCS. Leakage just above the 50 GPM threshold is well within the capacity of normal and emergency injection systems and is not a significant concern for core uncovery. However, 50 GPM is the minimum leak rate that would be classified under this EAL, with the maximum being equivalent to the leak rate that would result in either Reactor Water Level reaching -129" or Drywell Pressure reaching 1.68 PSIG, since these two conditions are obviously more recognizable to Control Room personnel, than an existing leak rate. Specifying an unisolable RCS leak as part of the threshold for this EAL, precludes classifying events such as an isolable Reactor Recirculation Pump dual seal failure under this EAL. Barrier Analysis RCS Barrier has been potentially lost. ESCALATION CRITERIA Emergency Classification will escalate based on the Potential Loss or Loss of additional Fission Product Barriers per EAL Section 3.0. EAL - 3 .2.2.a Rev. 00 Page 1 of 2

HCGS EAL!RALTechnical Basis DISCUSSION It is important to recognize that the unisolable RCS leak rate established in this EAL is inside the Primary Containment. The inability to isolate the leak would eventually lead to a High Drywell Pressure(> 1.68 PSIG) actuation of RPS, ECCS and PCIS. The actuation would lead to an isolation of the Drywell Floor and Equipment Drain sumps, complicating efforts to further identify and quantify any changes in the existing leak rate. In addition, monitoring oft~.~ leak rate could be limited by reaching the upper range (50 GPM) of the Drywell Leak Detection channels (9AX313 - Equipment, 9AX314- Floor Drain). For leakage outside Containment, since quantification of the leak rate is much more difficult due to the physical size of the Reactor Building, receipt of a Valid isolation signal has been established as the threshold for classification of this type of leakage. DEVIATION None REFERENCES NUMARC NESP-007, RCI NUMARC Questions and Answers, June I 993, "Fission Product Barrier Question #I I" HC.OP-SO.SM-OOOl(Q), Isolation Systems Operation HC.OP-AB.ZZ-OI I6(Q), Containment Isolations and Recovery from an Isolation HC.OP-AB.ZZ-020I(Q), Drywell High Pressure/Loss ofDrywell Cooling HC.RP-AR.SP-OOOI(Q), Radiation Monitoring System Alarm Response HC.OP-EO.ZZ-OIOO(Q)-FC, Reactor Scram HC.OP-EO.ZZ-OIOI(Q)-FC, Reactor Pressure Vessel (RPV) Control HC.OP-EO.ZZ-OI02(Q)-FC, Primary Containment Control HC.OP-EO.ZZ-OI03(Q)-FC, Secondary Containment Control HC.OP-GP.ZZ-0005(Q), Drywell Leakage Source Detection EAL - 3.2.2.a Rev. 00 Page 2 of 2

HCGS EALIRALTechnical Basis 3.0 Fission Product Barriers 3.2 RCS Barrier 3.2.2 RCS LEAK RA TE/DRYWELL PRESSURE 3.2.2.b IC Loss of RCS Barrier= 4 POINTS EAL Valid High Drywell Pressure Condition ( ~ 1.68 psig) OPERATIONAL CONDITION - I, 2, 3 BASIS A Valid High Drywell Pressure Condition (2: 1.68 PSIG) is indicative of the release of high energy Reactor Coolant from the RCS into the Drywell and hence is considered a Loss of the RCS Barrier. Valid is defined as the High Drywell Pressure condition specifically due to RCS leakage into the Drywell, ensuring that event classification under this EAL is truly reflective of a degraded RCS Barrier. This precludes unwarranted event declaration as the result of system malfunctions, including a loss of Drywell Cooling or inadvertent Drywell makeup. Indication of an RCS leak should be positively determined by observing Primary Containment parameters, including Drywell Pressure and Temperature trends, Drywell Equipment and Floor Drain sump levels, DAPA Radiation levels, atmospheric pressure, Torus Pressure, and the status ofDrywell Cooling systems. An isolable Reactor Recirculation Pump dual seal failure should not result in Drywell Pressure reaching the threshold for this EAL, hence classification under this EAL should not occur. Barrier Analysis RCS Barrier has been lost. ESCALATION CRITERIA Emergency Classification will escalate based on the Potential Loss or Loss of additional Fission Product Barriers per EAL Section 3.0. EAL - 3.2.2.b Rev. 00 Page 1 of 2

HCGS EALIRALTechnical Basis DISCUSSION RCS Leakage into the Drywell exceeding 50 GPM is substantially greater than the RCS leakage thresholds established in EAL Section 2.1.1, and represents further degradation of the RCS barrier. Inability to isolate the RCS leakage would eventually result in a High Drywell Pressure (> 1.68 PSIG) actuation of RPS, ECCS and PCIS. The actuation would lead to an isolation of the Drywell Floor and Equipment Drain sumps, complicating efforts to further identify and quantify any changes in the leak rate. In addition, monitoring of the leak rate could be limited by reaching the upper range (50 GPM) of the Drywell Leak Detection channels (9AX313 - Equipment, 9AX314 - Floor Drain). There are multiple Control Room indicators and alarms which can be used to determine the presence of a High Drywell Pressure condition. Overhead Annunciators will alarm at 1.5 PSIG and 1.68 PSIG. Plant automatic response to a High Drywell Pressure condition includes: a reactor scram, ECCS initiation, trip of the drywell cooling fans and isolation of the cooling water to the drywell. These actuations may mask the trend in drywell pressure. For example, the scram will result in less heat being added to the containment and the cooling water isolation will result in no heat being removed. Actions initiated as part of increasing drywell pressure condition include investigation of the source of the increased leakage into the drywell. maximizing drywell cooling and venting the Drywell (if release criteria can be satisfied). These actions are designed to control and relieve increasing drywell pressure. DEVIATION None REFERENCES NUMARC NESP-0007, RC2 NUMARC Questions and Answers, June 1993, "Fission Product Barrier Question #11" HC.OP-SO.SM-OOOl(Q), Isolation Systems Operation HC. OP-AB.ZZ-0116 (Q), Containment Isolation and Recovery From An Isolation HC.OP-AB.ZZ-0201 (Q), Drywell High Pressure/Loss ofDrywell Cooling HC.OP-EO.ZZ-0100 (Q)-FC, Reactor Scram HC.OP-EO.ZZ-0101 (Q)-FC, Reactor Pressure Vessel (RPV) Control HC.OP-EO.ZZ-0102 (Q)-FC, Primary Containment Control HC.OP-GP.ZZ-0005 (Q), Drywell Leak Source Detection Hope Creek Appendix A based on NED0-2121, Supplement A to BWR Owners Group Emergency Procedure Guidelines, Revision 4 HCGS Technical Specifications LCO 3/4.3, Instrumentation EAL - 3.2.2.b Rev. 00 Page 2 of 2

HCGS EAL/RALTechnical Basis 3.0 Fission Product Barriers 3.2 RCS Barrier 3.2.3 RCS LINE BREAK/CONTAINMENT BYPASS 3.2.3.a IC Potential Loss of RCS Barrier= 3 POINTS EAL Main Steam Line Break OUTSIDE Primary Containment , resulting in an AUTOMATIC MSIV Isolation Signal ALL 4 Main Steam Lines have been successfully isolated based on NO indication of CONTINUING FLOW I LEAKAGE OUTSIDE the Primary Containment AFTER valve closure from the Main Control Room has been attempted OPERATIONAL CONDITION - 1, 2, 3 BASIS A Main Steam Line Break outside the Primary Containment, resulting in an automatic MSIV Isolation Signal, could result in dose consequences offsite from a "puff' release in excess of I 0 millirem, based on design basis accident analysis, even if MSIV closure occurs within design limits. Hence this condition is classified as a Potential Loss of the RCS Barrier. Classification under this EAL is specifically for a Main Steam Line Break outside the Primary Containment, as evidenced by a rapid change in Main Steam Line Flow and Steam Tu.1nel Temperature, that results in automatic isolation with no indication of continuing leakage. Valve Packing leaks that result in elevated Steam Tunnel temperatures do not require classification under this EAL. A manual actuation of NS SSS or manual MSIV closure PRIOR to exceeding the setpoints that would result in an automatic isolation of the MSIV should not result in a "puff' release exceeding 10 millirem, and thus should not be classified under this EAL. Verification that continuing leakage does not exist, ensures that any potential release will not significantly exceed the I 0 CFR I 00 limits. This EAL is specific to a break outside the Primary Containment, since a break outside represents a potential challenge to Primary Containment Integrity due to the Containment Bypass condition that would exist until MSIV closure occurred . Failure to completely isolate the effected EAL - 3.2.3.a Rev.00 Page 1 of 3

HCGS EAL/RALTechnical Basis Main Steam Line(s) as determined by valve position and indication of continuing leakage would result in an additional Loss of the Primary Containment Barrier. In addition; this EAL ALLOWS for valve closure from the Main Control Room to isolate any Main Steam Line that did not completely isolate. Valve closure is defined as the closure of ANY valve from the Main Control Room associated with the effected Main Steam Line(s), that did not completely isolate. For example, ifthe isolation logic fails to cause valve closure, but operator actions implemented in the Main Control Room successfully i~ulates the effected Main Steam Line(s), then event classification under this EAL is warranted due to the consequences of the event previously discussed. This includes Motor Operated Valves that are not controlled by the isolation logic, but are manually controlled from the Main Control Room. (i.e. Main Steam Stop Valves lABHV-3631 A/B/C/D). In the event the effected Main Steam Line(s) can not be isolated, escalation of the classification will be required. Barrier Analysis RCS Barrier has been potentially lost ESCALATION CRITERIA Emergency Classification will escalate based on the Potential Loss or Loss of additional barriers per EAL section 3. 0. DISCUSSION The Main Steam System is associated with systems that are part of the RCS boundary and penetrate the Primary Containment. Isolation requirements for these lines are covered in 10CFR50, Appendix A, General Design Criteria 55. These systems form a closed loop outside the Primary Containment and are not open or potentially open to the environment. These systems represent an extension of the RCS Barrier beyond the Primary Containment. Positive identification of a Main Steam Line Break outside the Primary Containment can be based on receipt of the following Overhead Annunciators: NSSSS ISLN SIG - STM TNL TEMP HI (C8-C4) NSSSS ISLN SIG - MN STM FLOW HI (C8-B4) MSIV CLOSURE (C5-B3) as well as the following indications: MSIV TRIP LOGIC TRIPPED Rapid changes in Main Steam Line Flow and Steam Tunnel Temperatures EAL - 3.2.3.a Rev.00 Page 2 of 3

HCGS EAL/RALTechnical Basis DEVIATION This EAL is being maintained in the Fission Product Barrier Table for ease of use by the operators. It has been categorized as a "Potential loss" since the RCS leak is successfully isolated and an Alert classification will still be made as a result of the potential loss of RCS. REFERENCES NUMARC NESP-007, RC I NUMARC Question and Answer, June 1983, "Fission Product Barrier- BWR" Question #4 10 CFR50, App. A, GDC 55 10 CFR 100 HC.OP-SO.SM-OOOl(Q), Isolation Systems Operation HC.OP-AB.ZZ-Ol 14(Q), Loss of Primary Containment Integrity HC.OP-AB.ZZ-01l6(Q), Containment Isolations and Recovery from an Isolation HC.OP-AB.ZZ-0203(Q), Main Steam Line High Radiation HC.OP-AR.SP-OOOl(Q), Radiation Monitoring System Alarm Response HC.OP-AR.ZZ-001 l(Q), Annunciator Response Procedures, Window C6 HC.OP-AR.ZZ-0012(Q), Annunciator Response Procedures, Window C8 HC.OP-EO.ZZ-0100 (Q)-FC, Reactor Scram HC.OP-EO.ZZ-0101 (Q)-FC, Reactor Pressure Vessel (RPV) Control HC.OP-EO.ZZ-0102 (Q)-FC, Primary Containment Control HC.OP-EO.ZZ-0103 (Q)-FC, Reactor Building Control HC.OP-EO.ZZ-0104 (Q)-FC, Radioactive Release Control HCGS Technical Specifications, LCO 3/4.3 HCGS UFSAR, Section 6.2.4.3.1 EAL - 3.2.3.a Rev. 00 Page 3 of 3

HCGS EAL/RALTechnical Basis 3.0 Fission Product Barriers 3.2 RCS Barrier 3.2.3 RCS LINE BREAK/CONTAINMENT BYPASS 3.2.3.b IC Loss of RCS Barrier= 4 POINTS EAL RCS Line Break OUTSIDE Primary Containment, resulting in a Valid Isolation Signal for ANY one of the following systems:

NSSSS
HPCI
RCIC Indication of CONTINUING FLOW I LEAKAGE OUTSIDE the Primary Containment through the effected system AFTER valve closure from the Main Control Room has been attempted OPERATIONAL CONDITION- 1, 2, 3 BASIS An RCS Line Break outside Primary Containment that results in a Valid Isolation Signal for any of the systems listed in the EAL requires closure of the associated Primary Containment Isolation valves to maintain RCS and Primary Containment integrity under abnormal conditions. A failure of these isolation valves to isolate directly allows Reactor Coolant to be released outside the Primary Containment (Containment Bypass). resulting in a Loss of RCS and Loss of Containment. An RCS Line is ANY line that communicates directly with the Reactor. An RCS Line Break with indication of continuing flow is classified under thi~ EAL, due to the continuing discharge of Reactor Coolant outside the Primary Containment along with a potential for further "break propagation". This is the only condition that warrants classification under this EAL.

Valid is defined as the isolation signal specifically being the result of an RCS Line Break, thus ensuring that the RCS discharge is of significant.magnitude to pose a threat to the integrity of the RCS Barrier. This precludes unwarranted Event Classification as the result of condition that EAL - 3.2.3.b Rev. 00 Pag~ 1 of 3

HCGS EALIRALTechnical Basis result in limited leakage with no potential for "break propagation", including valve packing leaks outside Primary Containment and RWCU Pump Seal Leaks. In addition, isolation signal generated from known failures in other systems, that do not result in Reactor Coolant discharging outside the Primary Containment do not warrant Event Classification under this EAL either. Examples of such failures include a high temperature isolation resulting from a loss of ventilation or cooling water, spurious actuation during I&C surveillance testing or a low Reactor Water Level Condition due to a Loss of High Pressure injection capability. In addition, this EAL ALLOWS for valve closure from the Main Control Room to isolate any systems that did not completely isolate, prior to event classification. Valve closure is defined as the closure of ANY valve from the Main Control Room in the system( s) that did not completely isolate. For example, if the isolation logic fails to cause valve closure, but operator actions implemented in the Main Control Room successfully isolates the effected system, then classification under this EAL is not warranted. This includes Motor Operated Valves that are not control by the isolation logic, but are manually controlled from the Main Control Room. Effected system is defined as the system that is providing the flowpath outside the Primary Containment. Barrier Analysis RCS Barrier has been lost ESCALATION CRITERIA Emergency Classification will escalate based on the Potential Loss or Loss of additional Fission Product Barriers per EAL Section 3.0. DISCUSSION NSSSS isolations, as well as HPCI and RCIC steam line isolations, are associated with systems that are part of the RCS boundary and penetrate the Primary Containment. Isolation requirements for these lines are covered in IOCFR50, Appendix A, General Design Criteria 55. These systems form a closed loop outside the Primary Containment, and are not open or potentially open to the environment. They are included in this EAL since they represent an extension of the RCS boundary beyond the Primary Containment, and a potential release path from the RCS to the environment. Without a completed isolation, continuing flow/leakage represents a situation where Reactor Coolant is discharging outside the Primary Containment, including areas in the Reactor Building addressed in the EOPs. Indication of continuing flow/leakage includes: flow indication through isolated lines, increasing Reactor Building area temperatures, area radiation levels, sump levels, or room levels in spaces associated with affected lines, as well as increases in Plant Vent Eflluent levels. EAL - 3.2.3.b Rev. 00 Page 2 of 3

HCGS EAL!RALTechnical Basis DEVIATION This EAL is being considered a loss of the reactor coolant boundary since actuation of listed isolation system indicate a leak of significant magnitude, and an isolation failure. The classification for exceeding this EAL remains consistent with NUMARC guide lines. REFERENCES NUMARC NESP-007, RCl 10 CFR50, App. A, GDC 55 10 CFR 100 HC.OP-SO.SM-OOOl(Q), Isolation Systems Operation HC.OP-AB.ZZ-0114(Q), Loss of Primary Containment Integrity HC.OP-AB.ZZ-0116(Q), Containment Isolations and Recovery from an Isolation HC.OP-AB.ZZ-0203(Q), Main Steam Line High Radiation HC.OP-AR.SP-OOOl(Q), Radiation Monitoring System Alarm Response HC.OP-AR.ZZ-001 l(Q), Annunciator Response Procedures, Window C6 HC.OP-AR.ZZ-0012(Q), Annunciator Response Procedures, Window CS HC.OP-EO.ZZ-0100 (Q)-FC, Reactor Scram HC.OP-EO.ZZ-0101 (Q)-FC, Reactor Pressure Vessel (RPY) Control HC.OP-EO.ZZ-0102 (Q)-FC, Primary Containment Control HC.OP-EO.ZZ-0103 (Q)-FC, Reactor Building Control HC.OP-EO.ZZ-0104 (Q)-FC, Radioactive Release Control HCGS Technical Specifications LCO 3/4.3, Instrumentation HCGS UFSAR, Section 6.2.4.3. l EAL - 3.2.3.b Rev. 00 Page 3 of 3

HCGS EAL/RALTechnical Basis 3.0 Fission Product Barriers 3.2 RCS Barrier 3.2.4 EMERGENCY COORDINATOR JUDGMENT 3.2.4.a/ 3.2.4.b IC Potential Loss(= 3 POINTS) or Loss of RCS Barrier(= 4 POINTS) EAL ANY condition, in the opinion of the EC, that indicates EITHER a Potential Loss OR Loss of the RCS Barrier OPERATIONAL CONDITION - l, 2, 3 BASIS This EAL allows the Emergency Coordinator (EC) ~o address any condition that affects the integrity of the RCS Barrier that is not already covered elsewhere in the Fission Product Barrier Table. A complete loss of the ability to monitor the RCS barrier should be considered as a "Potential Loss" of that barrier. Barrier Analysis RCS Barrier has been potentially lost or lost. ESCALATION CRITERIA Emergency Classification will be escalate based on the Potential Loss or Loss of additional barriers per EAL section 3.0. DISCUSSION None DEVIATION None EAL - 3.2.4.a/ 3.2.4.b Rev. 00 Page 1 of 2

HCGS EALIRALTechnical Basis REFERENCES NUMARC NESP-007, RC6

                                       ~AL - 3.2.4.a/ 3.2.4.b Rev. 00 Page 2 of 2

HCGS EAL/RALTechnical Basis 3.0 Fission Product Barriers 3.3 Containment Barrier 3.3.1 REACTOR WATER LEVEL IC Potential Loss of Containment Barrier = 1 POINT EAL Reactor Water Level CANNOT BE RESTORED AND MAINTAINED above -200" (Minimum Zero Injection RPV Water Level) OPERATIONAL CONDITION - I, 2, 3 BASIS Inability to restore and maintain Reactor Water Level above - 200" (Minimum Zero Injection RPV Water Level), results in a loss of adequate core cooling by all mechanisms, causing a Potential Loss of the Fuel Clad Barrier. Without adequate core cooling, the integrity of the Containment is being challenged and can no longer be assured. Appropriate classification under this EAL is based on the failure of injection systems to restore and_maintain Reactor Water Level above -200", following a condition .that causes level to decrease below the threshold. For example, a large break LOCA is expected to cause Reactor Water Level to momentarily decrease below -200", due to the response time of Low Pressure ECCS. As these systems initiate and commence injection to the Reactor, water level will begin to increase and should be able to be maintained above -200". In this case, classification under this EAL is not appropriate as plant systems have performed their intended design function and will eventually restore adequate core cooling by core submergence. However, in the event that Low Pressure ECCS and alternate injection system, as defined in the EOPs are in a degraded condition (i e., Station Blackout, ECCS Suction Strainer plugging, etc.) and Reactor Water Level can not be restored and maintained above -200", then classification under this EAL should occur due to the Potential Loss of Containment from the release of energy to the containment from imminent fuel failure. Barrier Analysis Primary Containment Barrier has been potentially lost. EAL - 3.3.1 Rev.00 Page 1 of 3

HCGS EAL/RALTechnical Basis ESCALATION CRITERIA Emergency Classification will escalate based upon the Potential Loss or Loss of additional Fission Product Barriers per EAL Section 3.0. DISCUSSION Core submergence is the preferred method for maintaining adt..quate core cooling. The failure to reestablish Reactor Water Level above -16 l ", the Top of Active Fuel (T AF), for an extended period of time could lead to significant fuel damage. With Reactor Water Level below TAF, but above the Minimum Zero Injection RPV Water Level (-200"), adequate core cooling occurs due to the cooling effects of steam generated in the covered portion of the core flowing through the uncovered portion (Steam Cooling). The Minimum Zero Injection RPV Water Level is defined in the Emergency Operating Procedures. This method of cooling precludes any fuel clad temperature in the uncovered portion of the core from exceeding l 800°F. As Reactor Water Level drops below -200" with no injection available, this method of cooling becomes inadequate. Prolonged lack of cooling may result in severe overheating of the fuel clad, additional release of energy from accelerated clad oxidation, and eventual fuel melting. For events starting from full power operation, the failure to promptly reflood could result in some fuel melting. Even under these conditions vessel failure and containment failure with resultant release to the public would not be expected for some time. Reactor Water Level remaining below T AF for an extended amount of time represents an early indicator that significant core damage is in progress while providing sufficient time to initiate public protective actions. Ample time should be provided for Low Pressure ECCS and alternate injection systems restore Reactor Water Level prior to entry into this classification. The time basis for deciding whether or not Reactor Water can be maintained> -200" should be based on the rate of reactor depressurization, the availability of low pressure injection sources, (ECCS and alternate injection systems), and the rate of Reactor coolant inventory loss. Indications such as Reactor Water Level trend, injection flow rates, containment parameter trends, and low pressure injection system operability should also be considered. In the event, Reactor Water Level can not be restored> -200", containment flooding will be required by the EOPs. This will attempt to flood the containment as a means of flooding the RPV, and use a flooded containment as a heat sink for the nuclear fuel. DEVIATION None EAL - 3.3.1 Rev. 00 Page 2 of 3

HCGS EAL/RALTechnical Basis REFERENCES NUMARC NESP-007, PC4 HC.OP.EO.ZZ-0100 (Q)-FC, Reactor Scram HC.OP.EO.ZZ-0101 (Q)-FC, Reactor Pressure Vessel (RPV) Control HC.OP.EO.ZZ-0201 (Q)-FC, Alternate Level Control HC.OP.EO.ZZ-208 (Q)-FC, Primary Containment Flooding EWR Owners Group Emergency Procedure Guidelines, Revision 4 EAL - 3.3. l Rev. 00 Page 3 of 3

HCGS EALIRALTechnical Basis 3.0 Fission Product Barriers 3.3 Containment Barrier 3.3.2 DRYWELL PRESSURE/H2 3.3.2.a/ 3.3.2.c IC Potential Loss of Containment Barrier = l POINT EAL Suppression Chamber pressure CANNOT BE MAINTAINED below 65 psig Primary Containment H 2 concentration >4% and 0 2 concentration >5% OPERATIONAL CONDITION - I, 2, 3 BASIS Containment venting required by the EOPs indicates a degrading condition in containment and is implemented in an effort to preclude containment failure. Venting is required before Suppression Chamber pressure reaches 65 PSIG or Hydrogen concentration reaches the Lower Explosive Limit (LEL = 4%) and Oxygen concentration reaches 5%. Exceeding these parameters creates the potential for an unisolable breach of the primary containment, which could result in an uncontrolled, unmonitored, and untreated release of radioactivity to the environment. This EAL represents a Potential Loss of Containment, since containment venting is required due to Containment parameters potentially exceeding their design limits. The magnitude of any radiological release is dependent upon events leading to the requirement for emergency venting, including a loss of the RCS and a loss of the Fuel Clad Barriers. A Downcomer failure, by itself, does not represent a Loss of the Primary Containment Barrier. This failure does, however, render the Primary Containment inoperable per the Technical Specification, as Primary Containment integrity has been compromised. A Downcomer failure combined with a large break LOCA will likely result in a Potential Loss of Primary Containment under this EAL if Containment pressure can not be maintained below 65 PSIG and Containment Venting is required. EAL - 3.3.2.a/ 3.3.2.c Rev. 00 Page 1 of 2

HCGS EALIRALTechnical Basis Barrier Analysis Primary Containment Barrier has been potentially lost. ESCALATION CRITERIA Emergency Classification will escalate based on the Potential Loss or Loss of additional Fission Product Barriers per EAL Section 3.0. DISCUSSION Venting of the Primary Containment is initiated to preserve containment integrity under accident conditions. Primary Containment venting is required when Suppression Chamber cannot be maintained below 65 psig, which is well above the maximum pressure expected to be present in the Primary Containment during a design basis Loss of Coolant Accident (LOCA). Primary Containment venting is also required based on hydrogen concentrations exceeding 4%. Hz concentrations in excess of 6.0 % requires Emergency Depressurization and subsequent containment venting. Venting is continued until either Hz concentration has been reduced to <6.0% or 02 levels have been reduced to <5.0%. Venting with elevated hydrogen concentfation conditions ensures that containment failure resulting from a hydrogen detonation or deflagration does not occur. The elevated hydrogen in the containment may result from excessive zircaloy-water reaction occurring following a LOCA. Additionally, hydrogen and oxygen gas may be introduced into the containment environment from long term disassociation of water in the Suppression Chamber. EOP procedural guidance in these cases is provided to vent the Primary Containment regardless of off-site dose consequences. Although radiological releases resulting from venting containment may exceed EPA limits, a controlled, monitored, and isolable release is preferred to a potential uncontrolled, unmonitored radiological release that would result from a failure of containment. DEVIATION None REFERENCES NUMARC NESP-007, PC I, PC2 HC.OP-AB.ZZ-OZO I (Q), Drywell High Pressure/Loss of Drywell Cooling HC.OP-EO.ZZ-0101 (Q)-FC, Reactor Pressure Vessel (RPV) Control HC.OP-EO.ZZ-0102 (Q)-FC, Primary Containment Control HC.OP-EO.ZZ-0318 (Q)-FC, Containment Venting BWR Owners Group Emergency Procedure Guidelines, Revision 4 EAL - 3.3.Z.a/ 3.J.Z.c Rev.00 Page 2 of 2

HCGS EAL/RALTechnical Basis 3.0 Fission Product Barriers 3.3 Containment Barrier 3.3.2 DRYWELL PRESSURE/H2 3.3.2.b/ 3.3.2.d/ 3.3.2.e IC Loss of Containment Barrier= 2 POINTS EAL Containment Failure as indicated by a rapid drop in Drywell pressure following a rise in pressure above 1.68 psig Drywell pressure response not consistent with LOCA conditions Containment is Vented by the Emergency Operating Procedures (EOPs) OPERATIONAL CONDITION - I, 2, 3 BASIS Containment failure indicated by a rapid decrease in Drywell pressure following a significant rise in Drywell pressure is indicative of a Loss of the Containment barrier. This EAL specifically represents a Loss of Containment, whereby a unisolable breach of the Containment structure has occurred. Conditions that result in a drop in Drywell pressure following a pressure rise that are not the direct result of a Containment failure do not warrant classification under this EAL. These events include the initiation of Drywell Sprays, the re-establishment of Drywell Cooling, Containment Venting as required by the EOPs, and anticipated Drywell pressure drop due to ambient losses. Containment Venting is a controlled loss of containment. This venting is performed for the purpose of preventing an unisolable, unmonitored radiological release of containment gases. A Downcomer failure, by itself, does not represent a Loss of the Primary Containment Barrier. This failure does, however, render the Primary Containment inoperable per the Technical Specification, as Primary Containment integrity has been compromised. A Downcomer failure EAL - >.3.2.b/ 3.3.2.d/ 3.3.2.e Rev. 00 Page 1 of 2

HCGS EALIRALTechnical Basis combined with a large break LOCA will likely result in a Potential Loss of Primary Containment under EAL 3.3.2.a if Containment pressure cannot be maintained below 65 PSIG and Containment Venting is required. Barrier Analysis Primary Containment Barrier has been lost. ESCALATION CRITERIA Emergency Classification will escalate based on the Potential Loss or Loss of additional Fission Product Barriers per EAL Section 3.0. DISCUSSION Appropriate classification under this EAL occurs as the result of a Containment failure. Drywell pressure reaching I. 68 psig indicates that there is a significant release of reactor coolant to the Containment. Unless this source of leakage is isolated or the Reactor is depressurized, Drywell pressure would not be expected to drop in a rapid manner. Other indications such as Reactor Building Area Radiation Monitors (ARMs) radiation levels, Reactor Building area temperatures, Reactor Building floor and sump levels, Plant Effluent radiation levels, and containment isolation status should be used to confirm the loss of containment integrity if possible. Reactor Building to Torus vacuum breaker status should be monitored to ensure that this pathway does not result in a loss of containment integrity. DEVIATION None REFERENCES NUMARC NESP-007, PCl HC.OP-AB.ZZ-0114 (Q), Loss of Primary Containment Integrity HC.OP-AB.ZZ-0116 (Q), Containment Isolations and Recovery from an Isolation HC.OP-AB.ZZ-0201 (Q), Drywell High Pressure/Loss ofDrywell Cooling HC.OP-EO.ZZ-0100 (Q)-FC, Reactor Scram HC.OP-EO.ZZ-0101 (Q)-FC, Reactor Pressure Vessel (RPV) Control HC.OP-EO.ZZ-0102 (Q)-FC, Primary Containment Control HC.OP-EO.ZZ-0103 (Q)-FC, Reactor Building Control BWR Owners Group Emergency Procedure Guidelines, Revision 4 EAL - 3.3.2.b/ 3.3.2.d/ 3.3.2.e Rev. 00 Page 2 of 2

HCGS EALIRALTechnical Basis 3.0 Fission Product Barriers 3.3 Containment Barrier 3.3.3 DRYWELL ATMOSPHERE POST ACCIDENT (DAPA) RADIATION LEVEL IC Potential Loss of Containment Barrier = 1 POINT EAL DAP A Radiation Monitor reading 2::, 28000 R/hr OPERATIONAL CONDITION - 1, 2, 3 BASIS Drywell Atmosphere Post Accident (DAP A) monitor reading~ 28000 R/hr indicates significant fuel damage, well in excess of the level corresponding to the loss of the RCS and Fuel Clad barriers. This threshold corresponds to approximately 20% fuel clad damage. Regardless of whether or not containment is challenged, this amount of activity in containment, if released, could have severe consequences and it is prudent to treat this condition as a Potential Loss of containment. Barrier Analysis Primary Containment Barrier is potentially lost. ESCALATION CRITERIA Emergency Classification will escalate based on the Potential Loss or Loss of additional Fission Product Barriers per EAL Section 3.0. DISCUSSION NUREG-1228, "Source Term Estimation During Incident Response to Severe Nuclear Power Plant Accidents", states that releases of severe magnitude are not possible if plant systems function as designed, and any accident with a release of 20% or greater of the gap region must be considered severe. Using attachment 2 ofEPIP 205H, 10% clad damage is represented by a DAPA reading of l.4E4 R/hr at 0.1 hrs after shutdown (the most conservative). This is shown on the attachment as the 1% TID line. Extrapolating to 20% clad damage gives a reading of 2.8E4 R/hr. EAL- 3.3.3 Rev. 00 Page 1 of 2

HCGS EAL/RALTechnical Basis Exceeding a DAP A reading of 28000 R/hr should meet the criteria for declaration of a General Emergency. DEVIATION None REFERENCES NUMARC NESP-007, PC3 NUREG-1228 - Source Term Estimation During Incident Response to Severe Nuclear Power Plant Accidents EPIP 205H, TSC - Post Accident Core Damage Assessment EAL- 3.3.3 Rev. 00 Page 2 of 2

HCGS EALIRALTechnical Basis 3.0 Fission Product Barriers 3.3 Containment Barrier 3.3.4 RCS LINE BREAK/CONTAINMENT BYPASS 3.3.4.a IC Potential Loss of Containment Barrier = I POINT EAL RCS Line Break OUTSIDE Primary Containment, resulting in a Valid Isolation Signal for ANY one of the following systems:

NSSSS (excluding Main Steam Lines)
HPCI
RCIC NO indication of CONTINUING FLOW I LEAKAGE OUTSIDE the Primary Containment through the eITected system AFTER valve closure from the Main Control Room has been attempted OPERA TI ON AL CONDITION - I, 2, 3 BASIS An RCS Line Break outside Primary Containment that results in a Valid Isolation Signal for any of the systems listed in the EAL requires closure of the associated Primary Containment Isolation valves to mcintain RCS and Primary Containment integrity under abnormal conditions. Successful closure of the required isolation valves that results in NO indication of continuing FLOW I LEAKAGE is classified under this EAL as an Unusual Event, due to the significance of an RCS line break outside the Primary Containment for one of the systems listed in the EAL. An RCS Line is ANY line that communicates directly with the Reactor. A Main Steam Line Break with successful isolation is excluded from this EAL, since it is covered under EAL 3 .2.3 .a. An RCS Line Break with indication of successful isolation is the only condition that warrants classification under this EAL.

Valid is defined as the isolation signal specifically being the result of an RCS Line Break, thus ensuring that the RCS discharge is of significant magnitude to pose a threat to the integrity of the EAL - 3.3.4.a Rev. 00 Page 1 of 3

HCGS EAL/RALTechnical Basis Primary Containment Barrier. This precludes unwarranted Event Classification as the result of condition that result in limited leakage with no potential for "break propagation", including valve packing leaks outside Primary Containment and RWCU Pump Seal Leaks. In addition, isolation signal generated from known failures in other systems, that do not result in Reactor Coolant discharging outside the Primary Containment do not warrant Event Classification under this EAL either. Examples of such failures include a high temperature isolation resulting from a loss of ventilation or cooling water, spurious actuation during I&C surveillance testing or a low Reactor Water Level Condition due to a Loss of High Pressure injection capability. In addition, this EAL ALLOWS for valve closure from the Main Control Room to isolate any systems that did not completely isolate, prior to event classification. Valve closure is defined as the closure of ANY valve from the Main Control Room in the system(s) that did not completely isolate. For example, ifthe isolation logic fails to cause valve closure, but operator actions implemented in the Main Control Room successfully isolates the e:Tected system, then event classification under this EAL is warranted, due to the consequences of the event previously discussed. This includes Motor Operated Valves that are not control by the isolation logic, but are manually controlled from the Main Control Room. EfTected system is defined as the system that is providing the flowpath outside the Primary Containment. In the event the effected system(s) can not be isolated, escalation of the classification will be required. Barrier Analysis Primary Containment Barrier has been potentially lost ESCALATION CRITERIA Emergency Classification will escalate based on the Potential Loss or Loss of additional Fission Product Barriers per EAL Section 3.0. DISCUSSION NSSSS isolations, as well as HPCI and RCIC steam line isolations, are associated with systems that are part of the RCS boundary and penetrate the Primary Containment. Isolation requirements for these lines are covered in I OCFR50, Appendix A, General Design Criteria 55. These systems form a closed loop outside the Primary Containment, and are not open or potentially open to the environment. They are included in this EAL since they represent an extension of the RCS boundary beyond the Primary Containment, and a potential release path from the RCS to the environment. Indication of continuing flow/leakage includes: flow indication through isolated lines, increasing Reactor Building area temperatures, area radiation levels, sump levels, or room levels in spaces associated with affected lines, as well as increases in Plant Vent Eilluent levels. EAL - 3.3.4.a Rev. 00 Page 2 of 3

HCGS EAL/RALTechnical Basis DEVIATION None REFERENCES NUMARC NESP-007, PCS 10 CFRSO, App. A, GDC 55 10 CFR 100 HC.OP-SO.SM-OOOl(Q), Isolation Systems Operation HC.OP-AB.ZZ-Ol 14(Q), Loss of Primary Containment Integrity HC.OP-AB.ZZ-0116(Q), Containment Isolations and Recovery from an Isolation HC.OP-AB.ZZ-0203(Q), Main Steam Line High Radiation HC.OP-AR.SP-OOOl(Q), Radiation Monitoring System Alarm Response HC. OP-AR.ZZ-001 1(Q), Annunciator Response Procedures, Window C6 HC.OP-AR.ZZ-0012(Q), Annunciator Response Procedures, Window C8 HC.OP-EO.ZZ-0100 (Q)-FC, Reactor Scram HC.OP-EO.ZZ-0101 (Q)-FC, Reactor Pressure Vessel (RPV) Control HC.OP-EO.ZZ-0102 (Q)-FC, Primary Containment Control HC.OP-EO.ZZ-0103 (Q)-FC, Reactor Building Control HC.OP-EO.ZZ-0104 (Q)-FC, Radioactive Release Control HCGS Technical Specifications, LCO 3/4.3 HCGS UFSAR, Section 6.2.4.3.1 EAL - 3.3.4.a Rev. 00 Page 3 of 3

HCGS EAL/RALTechnical Basis 3.0 Fission Product Barriers 3.3 Containment Barrier 3.3.4 RCS LINE BREAK/CONTAINMENT BYPASS 3.3.4.b IC Loss of Containment Barrier= 2 POINTS EAL Isolation Signal for ANY one of the following systems:

NSSSS
PCIS
HPCI
RCIC Indication of CONTINUING FLOW I LEAKAGE OUTSIDE the Primary Containment through the effected system AFTER valve closure from the Main Control Room has been attempted OPERATIONAL CONDITION- I, 2, 3 BASIS An Isolation Signal for any of the systems listed in the EAL requires closure of the associated Primary Containment Isolation valves to maintain RCS and Primary Containment integrity under abnormal conditions. A failure of these isolation valves to isolate directly aliows the transport of Reactor Coolant or containment atmosphere to outside the Primary Containment (Containment Breach or Bypass}, resulting in a Loss of Containment.

This EAL addresses two conditions under which RCS is being transported OUTSIDE the Primary Containment. The first condition is associated with an Isolation signal being generated as the result of an RCS Line Break with a failure of the isolation valves to close. In this condition, an ABNORMAL FLOWP ATH exists for RCS to be discharged directly outside the Primary Containment. The second condition is associated with the failure of both Inboard and Outboard EAL - 3.3.4.b Rev.00 Page 1 of J

HCGS EALIRALTechnical Basis Isolation valves to FULLY close following an Isolation signal. In this condition, a flow path from containment atmosphere to areas outside of the Primary Containment exists. In addition, this EAL ALLOWS for valve closure from the Main Control Room to isolate any systems that did not completely isolate, prior to event classification. Valve closure is defined as the closure of ANY valve from the Main Control Room in the system( s) that did not completely isolate. For example, ifthe isolation logic fails to cause valve closure, but operator actions implemented in the Main Control Room successfully isolates the effected system, then Unusual Event declaration is not warranted. This includes Motor Operated Valves that are not control by the isolation logic, but are manually controlled from the Main Control Room. Effected system is defined as the system that is providing the flowpath outside the Primary Containment. Barrier Analysis Primary Containment has been lost. ESCALATION CRITERIA Emergency Classification will escalate based on the Potential Loss or Loss of additional Fission Product Barriers per EAL Section 3.0. DISCUSSION PCIS Isolations are associated with systems having lines that are either: 1) connect directly to the Primary Containment atmosphere and penetrate the Primary Containment; or 2) penetrate the Primary Containment and are neither part of the RCS boundary and are not connected directly to the Primary Containment atmosphere (e.g. RACS, Chilled Water). Isolation requirements for these lines are covered in 10CFR50, App. A , General Design Criteria 56 and 57 respectively. This event, therefore, may potentially connect the RCS or the Primary Containment atmosphere to the environment. Without a completed isolation, continuing flow/leakage represents a release path from the RCS or Primary containment to the environment. NSSSS isolations, as well as HPCI and RCIC steam line isolations, are associated with systems that are part of the RCS boundary and penetrate the Primary Containment. Isolation requirements for these lines are covered in 10CFR50, App. A, General Design Criteria 55. These systems form a closed loop outside the Primary Containment, and are not open or potentially open to the environment. They are included in this EAL since they represent an extension of the RCS boundary beyond the Primary Containment, and a potential release path from the RCS to the environment. Without a completed isolation, continuing leakage represents a Primary System discharging outside the Primary Containment (Containment Bypass), including areas in the Reactor Building addressed in the EOPs. Indication of continuing flow/leakage includes: flow indication through isolated lines, increasing Reactor Building area temperatures, area radiation levels, sump levels, or room levels in spaces associated with affected lines, as well as increases in Plant Vent Effluent levels. EAL - 3.3.4.b Rev.00 Page 2 of 3

HCGS EALIRALTechnical Basis The isolation valve status of all isolation groups is monitored for quick reference on SPDS, to be backed up by operator observation of valve status. DEVIATION NUMARC Primary Containment Barrier Example Flowchart (PC2) suggests that for the "Containment Isolation Valve Status after Containment Isolation Signal" EAL, a failure of both valves in any one line to close AND downstream pathway to the environment exists be included as a threshold for classification of an Unusual Event. In order to include the condition where the Inboard Valve fails to close and an RCS Lirie Break exists between the Primary Containment wall and Outboard Valve, the condition that both valves fail to close is NOT being included in the EAL. Indication of continuing flow I leakage OUTSIDE the Primary Containment will provide an adequate threshold for Event Classification, since both isolation valves must be open for continuing leakage Outside the Primary Containment, except as noted above. REFERENCES NUMARC NESP-007, PC2 10CFR50, App. A, GDC 55, 56, 57 10 CFR 100 HC.OP-SO.SM-OOOI(Q), Isolation Systems Operation HC.OP-AB.ZZ-OI 16(Q), Containment Isolations anJ Recovery from an Isolation HC.OP-AB.ZZ-0203(Q), Main Steam Line High Radiation HC.OP-AR.SP-OOOl(Q), Radiation Monitoring System Alarm Response HC.OP-AR.ZZ-001 l(Q), Annunciator Response Procedures, Window C6 HC.OP-AR.ZZ-0012(Q), Annunciator Response Procedures, Window CS HC.OP-EO.ZZ-0100 (Q)-FC, Reactor Scram HC.OP-EO.ZZ-0101 (Q)-FC, Reactor Pressure Vessel (RPV) Control HC.OP-EO.ZZ-0102 (Q)-FC, Primary Containment Control HC.OP-EO.ZZ-0103 (Q)-FC, Reactor Building Control HCGS Technical Specifications LCO 3/4.3, Instrumentation HCGS UFSAR Sections 6.2.4.3. I, 6.2.4.3.2, 6.2.4.3.3 EAL - 3.3.4.b Rev. 00 Page 3 of 3

HCGS EAL/RALTechnical Basis 3.0 Fission Product Barriers 3.3 Containment Barrier 3.3.5 EMERGENCY COORDINATOR JUDGMENT 3.3.5.a/ 3.3.5.b IC Potential Loss or Loss of Containment Barrier = 2 POINTS EAL ANY condition, in the opinion of the EC, that indicates EITHER a Potential Loss OR Loss of the Containment Barrier OPERATIONAL CONDITION - 1, 2, 3 BASIS This EAL allows the Emergency Coordinator (EC) to address any condition that effects the integrity of the Containment Barrier that is not already covered elsewhere in the Fission Product Barrier Table. A complete loss of the ability to monitor the Containment Barrier should be considered as a "Potential Loss" of that barrier. Barrier Analysis Containment Barrier has been potentially lost or lost. ESCALATION CRITERIA Emergency Classification will escalate based on the Potential Loss or Loss of additional barriers per EAL section 3.0. DISCUSSION None DEVIATION None EAL - 3.3.5.a/ 3.3.5.b Rev. 00 Page 1 of 2

HCGS EAL/RALTechnical Basis REFERENCES NUMARC NESP-007, PC6 EAL - 3.3.5.a/ 3.3.5.b Rev. 00 Page 2 of 2

HCGS EAL/RALTechnical Basis 4.0 EC Discretion

                                                                                  . ::~ u:;::y #

4.1 Emergency Coordinator Discfetion ccS UNUSUAL EVENT - 4.1.1 IC Other Conditions Exist Which In the Judgment of the Emergency Coordinator Warrant Declaration of an Unusual Event EAL Events are in progress or have occurred which, in the judgment of the Emergency Coordinator, indicate a Potential Degradation of Plant Safety OPERATIONAL CONDITION - All BASIS Emergency Coordinator (EC) judgment to declare an Unusual Event, based on the determination that the Potential Degradation of Plant Safety exists, should be implemented ONLY when conditions are not explicitly addressed elsewhere in the ECG. The phrase Potential Degradation of Plant Safety is intended to apply to those conditions that include a likely or actual breakdown of event mitigating actions or that hinder plant personnel from safely operating the plant. The following examples are by no means all inclusive and are not intended to limit the discretion of the SNSS. Examples for consideration include the following:

inadequate emergency response procedures
failure or unavailability of emergency systems during an accident/transient condition
insufficient availability of equipment or support personnel to deal with the ongoing or anticipated events
aircraft crash on or near site
explosions near site (within Owner Controlled Area)

Barrier Analysis Additional guidance on EC judgment for Fission Product Barriers is found on the Fission Product Barrier Table, Section 3.0. ESCALATION CRITERIA EAL - 4.1.1 Rev. 00 Page 1 of 2

HCGS EALIRALTechnical Basis Emergency Coordinator Judgment DISCUSSION Dose consequences from an Unusual Event, if a Radiological Release is involved, would not require offsite response or field monitoring since any release at this level would be < 20 mRem TEDE. Refer ~u Section 6 of the ECG if a Radiological Release is ongoing. DEVIATION None RI:FERENCES NUMARC NESP-007, HUl.3, HUS, Section 3.7. EAL- 4.1.1 Rev. 00 Page 2 of 2

HCGS EALIRALTechnical Basis 4.0 EC Discretion 4.1 Emergency Coordinator Discretion ALERT - 4.1.2 IC Other Conditions Exist Which In the Judgment of the Emergency Coordinator Warrant Declaration of an Alert EAL Events are in progress or have occurred which, in the judgment of the Emergency Coordinator, indicate EITHER one of the following:

Plant safety systems (more than one) are, or may be degraded
ANY Plant Vital Structure is degraded or potentially degraded Increased monitoring of Safety Functions is warranted OPERATIONAL CONDITION - All BASIS Emergency Coordinator (EC) judgment to declare an Alert, based on the determination that Plant Safety Systems are, or may be degraded, should be implemented ONLY when conditions are not explicitly addressed elsewhere in the ECG. This includes a determination by the SNSS that hazards exist that have, or may have caused damage to morJ! than one Safety System or to a Plant Vital Structure.

In addition, if plant conditions degrade to the point where increased monitoring of safety functions is warranted to better determine the plant's actual safety status, then an Alert classification may be appropriate. Barrier Analysis Additional guidance on EC judgment for Fission Product Barriers is found on the Fission Product Barrier Table, Section 3.0. ESCALATION CRITERIA Emergency Coordinator Judgment EAL- 4.1.2 Rev.00 Page 1 of 2

HCGS EAL/RALTechnical Basis DISCUSSION Dose consequences for an Alert, if a Radiological Release was ongoing, would only be a small fraction of the EPA Protective Action Guideline (PAG) plume exposure level, i.e., 10 to 100 mRem TEDE. Refer to ECG Section 6 if a Radiological Release is ongoing. DEVIATION None REFERENCES NUMARC NESP-007, HA6, HAI.4, Section 3.7. EPA-400 EAL - 4.1.2 Rev. 00 Page 2 of 2

HCGS EALIRALTechnical Basis 4.0 EC Discretion 4.1 Emergency Coordinator Discretion SITE AREA EMERGENCY - 4.1.3 IC Other Conditions Exist Which In the Judgment of the Emergency Coordinator Warrant Declaration of a Site Area Emergency EAL Events are in progress or have occurred which, in the judgment of the Emergency Coordinator, indicate an Actual or likely major failure of plant functions needed for the protection of the public OPERATIONAL CONDITION - All BASIS Emergency Coordinator (EC) judgment to decbre a Site Area Emergency, based on the determination that the potential exists for an uncontrolled Radiological Release or the source term available in the Containment atmosphere could result in Site Boundary dose rates in excess of 100 mRem/hr, should be implemented ONLY when conditions are not explicitly addressed elsewhere in the ECG. In addition, any criteria that satisfies the definition of a Site Area Emergency in the ECG Introduction Section, also warrants declaration under this EAL. A Site Area Emergency is intended to be anticipatory of potential fission product barrier failure, and allows offsite agencies to commence preparation for emergency response. Barrier Analysis Additional guidance on EC judgment for Fission Product Barriers is found on the Fission Product Barrier Table, Section 3. ESCALATION CRITERIA Emergency Coordinator Judgment EAL - 4.1.3 Rev. 00 Page 1 of 2

HCGS EAL/RALTechnical Basis DISCUSSION Radiological release rates during a Site Area Emergency declaration are not expected to result in exposure levels which exceed the EPA Protective Action Guideline threshold values except within the Site Boundary. However, plume exposure levels of 100 to < 1000 mRem TEDE may be possible offsite and levels > 1000 mRem TEDE could be experienced onsite. Refer to ECG Section 6 if a Radiological Release is ongoing. DEVIATION None REFERENCES NUMARC NESP-007, HS3, Section 3.7. EPA-400 EAL - 4.1.3 Rev. 00 Page 2 of 2

HCGS EALIRALTechnical Basis 4.0 EC Discretion 4.1 Emergency Coordinator Discretion GENERAL EMERGENCY - 4.1.4 IC Other Conditions Exist Which In the Judgment of the Emergency Coordinator Warrant Declaration of a General Emergency EAL Events are in progress or have occurred which, in the judgment of the Emergency Coordinator, indicate an Actual or imminent substantial core degradation with the potential for loss of containment OPERA TI ON AL CONDITION - All BASIS Emergency Coordinator (EC) judgment to declare a General Emergency , based on the determination that the potential for an uncontrolled Radiological Release exists, should be implemented ONLY when conditions are not explicitly addressed elsewhere in the ECG. In addition, any criteria that satisfies the definition of a General Emergency in the ECG Introduction Section, also warrants declaration under this EAL. A General Emergency is intended to be anticipatory of fission product barrier failure, and permits maximum offsite intervention time. Barrier Analysis Additional guidance on EC judgment for Fission Product Barriers is found on the Fission Product Barrier Table, Section 3.0. ESCALATION CRITERIA NIA DISCUSSION Radiological Release rates during a General Emergency may exceed the EPA Protective Action Guidelines, i.e., >1000 mRem TEDE, for more than the immediate site area. ECG Section 6, Radiological Releases/Occurrences should be consulted for releases of this magnitude. EAL- 4.1.4 Rev. 00 Page 1 of 2

HCGS EAL/RALTechnical Basis DEVIATION None REFERENCES NUMARC NESP-007, HG2, Section 3.7. EPA-400 EAL- 4.1.4 Rev. 00 Page 2 of 2

HCGS EAL/RALTecJmical..Basis

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* ~

5.0 Failure to Scram 5.1 ATWS ALERT- 5.1.2.a I 5.1.2.b IC Failure of the Reactor Protection System (RPS) to Successfully Complete a Reactor Scram (Automatic or Manual) EAL EITHER one of the following conditions:

An Automatic Reactor Scram Condition exists AND An Automatic Reactor Scram (RPS)

IS NOT successful

ANY Manually Initiated Reactor Scram (RPS) from the Control Room IS NOT successful OPERATIONAL CONDITION - I, 2 BASIS Failure of the RPS to successfully complete a Reactor Scram (automatic or manual) represents a significant degradation in plant safety, as the primary reactivity control system has failed to perform its design function. The intent of this EAL is to classify events in which either an automatic or manual RPS signal fails to initially complete a successful scram when required, even if a subsequent manual or automatic scram is successful . The failure of RPS to complete a successful scram, is the bases for Alert declaration under this EAL. A Successful scram (RPS automatic or RPS manual), as it relates to this EAL, results in a Control Rod configuration by which the Reactor will remain shutdown under all conditions without boron. The three criteria that satisfy this condition are :
All Control Rods are inserted to position 02 or beyond (Maximum Subcritical Banked Withdrawal Position)
All Control Rods but one being full inserted.
Reactor Engineering has determined that the Reactor will remain Shutdown under all conditions without Boron In addition, for a manual scram to be considered successful, it must be attempted from the Reactor Control console. In the event that ARI completes a successful Scram following a failure of automatic or manual RPS, the declaration of an Alert is still warranted, due to the failure of EAL - 5.1.2.a I 5.1.2.b Rev. 00 Page 1 of 3

HCGS EAL/RALTechnical Basis RPS. An inability to physically place the Reactor Mode Switch in the SHUTDOWN position, (i.e. broken key) does not constitute an RPS failure, since the RPS logic has not failed. Barrier Analysis This event does not reach the threshold for the loss of Fuel Clad or RCS Barriers, but conditions exist that could lead to a potential loss of those barriers. ESCALATION CRITERIA Emergency Classification will escalate to a Site Area Emergency ( 5 .1. 3) when a failure of both automatic or manual scram functions occurs, with Reactor power remaining ~4%. DISCUSSION The Reactor Protection System (RPS) is designed to function to shut down the reactor (either manually or automatically). The system is "fail safe", that is it deenergizes to function. An Anticipated Transient Without Scram (ATWS) event can be caused either by a failure of RPS (electrical/pneumatic failure) or a failure of the Control Rod Drive system to pennit the control rods to insert (hydraulic failure). The Alternate Rod Insertion (ARI) function of the Redundant Reactivity Control System (RRCS) provides an automatic backup function for an electrical/pneumatic failure of the RPS. A successful scram due to ARI following a failure of the RPS would still be classified under this EAL because of the potentially serious consequences of an RPS failure. Confinnation indications of an RPS failure to complete a successful scram include control room annunciators, control rod positions, APRM power and downscale indicating lights, IRM/SRM power level, SRM period, and control rod position indication. A manual scram is defined as any set of actions by the reactor operator( s) at the reactor controls which causes control rods to be rapidly inserted into the core via the RPS in an attempt to place the reactor in a subcritical condition (i.e. mode switch to shutdown, manual scram push buttons). This EAL addresses only those manual scram attempts that are initiated from the Control Room control panels. A failure of the RPS to initiate and complete a reactor scram can result iri the design limits of the nuclear fuel being compromised. RPS is designed to automatically detect and generate a reactor scram signal when a Technical Specification Limiting Safety System Setting (LSSS) is reached or exceeded. If an LSSS is exceeded without an automatic scram, consideration must be given to the possibility that a Technical Specification Safety Limit may have been exceeded. EAL - 5.1.2.a I 5.1.2.b Rev.00 Page 2 of 3

HCGS EALIRALTechnical Basis DEVIATION NUMARC EAL SA2 suggests that an Alert classification be based only on a failure of an automatic RPS scram followed by a successful manual RPS scram from the control room, with EAL SS2 escalating to a Site Area Emergency if a manual scram (RPS or ARI) fails to reduce Reactor Power below 4%. The Alert threshold is set so that unsuccessful manual RPS scrams from the control room, as well as unsuccessful automatic RPS scrams via RPS would be classified at the Alert level. This will cover those situations in which a manual RPS scram is attempted in anticipation of a continually degrading plant condition (i.e. degrading Main Condenser Vacuum). In addition, this threshold will also address those situations where a manual scram is required by procedure. (i.e. stuck open SRV, Main Steam Line Hi Hi Radiation, Dual Reactor Recirc Pump trip, Power Oscillations) and the manual scram is not successful. In either case, Alert declaration is appropriate when the RPS fails to perform its intended design function. The SAE threshold is set to include automatic and manual failure (for the reasons stated above), with resulting power 2:4% as suggested in NUMARC EAL SS2 bases. By defining a "Successful" scram as control rod being positioned such that the Reactor will remain Shutdown under all conditions, partial scrams that result in Reactor Power below 4% would be classified as an Alert, whether automatically or manually initiated. REFERENCES NUMARC NESP-007, SA2 NUMARC Questions and Answers, June 1993, "System Malfunctions Question #7" HC.OP-EO.ZZ-0100 (Q)-FC, Reactor Scram HC.OP-EO.ZZ-0101 (Q)-FC, Reactor Pressure Vessel (RPV) Control HC.OP-EO.ZZ-0207 (Q)-FC, Level/Power Control BWR Owners Group Emergency Procedure Guidelines, Revision 4 HCGS Technical Specifications 1.0, Definitions; SL/LSSS 2.1/2.2; LCO 3/4.1, Reactivity Control Systems; LCO 3/4.3, Instrumentation EAL - 5. l.2.a I 5.1.2.b Rev.00 Page 3 of 3

HCGS EALIRALTechnical Basis 5.0 Failure to Scram 5.1 ATWS SITE AREA EMERGENCY-5.1.3 IC Failure of the Reactor Protection System (RPS) to Successfully Complete a Reactor Scram (Automatic and Manual) and Reactor Power is above 4%. EAL EITHER one of the following conditions:

An Automatic Reactor Scram Condition exists AND An Automatic Reactor Scram (RPS)

IS NOT successful

ANY Manually Initiated Reactor Scram (RPS) from the Control Room IS NOT successful ALL Reactor Scram attempts from the Control Room (RPS and ARI)

DID NOT REDUCE and MAINTAIN Reactor Power to~ 4% OPERATIONAL CONDITION - I, 2 BASIS Failure of the RPS to successfully complete a Reactor Scram (automatic and manual) represents a significant degradation in plant safety, as the primary reactivity control system has failed to perform its design function. In addition, failure of subsequent Reactor Scram attempts (both RPS and ARI) to reduce Reactor Power to less than 4%, represents a potential challenge to the ability to provide continued heat removal from the Reactor. Thus, conditions exist that could lead to an imminent Joss or potential loss of both the Fuel Clad and RCS Barriers. The intent of this EAL is to classify events in which both automatic and manual RPS signals fail to complete a successful scram when required, and subsequent actions using ARI fail to reduce

Reactor Power to less than 4%. The failure of RPS and ARI to complete a successful scram with Reactor Power remaining above 4% is the bases for SAE declaration under this EAL. A Successful scram (RPS Automatic or Manual), as it relates to this EAL, results in a Control Rod configuration by which the Reactor will remain shutdown under all conditions without boron injection.

EAL - 5.1.3 Rev.00 Page 1 of 3

HCGS EALIRALTechnical Basis The three criteria that satisfy this condition are :

All Control Rods are inserted to position 02 or beyond (Maximum Subcritical Banked Withdrawal Position)
All Control Rods but one being full inserted.
Reactor Engineering has determined that the Reactor will remain Shutdown under all conditions without Boron In addition, for a manual scram to be considered successful, it must be attempted from the Reactor control console. In the event that ARI completes a successful Scram following a failure of automatic or manual RPS, the declaration of an SAE is not warranted.

Barrier Analysis This event does not reach the threshold for the loss of Fuel Clad or RCS Barriers, but conditions exist that could lead to an imminent loss or potential loss of those barriers. ESCALATION CRITERIA Emergency Classification will escalate to a General Emergency (5.1.4) when Reactor Water Level can not be maintained >-190", or Suppression Pool Temperature and Reactor Pressure can not be maintained below the HCTL. DISCUSSION The Reactor Protection System (RPS) is designed to function to shut down the reactor (either manually or automatically). The system is "fail safe", that is it deenergizes to function. An Anticipated Transient Without Scram (ATWS) event can be caused either by a failure of RPS (electrical/pneumatic failure) or a failure of the Control Rod Drive system to permit the control rods to insert (hydraulic failure). The Alternate Rod Insertion (ARI) function of the Redundant Reactivity Control System (RRCS) provides an automatic backup function for an electrical/pneumatic failure of the RPS. A failure of ARI to reduce Reactor Power to .:S 4% following a failure of the RPS is classified under this EAL because of the potentially serious consequences of a failure of RPS and ARI to reduce Reactor Power. Confirmation indications of an RPS failure to complete a successful scram include control room annunciators, control rod positions, APRM power and downscale indicating lights, IRM/SRM power level, SRM period, and control rod position indication. A failure of the RPS to initiate and complete a reactor scram can result in the design limits of the nuclear fuel being compromised. RPS is designed to automatically detect and generate a reactor scram signal when a Technical Specification Limiting Safety System Setting (LSSS) is reached or EAL - 5.1.3 Rev. 00 Page 2 of 3

HCGS EAL/RALTechnical Basis exceeded. If an LSSS is exceeded without an automatic scram, consideration must be given to the possibility that a Technical Specification Safety Limit may have been exceeded. Emergency Operating Procedures (EOPs) establish Reactor Power> 4% coincident with a scram condition as the initiating condition for various actions in response to an ATWS. If the Reactor is isolated (MSIVs closed), the heat generated is transferred to the Primary Containment, thus potentially threatening the integrity of Primary Containment. In an attempt to preclude this condition, EOP guidance includes restoration of the Main Condenser as a heat sink, provided there is no indication of gross fuel failure or a Main Steam Line break. EOP guidance also includes methods of alternate reactivity control, including the use of Standby Liquid Control (SLC), alternate control rod insertion, and intentional lowering of Reactor Water Level to control Reactor Power. DEVIATION None REFERENCES NUMARC NESP-007, SS2 NUMARC Questions and Answers, June 1993, "System Malfunctions Question #7" HC.OP-EO.ZZ-0 I 00 (Q)-FC, Reactor Scram HC.OP-EO.ZZ-0101 (Q)-FC, Reactor Pressure Vessel (RPV) Control HC.OP-EO.ZZ-0207 (Q)-FC, Level/Power Control BWR Owners Group Emergency Procedure Guidelines, Revision 4 HCGS Technical Specifications 1.0, Definitions; SL/LSSS 2.1/2.2; LCO 3/4.1, Reactivity Control Systems; LCO 3/4.3, Instrumentation EAL - 5.1.3 Rev. 00 Page 3 of 3

HCGS EAL/RALTechnical Basis 5.0 Failure to Scram 5.1 ATWS GENERAL EMERGENCY - 5.1.4 IC Failure of the Reactor Protection System (RPS) to Successfully C0mplete a Reactor Scram (Automatic and Manual) and there is indication of an Extreme Challenge to the Ability to Cool the Core EAL EITHER one of the following conditions:

An Automatic Reactor Scram Condition exists AND An Automatic Reactor Scram (RPS)

IS NOT successful

ANY Manually Initiated Reactor Scram (RPS) from the Control Room IS NOT successful ALL Reactor Scram attempts from the Control Room (RPS and ARI)

DID NOT REDUCE and MAINTAIN Reactor Power to~ 4% EITHER one of the following:

Reactor Water Level CANNOT BE MAINTAINED> -190"
The combination of Suppression Pool Temperature and RPV Pressure CANNOT BE MAINTAINED below the HCTL Curve OPERATIONAL CONDITION - I, 2 BASIS Failure of the RPS to successfully complete a Reactor Scram (Automatic and Manual) represents a significant degradation in plant safety, as the primary reactivity control system has failed to perform its design function. In addition, failure of subsequent scram attempts (ARI) to reduce EAL - 5.1.4 Rev.00 Page 1 of 4

HCGS EAL/RALTechnical Basis Reactor Power to less than 4%, resulting in an inability to MAINTAIN Reactor Water Level above -190" or Suppression Pool (SP) Temperature and Reactor Pressure below the Heat Capacity Temperature Limit (HCTL), represents an imminent loss or potential loss of all three fission product barriers. The inability to MAINTAIN Reactor Water Level above -190" was chosen based on the condition that core cooling is extremely challenged. This threshold corresponds directly to a decision step contained in EOP 207, Level /Power Control (Step LP-18), which requires a determination be made if Reactor Water Level can be MAINTAINED above -190". For cases where Reactor Water Level CAN NOT BE MAINTAINED> -190", a General Emergency declaration is warranted. The intent of this EAL is to classify those ATWS events that result in a challenge to the integrity of these barriers. A Successful scram (RPS Automatic and Manual), as it relates to this EAL, results in a Control Rod configuration by which the Reactor will remain shutdown under all conditions without boron. The three criteria that satisfy this condition are :

All Control Rods are inserted to position 02 or beyond (Maximum Subcritical Banked Withdrawal Position)
All Control Rods but one being full inserted.
Reactor Engineering has determined that the Reactor will remain Shutdown under all conditions without Boron Barrier Analysis This event reaches the threshold for the imminent loss of all three Fission Product Barriers.

ESCALATION CRITERIA NIA DISCUSSION The Reactor Protection System (RPS) is designed to function to shut down the reactor (either manually or automatically). The system is "fail safe", in that it deenergizes to function. An Anticipated Transient Without Scram (A TWS) event can be caused either by a failure of RPS (electrical/pneumatic failure) or a failure of the Control Rod Drive (CRD) system to permit the control rods to insert (hydraulic failure). The Alternate Rod Insertion (ARI) function of the Redundant Reactivity Control System (RRCS) provides an automatic backup function for an electrical/pneumatic failure of the RPS. Confirmation indications of an RPS failure to complete a successful scram include control room annunciators, control rod positions, APRM power and downscale indicating lights, IRM/SRM power level, SRM period, and control rod positior. indication. EAL - 5.1.4 Rev.00 Page 2 of 4

HCGS EAL/RALTechnical Basis A failure of the RPS to initiate and complete a reactor scram can result in the design limits of the nuclear fuel being compromised. RPS is designed to automatically detect and generate a reactor scram signal when a Technical Specification Limiting Safety System Setting (LSSS) is reached or exceeded. If an LSSS is exceeded without an automatic scram, consideration must be given to the possibility that a Technical Specification Safety Limit may have been exceeded. Emergency Operating Procedures (EOPs) establish Reactor Power> 4% coincident with a scram condition as the initiating condition for various actions in response to an ATWS. If the Reactor is isolated (MSIVs closed), the heat generated is transferred to the Primary Containment, thus potentially threatening the integrity of Primary Containment. In an attempt to preclude this condition, EOP guidance includes restoration of the Main Condenser as a heat sink, provided there is no indication of gross fuel failure or a main steam line break. EOP guidance also includes methods of alternate reactivity control, including the use of Standby Liquid Control (SLC), alternate control rod insertion, and intentional lowering of Reactor Water Level to control Reactor Power. During these actions, adequate core cooling is accomplished by maintaining Reactor Water Level above -190". Although this is below the Top of Active Fuel (Loss of Core Submergence), maintaining Reactor Water Level above -190" will ensure sufficient steam flow from the covered portion of the core to preclude Fuel Clad Temperatures in the uncovered portion of the core from exceeding 1500 Degrees F. This is referred to as the Minimum Steam Cooling RPV Water Level. Inability to maintain this level may result in damage to the fuel. The EOPs require the initiation of SLC before SP Temperature reaches 110 Degrees F. This threshold is referred to as the Boron Injection Initiation Temperature, and is defined as the highest SP Temperature at which initiation of boron injection will result in injection of the Hot Shutdown Boron Weight before SP Temperature exceeds the HCTL. Actions required by the EOPs when Reactor Water Level can not be maintained above -190" or the HCTL is exceeded include the initiation of Emergency Depressurization. DEVIATION None EAL - 5.1.4 Rev. 00 Page 3 of 4

HCGS EAL/RALTechnical Basis REFERENCES NUMARC NESP-007, SG2 NUMARC Questions and Answers, June 1993, "System Malfunctions Question #7" HC.OP-EO.ZZ-0100 (Q)-FC, Reactor Scram HC.OP-EO.ZZ-0101 (Q)-FC, Reactor Pressure Vessel (RPV) Control HC.OP-EO.ZZ-0207 (Q)-FC, Level/Power Control BWR Owners Group Emergency Procedure Guidelines, Revision 4 HCGS Technical Specifications 1.0, Definitions; SL/LSSS 2.1/2.2; LCO 3/4.1, Reactivity Control Systems; LCO 3/4.3, Instrumentation EAL - S. l.4 Rev. 00 Page 4 of 4

HCGS

                                                                      ...... .. .. EAL/RALTechnical
                                                                                   .                     Basis
                                                                                        '"":'"'.'J'."-~~7'.""j 6.0 Radiological Releases/OccurrFnc.,; ..

c3'"r r rr ! I 6.1 Gaseous Effluent Releas~

                                                                     !                                ~--J UNUSUAL EVENT- 6.1.1.a IC      Any Unplanned Release of Gaseous Radioactivity to the Environment that Exceeds 2 Times the Radiological Technical Specifications for 60 minutes or longer EAL Dose Assessment indicates EITHER one of the following at the MEA or beyond as calculated on the SSCL:

TEDE 4-Day Dose of~ 2.0E-01 mRem
Thyroid-COE Dose of~ 6.SE-01 mRem based on Plant Vent effluent sample analysis and NOT on a default Noble Gas to Iodine Ratio Release is ongoing for~ 60 minutes OPERATIONAL CONDITION - All BASIS Dose Assessment at or beyond the MEA exceeding the EAL threshold, can result from a Gaseous Radiological Release in excess of 2 times Technical Specifications. This condition results from an uncontrolled release of radioactivity to the environment, resulting in elevated offsite dose rates.

The threshold for this EAL is NOT based on a specific offsite dose rate, but rather on the loss of plant control implied by a radiological release of this magnitude that was not isolated within 60 minutes. The final integrated dose is very low and is not the primary concern. Classification is based on an ongoing release that does not comply with a license condition. Unplanned is defined as any release for which a radioactive discharge permit was not prepared, or a release that exceeds the conditions on the applicable permit. Dose Assessment using actual meteorological data provides an accurate indication of release magnitude. The use of dose assessment based EALs is therefore preferred over the use of Release Rate based EALs which utilize calculations which have built-in inaccuracies because ODCM default meteorological data is used. As long as dose assessment is available, this EAL should be used in place of EAL 6.1.1.d. EAL - 6.1.1.a Rev. 00 Page 1 of 3

HCGS EAL/RALTechnical Basis It is not intended that the release be averaged over 60 minutes, but exceed 2 times the Technical Specification limit for 60 minutes or longer. In addition, it is intended that the event be declared as soon as it is determined that the release will exceed 2 times the limit for 60 minutes or longer. Barrier Analysis NIA ESCALATION CRITERIA Emergency Classification will escalate to an Alert when the effluent release concentration increases to 200 times the Technical Specification limit. DISCUSSION Prorating the 500 mRem/yr criterion for the TEDE 4-day dose: time (8766 hr/yr); the 2 x Tech. Spec. multiplier; and, Artificial Island's Allocation Factor of 0.5 (50% per site), the associated site boundary dose rate would be 0.057 mRem/hr. 500mRem/ yr TEDE 4-Day MEA Dose Rate= ( )(2)(.5) = 0.057 mRem/hr 8766hr I yr This is rounded to .05 mRem/hr. The TEDE 4-day Dose is based on a 4 hour release duration. Therefore .05 mRem/hr

4 hours=

0.2 mRem. Prorating the 1500 mRem/yr criterion for the Thyroid-COE Dose: time (8766 hr/yr); the 2 x Tech. Spec. multiplier; and, Artificial Island's Allocation Factor of 0.5 (50% per site), the associated site boundary dose rate would be 0.17 mRem/hr.

       .                                1500mRem/ yr Thyr01d-CDE MEA Dose Rate= (                             )(2)(.5) = 0.17 mRem/hr 8766hr I yr The Thyroid-COE Dose is based on a 4 hour release duration. Therefore 0.17 mRem/hr            *4 hours= 0.68 mRem.

DEVIATION None EAL - 6.1.1.a Rev.00 Page 2 of 3

HCGS EAL/RALTechnical Basis REFERENCES NUMARC NESP-007, AUi .4 Off-Site Dose Calculation Manual, Section 2.0 - Gaseous Efiluents NUMARC Draft White Paper, 7-25-94, 9-10-94. Technical Specification 3.11.2. l EAL - 6.1.1.a Rev. 00 Page 3 of 3

HCGS EAL/RALTechnical Basis 6.0 Radiological Releases/Occurrences 6.1 Gaseous Effluent Release UNUSUAL EVENT- 6.1.1.b IC Any Unplanned Release of Gaseous Radioactivity to the Environment that Exi:eeds 2 Times the Radiological Technical Specifications for 60 minutes or longer EAL Dose Rate measured at the Protected Area Boundary or lieyond EXCEEDS

    .05 mRem/hr above nonnal background Release is ongoing for   ~  60 minutes OPERATIONAL CONDITION - All BASIS Measured Dose Rate at or beyond the Protected Area Boundary exceeding the EAL threshold can result from a Gaseous Radiological Release in excess of 2 times Technical Specifications. This condition results from an uncontrolled release of radioactivity to the environment, resulting in elevated offsite dose rates. The threshold for this EAL is NOT based on a specific offsite dose rate, but rather on the loss of plant control implied by a radiological releasP. of this magnitude that was not isolated within 60 minutes. The final integrated dose is very low and is not the primary concern. Classification is based on an ongoing release that does not comply with a license condition. Unplanned is defined as any release for which a radioactive discharge permit was not prepared, or a release that exceeds the conditions on the applicable pennit.

It is not intended that the release be averaged over 60 minutes, but exceed 2 times Tech. Spec. limits for 60 minutes or longer. Further, it is intended that the event be declared as soon as it is determined that the release will exceed 2 times the limit for 60 minutes or longer. Barrier Analysis NIA EAL - 6.1.1.b Rev.00 Page 1 of 2

HCGS EAL!RALTechnical Basis ESCALATION CRITERIA Emergency Classification will escalate to an Alert when effluent release concentration increases to 200 times the Technical Specification limit. DISCUSSION Prorating the 500 mRem/y1 criterion for: time (8766 hr/yr); the 2 x Tech. Spec. multiplier; and, Artificial Island's Allocation Factor of 0.5 (50% per site), the associated site boundary (MEA) dose rate would be 0.057 mRem/hr. 500m Rem I yr

Protected Area Boundary Dose Rate= ( )(2)(.5) = 0.57 mRem/hr 8766hr I yr This is rounded to .05 mRem/hr DEVIATION None REFERENCES NUMARC NESP-007, AUl.3 Off-Site Dose Calculation Manual, Section 2.0 - Gaseous Effluents NUMARC Draft White Paper, 7-25-94, 9-10-94.

Technical Specification 3 .11.2.1 EAL - 6.1.1.b Rev. 00 Page 2 of 2

HCGS EAL/RALTechnical Basis 6.0 Radiological Releases/Occurrences 6.1 Gaseous Effluent Release UNUSUAL EVENT - 6.1.1.c IC Any Unplanned Release of Gaseous Radioactivity to the Environment that Exceeds 2 Times the l OCFR20, Appendix B limits for 60 minutes or longer EAL Gaseous effluent release sample analysis for ANY one of the following indicates a concentration of:

FRVS:

          ~ l.13E-03 µCi/cc Total Noble Gas
           ~ 2.71E-07 µCi/cc 1-131

NPV:

          ~ 2.43E-04 µCi/cc Total Noble Gas
          ~ 5.SlE-08 µCi/cc 1-131

SPV:

          ~  2.27E-05 µCi/cc Total Noble Gas
            ~ 5.44E-09 µCi/cc 1-131 Dose Assessment results NOT available Release is ongoing  for~ 60 minutes OPERATIONAL CONDITION - All BASIS Gaseous eflluent release sample analysis exceeding the EAL threshold for any of the plant vents listed (FRVS, NPV, SPY), can result from a Gaseous Radiological Release in excess of2 times l OCFR20, Appendix B limits. This condition results from an uncontrolled release of radioactivity to the environment, resulting in elevated offsite dose rates.

EAL - 6.1. l.c Rev. 00 Page 1 of 3

HCGS EALIRALTechnical Basis The threshold for this EAL is NOT based on a specific offsite dose rate, but rather on the loss of plant control implied by a radiological release of this magnitude that was not isolated within 60 minutes. The final integrated dose is very low and is not the primary concern. Classification is based on an ongoing release that does not comply with a license condition. The HTV is not included under this EAL since there are no provisions for collecting a HTV grab sample. Unplanned is defined as any release for which a radioactive discharge permit was not prepared, or a release that exceeds the conditions on the applicable permit. It is not intended that the release be averaged over 60 minutes, but exceed 2 times the I OCFR20, Appendix B limit for 60 minutes or longer. In addition, it is intended that the event be declared as soon as it is determined that the release will exceed 2 times the limit for 60 minutes or longer. Barrier Analysis NIA ESCALATION CRITERIA Emergency Classification will escalate to an Alert when the effluent release concentration increases to 200 times the 10CFR20, Appendix B limit. ~ DISCUSSION Refer to Basis Section for EAL 6.1.1.d for the 10CFR20, Appendix B Noble Gas and Thyroid Committed Dose release rate calculations. Calculation of the threshold sample concentrations are as follows:

                                       .      4.80£ + 03µCi I sec                 .

FRVS Noble Gas Sample Co11ce11trat1011 = = l .1 JE-03 µCi/cc 472x9000cfin 5 1 FRVS 1-131 Sample Co11ce11tration = 1.1 E + OO µCi sec -- 2. 71 E-07 µCi/cc 472x9000cfm 4.80£ + 03µCi I sec NPV Noble Gas Sample Co11cel/fratio11= = 2.43E-04 µCi/cc 472x4.19E + 04cjin NPV 1-131 Sample Co11ce11tratio11 = l.1 5E + OOµCil sec -- 5.81E-08 µCi/cc 472x4.19E + 04cfm 4.80£ + 03µCi I sec SPV Noble Gas Sample Co11ce11tralio11 = = 2.27E-05 µCi/cc 472x4.48£ + 05cfm EAL - 6.1.1.c Rev.00 Page 2 of 3

HCGS EAL/RALTechnical Basis 1.15£ +OOµCil sec . SPV I-131 Sample Concentration= = 5.44E-09 µCi/cc 472x4.48E + 05cjm Where: 472 =conversion factor (28,317 cc/ft3 x 1 min./60 sec.) 9000 cfm = FRVS Vent Flow (maximum) 4.19E+04 cfm = NPV Vent Flow (maximum) 4.48E+05 cfm = SPV Vent Flow (maximum) The noble gas release rate of 4.80E+03 µCi/sec is obtained by multiplying the IOCFR20, Appendix B limit release rate of 2.40E+03 µCi/sec times 2. The iodine release rate of 1.15E+OO µCi/sec is obtained by multiplying the IOCFR20, Appendix B limit release rate of 5.75E-01 µCi/sec times 2. DEVIATION The value for EAL 6.1.1.c is based on one meteorological case and one isotopic mixture found in the ODCM. A radiological release based on this specific release rate could produce a TEDE Dose which would require an Alert classification or not meet the Unusual Event classification, depending on the meteorological conditions and the isotopic mixture. EAL 6.1.1.c would not be used unless EAL 6.1.1.a (Dose Assessment) can not be used to determine the classification, if any, due to the potential uncertainty of this "default" EAL. Two times the 10CFR20, Appendix B limits for noble gas and Iodine 131 are being used for this EAL, due to concerns that the State of New Jersey have pertaining to this EAL and based on the above mentioned uncertainties. REFERENCES NUMARC NESP-007, AUI.2, AUI.1, AUl.4 Off-Site Dose Calculation Manual, Section 2.0 NUMARC Draft White Paper, 7-25-94, 9-10-94. Technical Specification 3.11.2.1 EAL - 6.1.1.c Rev.00 Page 3 of 3

HCGS EAL/RALTechnical Basis 6.0 Radiological Releases/Occurrences 6.1 Gaseous Effluent Release UNUSUAL EVENT- 6.1.1.d IC Any Unplanned Release of Gaseous Radioactivity to the Environment that Exceeds 2 times. the I OCFR20, Appendix B limits for 60 minutes or longer EAL Valid High Alarm received from ANY one of the following Plant Effluent RMS Channels:

FRVS Noble Gas (Grid 1/3; 9RX680)
NPV Noble Gas (Grid 1/3; 9RX590)
NPV Iodine (Grid 3; 9RX601)
SPV Noble Gas (Grid 1/3; 9RX580)
SPV Iodine (Grid 3; 9RX605)
HTV Noble Gas (Grid 3; 9RX516)

Total Plant Vent release rate EXCEEDS EITHER one of the following limits:

4.80E+03 µCi/sec Total Noble Gas
l. l 5E+OO µCi/sec 1-131 (NPV & SPV ONLY)

Dose Assessment results NOT available Release is ongoing for ~ 60 minutes OPERATIONAL CONDITION - All BASIS Valid High alarm and effluent release rate values exceeding the EAL threshold, can result from a Gaseous Radiological Release in excess of2 times 10CFR20, Appendix B limits. This condition results from an uncontrolled release of radioactivity to the environment, resulting in elevated offsite dose rates. The threshold for this EAL is NOT based on a specific offsite dose rate, but rather on the loss of plant control implied by a radiological release of this magnitude that was not EAL - 6.1. l.d Rev. 00 Page 1 of 4

HCGS EAL/RALTechnical Basis isolated within 60 minutes. The final integrated dose is very low and is not the primary concern. Valid is defined as the High alarm actuating specifically due to a Gaseous Release exceeding 10 CFR 20, Appendix B limits, thus precluding unwarranted event declaration as the result of spurious actuation. Classification is based on an ongoing release that does not comply with a license condition. Unplanned is defined as any release for which a radioactive discharge permit was not prepared, or a release that exceeds the conditions on the applicable permit. The EAL value for Total Plant Vent release rate was determined using default X/Q values from the ODCM which provides a less accurate method of evaluation release magnitude than using dose assessment with real time meteorological data. For that reason, this EAL should not be utilized if Dose Assessment is available. Dose Assessment will take in account actual meteorological conditions, plant vent flows and plant vent effluent concentrations to provide a more accurate assessment of a radiological release. If Dose Assessment is available then refer to EAL 6.1.1.a for classification. It is not intended that the release be averaged over 60 minutes, but exceed 2 times 10 CFR20, Appendix B limits for 60 minutes or longer. In addition, it is intended that the event be declared as soon as it is determined that the release will exceed 2 times the limit for 60 minutes or longer. Barrier Analysis NIA ESCALATION CRITERIA Emergency Classification will be escalated to an Alert when the effluent release concentration increases to 200 times the I OCFR20, Appendix B limits. DISCUSSION The release rate thresholds for this EAL are obtained by multiplying the Technical Specification release rates of2.4E+03 µCi/sec and 5.78E-01 µCi/sec, for Noble Gases and Iodine-131 respectively, times 2. Total Noble Gas release rate is the summation of all plant vent release rates. This EAL includes Iodine Release Rates for the NPV and SPV, since these vents have an Iodine monitor. Determination of the Iodine Release Rate from the Iodine monitor is accomplished by multiplying the Iodine reading (in uCi/cc) by the applicable vent flow rate, and 472 (Conversation factor). Iodine Release rates for FRVS and the HTV are excluded since these vents do not include an Iodine detector. The SPDS Total Iodine Offsite Release Rate does not provide useful information because this is based on a default value of 1000 times less than the Total Noble Gas Offsite Release Rate, which could be grossly inaccurate. Release rates for FR VS and the HTV are not included since these vents do not have an Iodine detector. A gaseous effluent sample is needed to accurately quantify the Iodine Release rate. The SPDS Total Iodine Offsite Release Rate should not be used, as this is based on a default value of EAL - 6.1.1.d Rev. 00 Page 2 of 4

HCGS EAL/RALTechnical Basis 1000 times less than the Total Noble Gas Offsite Release Rate. The 10CFR20, Appendix B limits are based on ODCM calculations. 10CFR20, Appendix B Calculation for Noble Gas (100 mRem I year)* (Allocation Factor) u Ci/s econ d = (ODCM X /Q)*(ODCM DRCF). WHERE: uCi/Second = Total Noble Gas Release Rate from Salem (Unit 1 & Unit 2) or Hope Creek (all Vents; NPV, SPV, FRVS, and HTV) which would result in a TEDE Dose Rate of 50 mRem/year. ODCM X/Q = Site Specific (Salem or Hope Creek) dispersion factor at the Site Boundary in sec/m3. ODCM DRCF = Site Specific (Salem or Hope Creek) dose rate conversion factor in mRem/year/uCifm3. ODCM X/Q = 2.67E-06 sec/m 3 ODCM DRCF = 7.80E+03 mRem/yr/uCi/m3 Allocation Factor= 5.00E-01 2.40E+03 uCi/Second = (I 00 m Rem I yr)* (5.00E - OI) (2.67 E - 06 sec/ m3 ) * (7.80£ + 03mRem I yr I µCi I m3 ) 2.40E+03 µCi/sec *2 =EAL value. 4.80E+03 µCi/sec is the EAL value. 10CFR20, Appendix B Calculation for Thyroid Committed Dose uCi/Second = 50 mRem/year * (Allocation Factor) (ODCM X/Q) * (ODCM THY DRCF) WHERE: uCi/Second = Total Iodine 131 release rate from Salem (Unit I or 2) or Hope Creek (all Vents; NPV, SPY, FRVS and HTV). ODCM X/Q = Site Specific (Salem or Hope Creek) dispersion factor at the Site Boundary in sec/m3. EAL - 6.1.1.d Rev.00 Page 3 of 4 L___ _

HCGS EAL/RALTechnical Basis ODCM THY DRCF = is the most limiting potential pathway (inhalation, child, thyroid I-131) dose rate conversion factor in mRem/year/uCi/m3. ODCM X/Q = 2.67E-06 sec/m3 ODCM THY DRCF = 1.62E+07 mRem/yr/uCi/m3 Allocation Factor= 5.00E-01 5.78E-Ol uCi/Second = (50 mRem/year) * (5.00E-01) (2.67E-06 sec/m3)* (1.62E+07 mRem/yr/uCi/m3) 5.78E-Ol µCi/sec* 2 =EAL value. 1.15E+OO µCi/sec is the EAL value. DEVIATION The value for EAL 6.1.1.d is based on one meteorological case and one isotopic mixture found in the ODCM. A radiological release based on this specific release rate could produce a TEDE Dose which would require an Alert classification or not meet the Unusual Event classification, depending on the meteorological conditions and the isotopic mixture. EAL 6.1.1.d would l)Ot be used unless EAL 6.1.1.a (Dose Assessment) can not be used to determine the classification, if any, due to the potential uncertainty of this "default" EAL. Two times the I OCFR20, Appendix B limits for noble gas and Iodine 131 are being used for this EAL, due to concerns that the State of New Jersey have pertaining to this EAL and based on the above mentioned uncertainties. REFERENCES NUMARC NESP-007, AUI.l, AUl.4 HC.OP-AB.ZZ-126(Q), Abnormal Releases of Gaseous Radioactivity HC.RP-AR.SP-OOOI(Q), Radiation Monitoring System Alarm Response Off-Site Dose Calculation Manual, Section 2.0 - Gaseous Effluents NUMARC Draft White Paper, 7-25-94, 9-10-94. Technical Specification 3.11.2.1 EAL - 6.1. l.d Rev.00 Page 4 of 4

HCGS EALIRALTechnical Basis 6.0 Radiological Releases/Occurrences 6.1 Gaseous Effluent Release ALERT- 6.1.2.a IC Any Unplanned R~!ease of Gaseous Radioactivity to the Environment that exceeds 200 Times Radiological Technical Specifications for 15 minutes or longer EAL Dose Assessment indicates EITHER of the following at the :MEA or beyond as calculated on the SSCL:

TEDE 4-Day Dose of?: 2.0E+Ol mRem
Thyroid-CDE Dose of?! 6.8E+Ol mRem based on Plant Vent effluent sample analysis and NOT on a default Noble Gas to Iodine Ratio Release is ongoing for ?! 15 minutes OPERA TI ON AL CONDITION - All BASIS Dose Assessment at or beyond the MEA exceeding the EAL threshold , can result from a Gaseous Radiological Release in excess of 200 times Technical Specifications. This condition results from an uncontrolled release of radioactivity to the environment, resulting in significantly elevated offsite dose rates. The threshold for this EAL is NOT based on a specific offsite dose rate, but rather on the loss of plant control implied by a radiological release of this magnitude that was not isolated within 15 minutes. Classification is based on an ongoing release that does not comply with a license condition. Unplanned is defined as any release for which a radioactive discharge permit was not prepared, or a release that exceeds the conditions on the applicable permit.

Dose Assessment using actual meteorological data provides an accurate indication of release magnitude. The use of dose assessment based EALs is therefore preferred over the use of Release Rate based EALs which utilize calculations which have built-in inaccuracies because ODCM default meteorological data is used. As long as dose assessment is available, this EAL should be used in place of EAL 6.1.2.d. EAL - 6.1.2.a Rev.00 Page 1 of J

HCGS EAL/RALTechnical Basis It is not intended that the release be averaged over 15 minutes, but exceed 200 times the Technical Specification limit for 15 minutes or longer. In addition, it is intended that the event be declared as soon as it is determined that the release will exceed 200 times the limit for 15 minutes or longer. Barrier Analysis NIA ESCALATION CRITERIA Emergency Classification will escalate to a Site Area Emergency when the effluent release concentration increases to a level that would cause a I 00 mRem dose at the Protected Area Boundary. DISCUSSION Prorating the 500 mRem/yr criterion for the TEDE 4-day dose: time (8766 hr/yr); the 200 x Tech. Spec. multiplier; and, Artificial Island's Allocation Factor of 0.5 (50% per site), the associated site boundary dose rate would be 5.7 mRem/hr. 500mReml yr TEDE 4-Day MEA Dose Rate= ( )(200)(05) = 5.7 mRem/hr 8766hr I yr This is rounded to 5.0 mRem/hr. The TEDE 4-day Dose is based on a default (assumed) 4 hour release duration. Therefore 5.0 mRem/hr

4 hours = 20 mRem.

Prorating the 1500 mRem/yr criterion for the Thyroid-COE Dose: time (8766 hr/yr); the 200 x Tech. Spec. multiplier; and, Artificial Island's Allocation Factor of 0.5 (50% per site), the associated site boundary dose rate would be 17 mRem/hr. 1500mRem/ yr Thyroid-CD£ MEA Dose Rate= ( )(200)(.5) = 0.17 mRem/hr 8766hr I yr The Thyroid-COE Dose is based on a 4 hour release duration. Therefore 17 mRem/hr

4 hours

= 68 mRem.

DEVIATION None EAL - 6.1.2.a Rev.00 Page 2 of 3

HCGS EAL/RALTechnical Basis REFERENCES NUMARC NESP-007, AAl .4 Off-Site Dose Calculation Manual, Section 2.0 - Gaseous Effluents NUMARC Draft White Paper, 7-25-94, 9-10-94. Technical Specification 3.11.2.1 EAL - 6. 1.2.a Rev. 00 Page 3 of 3

HCGS EAL/RALTechnical Basis 6.0 Radiological Releases/Occurrences 6.1 Gaseous Effluent Release ALERT- 6.1.2.b IC Any Unplanned Release of Gaseous Radioactivity to the Environment that exceeds 200 Times Radiological Technical Specifications for 15 minutes or longer EAL. Dose Rate measured at the Protected Area Boundary or beyond EXCEEDS 5 mRem/hr Release is ongoing for ~ 15 minutes OPERATIONAL CONDITION - All BASIS Measured Dose Rates at or beyond the MEA exceeding the EAL threshold , can result from a Gaseous Radiological Release in excess of 200 times Technical Specifications. This condition results from an uncontrolled release of radioactivity to the environment, resulting in significantly elevated offsite dose rates. The threshold for this EAL is NOT based on a specific offsite dose rate, but rather on the loss of plant control implied by a radiological release ofthis magnitude that was not isolated within 15 minutes. Classification is based on an ongoing release that does not comply with a license condition. Unplanned is defined as any release for which a radioactive discharge permit was not prepared, or a release that exceeds the conditions on the applicable permit. It is not intended that the release be averaged over 15 minutes, but exceed 200 times the Technical Specification limit for 15 minutes or longer. In addition, it is intended that the event be declared as soon as it is determined that the release will exceed 200 times the limit for 15 minutes or longer. Barrier Analysis NIA EAL - 6.1.2.b Rev. 00 Page 1 of 2

HCGS EAL/RALTechnical Basis ESCALATION CRITERIA Emergency Classification will escalate to a Site Area Emergency when effluent release concentration increases to a level that would cause a 100 mRem dose at the Protected Area Boundary. DISCUSSION Prorating the 500 mRem/yr criterion for: time (8766 hr/yr); the 200 x Tech. Spec. multiplier; and, Artificial Island's Allocation Factor of 0.5 (50% per site), the associated site boundary dose rate would be 5. 7 mRem/hr. 500mRem/yr Protected Area Boundary Dose Rate= ( )(200)(.5) = 5.7 mRem/hr 8766hr I yr This is rounded to 5.0 mRem/hr DEVIATION None REFERENCES NUMARC NESP-007, AAI.3 Off-Site Dose Calculation Manual, Section 2.0 - Gaseous Effluents NUMARC Draft White Paper, 7-25-94, 9-10-94. Technical Specification 3. 11.2.1 EAL - 6.1.2.b Rev.00 Page 2 of 2

HCGS EAL/RALTechnical Basis 6.0 Radiological Releases/Occurrences 6.1 Gaseous Effluent Release ALERT- 6.1.2.c IC Any Unplanned Release of Gaseous Radioactivity to the Environment that exceeds 200 Times the I OCFR20, Appendix B limits for 30 minutes or longer EAL

   . Gaseous effluent release sample analysis for ANY one of the following indicates a concentration of:

FRVS:

          ~ l.13E-Ol µCi/cc Total Noble Gas
           ~ 2.71E-05 µCi/cc 1-131

NPV:

          ~ 2.43E-02 µCi/cc Total Noble Gas
           ~ 5.SlE-06 µCi/cc 1-131

SPV:

          ~ 2.27E-03 µCi/cc Total Noble Gas
           ~ 5.44E-07 µCi/cc 1-131 Dose Assessment results NOT available Release is ongoing for~  30 minutes OPERATIONAL CONDITION - All BASIS Total gaseous effluent release sample analysis exceeding the EAL threshold for any of the plant vents listed (FRVS, NPV, SPY), can result from a Gaseous Radiological Release in excess of 200 times 10CFR20, Appendix B limits. This condition results from an uncontrolled release of radioactivity to the environment, resulting in elevated offsite dose rates.

EAL - 6. 1.2.c Rev. 00 Page 1 of 3

HCGS EAL/RALTechnical Basis The threshold for this EAL is NOT based on a specific offsite dose rate, but rather on the loss of plant control implied by a radiological release of this magnitude that was not isolated within 30 minutes. The final integrated dose is very low and is not the primary concern. Classification is based on an ongoing release that does not comply with a license condition. The HTV is not included under this EAL since there are no provisions for collecting a HTV grab sample. Unplanned is defined as any release for which a radioactive discharge permit was not prepared, or a release that exceeds the conditions on the applicable permit. It is not intended that the release be averaged over 30 minutes, but exceed 200 times the 10CFR20, Appendix B limit for 30 minutes or longer. In addition, it is intended that the event be declared as soon as it is determined that the release will exceed 200 times the limit for 30 minutes or longer. Barrier Analysis NIA ESCALATION CRITERIA Emergency Classification will escalate to a Site Area Emergency when effluent release concentration increases to a level that would cause a 100 mRem TEDE dose or Thyroid-CDE of 500 mRem for I-131 at the Protected Area Boundary. DISCUSSION Refer to Basis Section for EAL 6.1.2.d for the 10CFR20, Appendix B Noble Gas and Thyroid Committed Dose release Rate Calculations. Calculation of the threshold sample concentrations are as follows: z:-Rr,'S JvO

r. ,.,

          ,.., ble Gas .Jamp

(' . = 4.80£ + 05µCi I sec le C011centrat1011 .

                                                                        = l.13E-Ol µCi/cc 472x9000cfm l.15£ + 02µCi I sec FRVS 1-131             Sample Concentration      =                      = 2. 71 E-05 µCi/cc 472x9000cfm
                                            .      4.80£ + OSµCi I sec NPV Noble Gas Sample Concentrat1on =                                     = 2.43E-02 µCi/cc 472x4.19£ + 4cfm l.15£ + 02µCi I sec NPV 1-131             Sample Concentration =                            = 5. 81 E-06 µCi/cc 472x4.19£ + 04cfm EAL - 6.1.2.c Rev. 00 Page 2 of 3

HCGS EAL/RALTechnical Basis 4.80£ + 05µCi I sec . SPV Noble Gas Sample Concentration = = 2.27E-03 µCi/cc 472x4.48£ + 5cfm 1.15£ + 02µCi I sec . SPV /-131 Sample Concentration= = 5.44E-07 µCi/cc 4 72x4.48£ + 05cfm Where: 472 =conversion factor (28,317 cc/ft3 x 1 min./60 sec.) 9000 cfm = FRVS Vent Flow (maximum) 4.19E+04 cfm = NPV Vent Flow (maximum) 4.48E+05 cfm = SPV Vent Flow (maximum) The noble gas release rate of 4.80E+05 µCi/sec is obtained by multiplying the 10CFR20, Appendix B limit release rate of 2AOE+03 µCi/sec times 200. The iodine release rate of l. l 5E+02 µCi/sec is obtained by multiplying the IOCFR20, Appendix B limit release rate of 5.75E-Ol µCi/sec times 200. DEVIATION The value for EAL 6.1.2.c is based on one meteorological case and one isotopic mixture found in the ODCM. A radiological release based on this specific release rate could produce a TEDE Dose which would require a General Emergency classification or not meet the Alert classification, depending on the meteorological conditions and isotopic mixture. EAL 6.1.2.c would not be used unless EAL 6. 1.2.a (Dose Assessment) can not be used to determine the classification, if any, due to the potential uncertainty of this "default" EAL. Two hundred times the 10CFR20, Appendix B limit noble gas and Iodine 131 are being used for this EAL, due to concerns that the State of New Jersey had pertaining to this EAL and based on the above mentioned uncertainties. The time limit has been increased from 15 minutes to 30 minutes, to allow additional time to perform dose assessment, since the threshold fore this EAL is only 20% of the value allowed per NESP-007 and we do not wish to use this "default" EAL, unless absolutely necessary. REFERENCES NUMARC NESP-007, AAl.2, AAl. I, AAl.4 Off-Site Dose Calculation Manual, Section 2.0 NUMARC Draft White Paper, 7-25-94, 9-10-94. Technical Specification 3.11.2.1 EAL - 6.1.2.c Rev. 00 Page 3 of 3

HCGS EAL/RALTechnical Basis 6.0 Radiological Releases/Occurrences 6.1 Gaseous Effluent Release ALERT- 6.1.2.d IC Any Unplanned Rel~ase of Gaseous Radioactivity to the Environment that exceeds 200 Times 10 CFR20, Appendix B Limits for 30 minutes or longer EAL Valid High Alann received from ANY one of the following Plant Effluent RMS Channels:

FRVS Noble Gas (Grid 1/3; 9RX680)
NPV Noble Gas (Grid 1/3; 9RX590)
NPV Iodine (Grid 3; 9RX601)
SPV Noble Gas (Grid 1/3; 9RX580)
SPV Iodine (Grid 3; 9RX605)
HTV Noble Gas (Grid 3; 9RX516)

Total Plant Vent release rate EXCEEDS EITHER one of the following limits:

4.80E+05 µ.Ci/sec Total Noble Gas
1.15E+02 µ.Ci/sec 1-131 (NPV & SPY ONLY)

Dose Assessment results NOT available Release is ongoing for ~ 30 minutes OPERATIONAL CONDITION - All BASIS Valid High alarm and effluent release rate values exceeding the EAL threshold, can result from a Gaseous Radiological Release in excess of 200 times I OCFR20, Appendix B limits . This condition results from an uncontrolled release of radioactivity to the environment, resulting in elevated offsite dose rates. The threshold for this EAL is NOT based on a specific offsite dose EAL - 6.1.2.d Rev. 00

                                          .Page 1 of 4

HCGS EAL/RALTechnical Basis rate, but rather on the loss of plant control implied by a radiological release ofthis magnitude that was not isolated within 30 minutes. The final integrated dose is very low and is not the primary concern. Valid is defined as the High alarm actuating specifically due to a Gaseous Release exceeding Technical Specification limits, thus precluding unwarranted event declaration as the result of spurious actuation. Classification is based on an ongoing release that does not comply with a license condition. Unplanned is defined as any release for which a radioactive discharge permit was not prepared, or a release that exceeds the conditions on the applicable permit. The EAL value for Total Plant Vent release rate was determined using default X/Q values from the ODCM which provides a less accurate method of evaluation release magnitude then using dose assessment with real time meteorological data. For that reason, this EAL should not be utilized if Dose Assessment is available. Dose Assessment will take in account actual meteorological conditions, plant vent flows and plant vent effluent concentrations to provide a more accurate assessment of a radiological release. If Dose Assessment is available than refer to EAL 6.1.2.a for classification. The Total Plant Vent release rate can be obtained from SPDS or by adding up NPV, SPY, FRVS and HTV noble gas readings. It is not intended that the release be averaged over 30 minutes, but exceed 200 times IOCFR20, Appendix B limits for 30 minutes or longer. In addition, it is intended that the event be declared as soon as it is determined that the release will exceed 200 times the limit for 30 minutes or longer. Barrier Analysis NIA ESCALATION CRITERIA Emergency Classification will escalate to a Site Area Emergency when effluent release concentration increases to a level that would cause a I 00 mRem dose at the Protected Area Boundary DISCUSSION The release rate thresholds for this EAL are obtained by multiplying the I OCFR20, Appendix B Limit release rates of2.4E+03 µCi/sec and 5.78E-Ol µCi/sec for Noble Gases and 1-131 respectively, times 200. Total Noble Gas release rate is the summation of all plant vent release rates. This EAL includes an Iodine Release rate for NPV & SPY since these vents have Iodine monitors. Determination of the Iodine Release Rate from the Iodine monitor is accomplished by multiplying the Iodine reading (in µCi/cc) by the applicable vent flow rate and 472 (conversion factor). Iodine Release Rates for FRVS and HTV are excluded since these vents do not include EAL - 6.1.2.d Rev. 00 Page 2 of 4

HCGS EAL/RALTechnical Basis an Iodine detector. The SPDS Total Iodine offsite Release Rate does not provide useful information, since this based on a default valve of 1000 times less than the Total Noble Gas offsite Release Rate , which could be grossly inaccurate. A gaseous effluent sample is needed to accurately quantify the Iodine Release Rate. 10CFR20, Appendix B Limit Calculation for Noble Gas uCi/Second = 100 mRem/year * (Allocation Factor) (ODCM X/Q) * (ODCM DRCF) WHERE: uCi/Second = Total Noble Gas Release Rate from Salem (Unit 1 & Unit 2) or Hope Creek (all Vents; NPV, SPV, FRVS, and HTV) which would result in a TEDE Dose Rate of 50 mRem/year. ODCM X/Q = Site Specific (Salem or Hope Creek) dispersion factor at the Site Boundary in sec/m3. ODCM DRCF = Site Specific (Salem or Hope Creek) dose rate conversion factor in mRem/year/uCifm3. ODCM XJQ = 2.67£-06 sec/m3 ODCM DRCF = 7.80E+03 mRem/yr/uCi/m3 Allocation Factor= 5.00E-01 2.40E+03 uCi/Second = (100 mRem/year~ * (5.00E-01) . (2.67E-06 sec/m ) * (7.80E+03 mRem/yr/uCi/m3) 2.40E+03 µCi/sec

200 = EAL value.

4.80E+05 µCi/sec = EAL value 10CFR20, Appendix B Limit Calculation for Thyroid Committed Dose

µCi/Second = 50 rnRem/year     *  (Allocation Factor)

(ODCM X/Q) * (ODCM THY DRCF) WHERE: µCi/Second= Total Iodine 131 release rate from Salem (Unit 1 or 2) or Hope Creek (all Vents; NPV, SPY, FRVS, and HTV). ODCM X/Q = Site Specific (Salem or Hope Creek) dispersion factor at the Site Boundary in sec/m~. ODCM THY DRCF =is the most limiting potential pathway (inhalation, child thyroid 1-13 1) dose rate conversion factor in mRem/year/µCi/m 3 . EAL - 6.1.2.d Rev. 00 Page 3 of 4

HCGS EALIRALTechnical Basis 3 ODCM X/Q = 2.67E-06 sec/m 3 ODCM THY DRCF = I.62E+07 mRem/yr/µCi/m Allocation Factor= 5.00E-01 5.78E-01 µCi/Second= (50 mRem/year) * (5.00E-01) 3 (2.67E-06 sec/m3) * (1.62E+07 mRem/yr/µCi/m ) 5.78E-01 µCi/sec* 200 =EAL value. l.15E+02 µCi/sec = EAL value. DEVIATION The value for EAL 6.1.2.d is based on one meteorological case and one isotopic mixture found in the ODCM. A radiological release based on this specific release rate could produce a TEDE Dose which would require a General Emergency classification or not meet the Alert classification, depending on the meteorological conditions and the isotopic mixture. EAL 6.1.2.d would not be used unless EAL 6.1.2.a (Dose Assessment) can not be used to determine the classification, if any, due to the potential uncertainty of this "default" EAL. Two hundred times the IOCFR20, Appendix B limits of noble gas and Iodine 131 are being used for this EAL, due to concerns that the State of New Jersey had pertaining to this EAL and based on the above mentioned uncertainties. The time limit has been increased from 15 minutes to 30 minutes, to allow additional time to perform dose assessment, since the threshold for this EAL is only 20% of the value allowed per NESP-007 and we do not wish to use this "default" EAL, unless absolutely necessary. REFERENCES NUMARC NESP-007, AAl.1, AAl.4 OP-AB.ZZ-126(Q), Abnormal Releases of Gaseous Radioactivity Off-Site Dose Calculation Manual, Section 2.0 - Gaseous Effluents NUMARC Draft White Paper, 7-25-94, 9-10-94. Technical Specification 3.11.2.1 EAL - 6.1.2.d Rev.00 Page 4 of 4

HCGS EALIRALTechnical Basis 6.0 Radiological Releases/Occurrences 6.1 Gaseous Effluent Release SITE AREA EMERGENCY - 6.1.3.a IC Boundary Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity Exceeds 100 mRem Total Effective Dose Equivalent {TEDE) or 500 mRem Thyroid CDE Dose for the actual or projected duration of the release EAL Dose Assessment indicates EITHER one of the following at the MEA or beyond as calculated on the SSCL:

TEDE 4-Day Dose of~ 1.0E+02 mRem
Thyroid-COE Dose of ~ 5.0E+02 mRem based on Plant Vent effluent sample analysis and NOT on a default Noble Gas to Iodine Ratio OPERATIONAL CONDITION - All BASIS The TEDE 4-Day Dose of~ l.OE+02 mRem corresponds directly to the NUMARC dose of 100 mRem.
The Thyroid-COE Dose of~ 5.0E+02 mRem corresponds directly to the NUMARC dose of 500 mRem.

Dose Assessment using actual meteorological data provides an accurate indication of release magnitude. The use of dose assessment based EALs is therefore preferred over the use of Release Rate based EALs which utilize calculations which have built-in inaccuracies because ODCM. default Meteorological data is used. Imminent is defined as expected to occur within 2 hours. Barrier Analysis NIA EAL - 6.1.3.a Rev.00 Page 1 of 2

HCGS EALIRALTechnical Basis ESCALATION CRITERIA Emergency Classification escalates to a General Emergency when actual or projected doses exceed EPA Protective Action Guidelines. DISCUSSION This value provides a desirable gradient (one order of magnitude) between the Site Area Emergency and General Emergency classifications. No site allocation factor (.5) is used in this calculation due to the assumption that releases of this magnitude will be from one site. The dose projection code assumes a 4 hour release utilizing current 15 minute average release rate data. For the TEDE 4-Day Dose, 100 mRem/hr

4 hr= 400 mRem. For the Thyroid-COE Dose, 500 mRem/hr
4 hr = 2000 mRem.

DEVIATION None REFERENCES NUMARC NESP-007, ASI.3 EPA 400-R-92-001, Manual of Protective Action Guides and Protective Actions for Nuclear Incidents NUMARC Draft White Paper, 7-25-94, 9-10-94 EAL - 6.1.3.a Rev. 00 Page 2 of 2

HCGS EALIRALTechnical Basis 6.0 Radiological Releases/Occurrences 6.1 Gaseous Effluent Release SITE AREA EMERGENCY-6.1.3.b IC Boundary Dose Resulting from an Actual or ImmL~ent Release of Gaseous Radioactivity Exceeds 100 mRem Total Effective Dose Equivalent (TEDE) or 500 mRem Thyroid CDE Dose for the actual or projected duration of the release EAL Dose Rate measured at the Protected Area Boundary or beyond EXCEEDS 100 mRem/hr Release is expected to continue for ~ 15 minutes OPERA TI ON AL CONDITION - All BASIS An actual dose rate of 100 mRem/hr which is expected to continue for 2: 15 minutes indicates a . substantial radiological release which could exceed the 10CFR20 annual average population exposure limit of I 00 mRem TEDE, using the assumption of a one hour release duration. Barrier Analysis NIA ESCALATION CRITERIA Emergency Classification will escalate to a General Emergency when actual or projected doses exceed EPA Protective Action Guidelines. DISCUSSION An actual dose of 100 mRem Total Effective Dose Equivalent (TEDE) is based on the IOCFR20 annual average population exposure limit. Unless otherwise indicated, the conversion from whole body dose to TEDE is I: I. Measured dose rates will be taken at the Protected Area Boundary, and a~ 15 minute release duration threshold will be applied to be conservative. EAL - 6.1.3.b Rev. 00 Page 1 of 2

HCGS EALIRALTechnical Basis DEVIATION None REFERENCES NUMARC NESP-007, ASI.4 EPA 400-R-92-001, Manual of Protective Action Guides and Protective Actions for Nuclear Incidents NUMARC Draft White Paper, 7-25-94, 9-10-94 EAL - 6.1.3.b Rev.00 Page 2 of 2

HCGS EAL/RALTechnical Basis 6.0 Radiological Releases/Occurrences 6.1 Gaseous Effluent Release SITE AREA EMERGENCY- 6.1.3.c IC Boundary Dose Resdting from an Actual or Imminent Release of Gaseous Radioactivity Exceeds 100 mRem Total Effective Dose Equivalent (TEDE) or 500 mRem Thyroid CDE Dose for the actual _or projected duration of the release EAL Analysis of field survey samples at the Protected Area Boundary indicates EITHER one of the following:

 *   ~ 4.36E+02 CCPM
 *   ~ 3.85E-07 µCi/cc 1-131 OPERATIONAL CONDITION - All BASIS The Corrected Counts per Minute (CCPM) value is based on reading(s) obtained using a radiation count rate meter such as a RM-14 or E-l 40N with an HP260 probe attached. The Iodine-131 field survey sample concentration threshold is based on 1-131 dose conversion factors from EPA-400. The thresholds are based on a Thyroid-COE dose rate of 500 mRem/hr for 1-131.

Barrier Analysis NIA ESCALATION CRITERIA Emergency Classification will escalate to a General Emergency when actual or projected doses exceed EPA Protective Action Guidelines. EAL - 6.1.3 .c Rev. 00

                                           .t>age 1 of 3

HCGS EAL/RALTechnical Basis DISCUSSION The release sample concentration calculations are as follows. The sample concentration is calculated using the 1-131 Dose Conversion Factor from EPA-400: Solving the following equation for µCi/cc: mRem/hr =(µCi/cc) (Dose Conversion Factor) Then;

                                    .                         SOOmRem I hr

1-131 Sample Co11centrat10n = ( ) = 3.85E-07 µCi/cc 1.30£ + 09mRem I µCi I cc I hr Where 1.30E+09 mRem/µCi/cc/hr is the Dose Conversion Factor from EPA-400, Table 5-4 and includes the EP A-400 breathing rate .

The Corrected Counts per Minute reading is calculated using the 1-131 Sample concentration, and factors for using an RM-14 or E- I 40N with an HP260 probe. Solving the following equation for CCPM: µCi/cc= CCPl\f . (Detector Eflicicncy)(Collcclion Efficiency)( Conversion Factor* DPll.f to µCi)(Volumc - ft3 )(Conversion Factor* cc to ft3 ) Then; CCPM = (3.85E-07 µCi/cc) (2.00E-03 CCPM/DPM) (0.9) (2.22E+06 DPM/µCi)

3 (IO ft ) (2.832E+04 cc/ft 3 ) = 4.36E+02 CCPM Where:

2.00E-03 = Detector Efficiency- CCPMDPM 0.9 (or 90%) = Collection Efficiency 2.22£+06 = Conversion factor - DPM*'µCi 10ft3 = Volume 2.832£+0./ = (0111*ersio11 factor - cc to ft3 CCPM= Corrected Counts per Minute using an RM-14 or E-140N with an HP260 probe. EAL - 6.1.3.c Rev. 00 Page 2 of 3

HCGS EALIRALTechnical Basis DEVIATION None REFERENCES NUMARC NESP-007, AS 1.4 EPA 400-R-92-001, Manual of Protective Action Guides and Protective Actions for Nuclear Incidents FEMA REP-2, Rev. 1, 7/87, Guidance on Offsite Emergency Radiation Measurement Systems, Phase-1 Airborne Release SORC Summary 07/10/89 RPCS Thyroid Dose Commitment Factor Paper (NRP-94-0557), 11122/94 EAL - 6.1.3.c Rev. 00 Page 3 of 3

HCGS EAL/RALTechnical Basis 6.0 Radiological Releases/Occurrences 6.1 Gaseous Effluent Release SITE AREA EMERGENCY-6.1.3.d IC Boundary Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity Exceeds 100 mRem Total Effective Dose Equivalent (TEDE) or 500 mRem Thyroid COE Dose for the actual or projected duration of the release EAL Valid High Alarm received from ANY of the following Effluent RMS Channels:

FRVS Noble Gas (Grid 1/3; 9RX680)
NPV Noble Gas (Grid 1/3; 9R..'C590)
SPV Noble Gas (Grid 1/3; 9RX580)
HTV Noble Gas (Grid 3; 9RXS 16)

Total Plant Vent release rate EXCEEDS 7.6E+07 µ.Ci/sec Total Noble Gas Dose Assessment results NOT available Release is ongoing for ~ 15 minutes OPERATIONAL CONDITION - All BASIS Valid High alarm and effluent release rate values exceecing the EAL threshold, indicates a substantial Gaseous Radiological Release which could exceed the I OCFR20 average annual population exposure limit of 100 mRem TEDE, using the assumption of a one hour release duration. The EAL value for Total Plant Vent release rate wa: determined using default X/Q values from the ODCM which provides a less accurate method of evaluation release magnitude then using EAL - 6.1.3.d Rev.00 Page 1 of 3

HCGS EAL/RALTechnical Basis dose assessment with real time meteorological data. For that reason, this EAL should not be utilized if Dose Assessment is available. Dose Assessment will take in account actual meteorological conditions, plant vent flows and plant vent effluent concentrations to provide a more accurate assessment of a radiological release. If Dose Assessment is available then refer to EAL 6.1.3 .a for classification. The Total Plant Vent release rate can be obtained from SPDS or by adding up NPV, SPV, FRVS and HTV noble gas readings. It is not intended that the release be averaged over 15 minutes, but that the Release Rate exceed the EAL value for > 15 minutes. Barrier Analysis NIA ESCALATION CRITERIA Emergency Classification will escalate to a General Emergency when effluent release concentration increases to a level that would cause a I 000 mRem dose at the Protected Area Boundary. DISCUSSION This EAL does not utilize an Iodine Release rate because the corresponding Site Area threshold for Iodine is above the upper range of the NPV and SPY Iodine monitoring channels. Iodine Release rates for FRVS and the HIV are excluded since these do not include an Iodine detector. To obtain a site specific value to trigger the performance of dose assessment is not necessary, since this will be done when the UE value is reached. This value will supply a set point to classify a Site Area Emergency (SAE), if dose assessment has not been performed within 15 minutes. A release rate of 7.6E+07 µCi/sec was backcalculated from a TEDE Dose of 100 mRem/hour at the Site MEA. The assumptions that went into this calculation were as follows: Release Point: FRVS Release Rate: 9000cfm ODCM XJQ = 2.67£-06 sec/m 3 Isotopic mixture: FSAR isotopic mixture for a design basis LOCA EAL - 6.1.3.d Rev.00 Page 2 of 3

HCGS EALIRALTechnical Basis Dose Rate Conversion Factors: EPA 400-R-92-00 I (Manual of Protective Actions for Nuclear Incidents) Dose Rate Conversion Factors. DEVIATION The NUMARC basis states that the FSAR source term assumptions should be used in determining the indications for monitors. The NUMARC Draft White Paper states the FSAR source term should not be used unmodified. This NUMARC EAL is calculated using the FSAR Isotopic Mixture for a Design Basis LOCA and the Dose Rate Conversion Factors found in EPA 400-R-OO 1. The combination of using the FSAR Isotopic mixture and the EPA 400 dose Rate Conversion Factors calculate an accurate accident source term. REFERENCES NUMARC NESP-007, ASl.l, ASI.4 OP-AB.ZZ-126(Q), Abnormal Releases of Gaseous Radioactivity Off-Site Dose Calculation Manual, Section 2.0 - Gaseous Effluents NUMARC Draft White Paper, 7-25-94, 9-10-94. Technical Specification 3.11.2.1 FSAR Section 15 EPA 400-R-OO 1, Manual of Protective Action Guides and Protective Actions for Nuclear Incidents EAL - 6.1.3.d Rev. 00 Page 3 of 3

HCGS EAL/RALTechnical Basis 6.0 Radiological Releases/Occurrences 6.1 'Gaseous Effluent Release GENERAL EMERGENCY - 6.1.4.a IC Boundary Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity Exceeds 1000 mRem Total Effective Dose Equivalent (TEDE) or 5000 mRem Thyroid CDE Dose for the actual or projected duration of the release EAL Dose Assessment indicates EITHER one of the following at the MEA or beyond as calculated on the SSCL:

TEDE 4-Day Dose of ~ 1.0E+03 mRem
Thyroid-COE Dose of ~ 5.0E+03 mRem based on Plant Vent effluent sample analysis and NOT on a default Noble Gas tG Iodine Ratio OPERATIONAL CONDITION - All BASIS The TEDE 4-Day Dose of~ 1.0E+03 mRem corresponds directly to the NUMARC dose of 1000 mRem which exceeds EPA Protective Action Guideline criteria for a General Emergency.

The Thyroid-CDE Dose of~ 5.0E+03 mRem corresponds directly to the NUMARC dose of 5000 mRem which exceeds EPA Protective Action Guideline criteria for a General Emergency. Imminent is defined as expected to occur within 2 hours. Barrier Analysis NIA ESCALATION CRITERIA NIA EAL - 6.1.4.a Rev. 00 Page 1 of 2

HCGS EALIRALTechnical Basis DISCUSSION No site allocation factor (.5) is used in this calculation due to the assumption that releases of this magnitude will be from one site. DEVIATION None REFERENCES NUMARC NESP-007, AG 1.3 EPA 400-R-92-001, Manual of Protective Action Guides and Protective Actions for Nuclear Incidents NUMARC Draft White Paper 7-25-94, 9-10-94 EAL - 6.1.4.a Rev. 00 Page 2 of 2

HCGS EALIRALTechnical Basis 6.0 Radiological Releases/Occurrences 6.1 Gaseous Effluent Release GENERAL EMERGENCY - 6.1.4.b IC Boundary Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity Exceeds 1000 mRem Total Effective Dose Equivalent (TEDE) or 5000 mRem Thyroid CDE Dose for the actual or projected duration of the release EAL Dose Rate measured at the Protected Area Boundary or beyond EXCEEDS 1000 mRem/hr Release is expected to continue for ~ 15 minutes OPERATIONAL CONDITION - All BASIS An actual dose rate of I 000 mRem/hr indicates the EPA Protective Action Guide may be exceeded for the general public. Barrier Analysis NIA ESCALATION CRITERIA NIA DISCUSSION An actual projected dose of IOOO mRem Total Effective Dose Equivalent (TEDE) is based on the EPA protective action guidance which indicates that public protective actions are indicated if the dose exceeds I Rem whole body. This is consistent with the emergency class description for a General Emergency. A release rate equivalent to I 000 mRem/hr boundary dose rate may also be used if TEDE projections are not available. Unless otherwise indicated, the conversion from whole body dose to TEDE is I: I. EAL- 6.1.4.b Rev.00 P:ige 1 of 2

HCGS EALIRALTechnical Basis DEVIATION None REFERENCES NUMARC NESP-007, AG 1.4 EPA 400-R-92-00 I, Manual of Protective Action Guides and Protective Actions for Nuclear Incidents EAL - 6.1.4.b Rev. 00 Page 2 of 2

HCGS EALIRALTechnical Basis 6.0 Radiological Releases/Occurrences 6.1 Gaseous Effluent Release GENERAL EMERGENCY - 6.1.4.c IC Boundary Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity Exceeds 1000 rnRem Total Effective Dose Equivalent (TEDE) or 5000 rnRem Thyroid CDE Dose for the actual or projected duration of the release EAL Analysis of field survey samples at the Protected Area Boundary indicates EITHER one of the following:

         =: 4.36E+03 CCPM
         =: 3.85E-06 µCi/cc 1-131 OPERATIONAL CONDITION - All BASIS The Corrected Counts per Minute (CCPM) value is based on reading(s) obtained using a radiation count rate meter such as a RM-14 or E- l 40N with an HP260 probe attached. The Iodine-131 field survey sample concentration threshold is based on I-131 dose factors from EPA-400. The thresholds are based on a dose rate of 5000 mRem/hr Thyroid-CDE for I-131.

Barrier Analysis NIA ESCALATION CRITERIA NIA DISCUSSION The release sample concentration calculatio;is are as follows. The sample concentration is calculated using the I-131 Dose Conversion Factor from EPA-400: EAL - 6.1.4.c Rev. 00 Page 1 of 3

HCGS EALIRALTechnical Basis Solving the following equation for µCi/cc: mRem/hr =(µCi/cc) (Dose Conversion Factor) Then; 5000m Rem I hr . 1-131 Sample Concentration = ( ) = 3.SSE-06 µCi/cc 1.30£ + 09m Rem I µCi I cc I hr Where 1.30E+09 mRern/µCi/cc/hr is the Dose Conversion Factor from EPA-400, Table 5-4 and includes the EPA-400 breathing factor. The Corrected Counts per Minute reading is calculated using the I-131 Sample concentration, and factors for using an RM-14 or E-140N with an HP260 probe. Solving the following equation for CCPM: µCi/cc= CCPI\! (Detector Efficiency) (Collection Efficiency) (Conversion Factor - DPM lo µCi) (Volume - ft3 ) (Conversion Factor - cc to ft3 ) ~ Then; CCPM = (3.85E-06 µCi/cc) (2.00E-03 CCPM/DPM) (0.9) (2.22E+06 DPM/µCi) (IO ft 3) * (2.832E+04 cc/ft 3 ) = 4.36E+03 CCPM Where: 2.00E-03 = Detector Efficiency - CCPMIDPM 0.9 (or 90%) = Collection Efficiency 2.22£+06 = Co11versio11factor -DPMlµCi 10ft3 = Volume 2.832E+D-I = Conversion factor - cc to jt3 CCPM= Corrected Counts per Minute using an RM-14 or E-140N with an HP260 probe. DEVIATION None EAL - 6.1.4.c Rev. 00 Page 2 of 3

HCGS EALIRALTechnical Basis REFERENCES NUMARC NESP-007, AG 1.4 EPA 400-R-92-001, Manual of Protective Action Guides and Protective Actions for Nuclear Incidents FEMA REP-2, Rev. I/July 1987, Guidance on Offsite Emergency Radiation Measurement Systems, Phase- I Airborne Release SORC Summary 07/10/89 RPCS Thyroid Dose Commitment Factor paper NRP-94-0557, 11-22-94 EAL - 6.1.4.c Rev. 00 Page 3 of 3

HCGS EALIRALTechnical Basis 6.0 Radiological Releases/Occurrences 6.1 Gaseous Effluent Release GENERAL EMERGENCY - 6.1.4.d IC Boundary Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity Exceeds 1000 rnRem Total Effective Dose Equivalent (TEDE) or 5000 rnRem Thyroid CDE Dose for the actual or projected duration of the release EAL Valid High Alarm received from ANY one of the following Plant Effluent RMS Channels:

FRVS Noble Gas (Grid 1/3; 9RX680)
NPV Noble Gas (Grid 1/3; 9RX590)
SPV Noble Gas (Grid 1/3; 9RX580)
HTV NobleGas (Grid3; 9RX516)

Total Plant Vent release rate EXCEEDS 7.6E+08 µCi/sec Total Noble Gas Dose Assessment results NOT available Release is ongoing for .:::, 15 minutes OPERATIONAL CONDITION - All BASIS Valid High alarm and effluent release rate values exceeding the EAL threshold, indicates a substantial Gaseous Radiological Release which could exceed the EPA Protective Action Guide exposure of I 000 mRem TEDE, using the assumption of a one hour release duration. The EAL value for Total Plant Vent release rate was determined using default X/Q values from the ODCM which provides a less accurate method of evaluation release magnitude then using dose assessment with real time meteorological data. For that reason, this EAL should not be EAL - 6.1.4.d Rev. 00 Page 1 of 3

HCGS EAL/RALTechnical Basis utilized if Dose Assessment is available. Dose Assessment will take into account actual meteorological conditions, plant vent flows and plant vent effluent concentrations to provide a more accurate assessment of a radiological release. If Dose Assessment is available then refer to EAL 6.1.4.a for classification. The Total Plant Vent release rate can be obtained from SPDS or by adding up NPV, SPY, FRVS and HTV noble gas readings. It is not intended that the release be averaged over 15 minutes, but that the Release Rate exceed the EAL value for > 15 minutes. Barrier Analysis NIA ESCALATION CRITERIA None DISCUSSION This EAL does not utilize an Iodine Release rate because the corresponding General threshold for Iodine is above the upper range of the NPV and SPV Iodine monitoring channels. Iodine Release rates for FRVS and the HTV are excluded since these do not include an Iodine detector. To obtain a site specific value to trigger the performance of dose assessment is not necessary, since this will be done when the UE value is reached. This value will supply a set point to classify a General Emergency (GE), if dose assessment has not been performed within 15 minutes. A release rate of7.6E+08 µCi/sec was backcalculated from a TEDE Dose of IOOOmRem/hour at the Site MEA. The assumptions that went into this calculation were as follows: Release Point: FR VS Release Rate: 9000cfm ODCM XJQ = 2.67E-06 seclm' Isotopic mixture: FSAR isotopic mixture for a design basis LOCA Dose Rate Conversion Factors: EPA 400-R-92-001 (Manual of Protective Actions for Nuclear Incidents) Dose Rate Conversion Factors. DEVIATION EAL - 6. I .4.d Rev.00 Page 2 of 3

HCGS EALIRALTechnical Basis The NUMARC basis states that the FSAR source term assumptions should be used in determining the indications for monitors. The NUMARC Draft White Paper states the FSAR source term should not be used unmodified. This NUMARC EAL is calculated using the FSAR Isotopic Mixture for a Design Basis LOCA and the Dose Rate Conversion Factors found in EPA 400-R-001. The combination of using the FSAR Isotopic mixture and the EPA 400 dose Rate Conversion Factors calculate an accurate accident source term. REFERENCES NUMARC NESP-007, AG I. I, AG I .4 OP-AB.ZZ-I26(Q), Abnormal Releases of Gaseous Radioactivity Off-Site Dose Calculation Manual, Section 2.0 - Gaseous Effiu~nts NUMARC Draft White Paper, 7-25-94, 9-I0-94. Technical Specification 3.11.2. l FSAR Section I 5 EPA 400-R-OOI, Manual of Protective Action Guides and Protective Actions for Nuclear Incidents EAL - 6.1.4.d Rev. 00 Page 3 of 3

HCGS EAL/RALTechnical Basis 6.0 Radiological Releases/Occurrences 6.2 Liquid Effluent Release UNUSUAL EVENT- 6.2.1 IC Any Unplanned Release of Liquid Radioactivity to the Environment that Exceeds 2 Times the Radiological Technical Specifications for 60 minutes or longer EAL Valid Cooling Tower Blowdown Effluent Radiation Monitor High Alann Condition Sample analysis of liquid effluent indicates concentration in excess of 2 times Technical Specification limits Release continues for .::_60 minutes after the alarm occurs OPERATIONAL CONDITION - All BASIS A Valid Cooling Tower Blowdown Effluent Radiation Monitor High alann condition corresponds to the Technical Specification Liquid Effluent Release Limit. Despite this limit being below the EAL threshold, exceeding this limit with a failure to terminate the discharge may be a precursor to an Unplanned Liquid Radiological Release in excess of 2 times Technical Specifications that continues for greater than 60 minutes. The threshold for this EAL is NOT based on a specific offsite dose rate, but rather on the loss of plant control implied by a radiological release of this magnitude, that is not isolated in 60 minutes. The final integrated dose is very low and is not the primary concern. Valid is defined as the Cooling Tower Blowdown Effluent Radiation Monitor High Alarm actuating specifically due to a Liquid Release exceeding the Technical Specification limit, thus precluding unwarranted event declaration as the result of spurious actuation. Unplanned is defined as any release for which a radioactive discharge pennit was not prepared, or a release that exceeds the conditions on the applicable permit. It is not intended that the release be averaged over 60 minutes, but exceed 2 times the Technical Specification limit for 60 minutes or longer. In addition, it is intended that the event be declared as soon as it is determined that the release will exceed 2 times the limit for 60 minutes or longer. EAL- 6.2.1 Rev. 00 P~gc 1 of 2

HCGS EALIRALTechnical Basis Barrier Analysis NIA ESCALATION CRITERIA Emergency Classification will escalate to an Alert when the Liquid Effluent Release exceeds 200 times Technical Specification limits. DISCUSSION The Cooling Tower Blowdown Effluent Radiation Monitor (9RX506) monitors radioactivity in the cooling tower blowdown before it is discharged into the Delaware River and warns personnel of an excessive amount of radioactivity (greater than Technical Specification limits) being released to the environment. DEVIATION None REFERENCES NUMARC NESP-007, AUi .2 Off-Site Dose Calculation Manual, Section 1.0 - Liquid Effluents Technical Specifications LCO 3.11.1.1 HC.RP-AR.SP-OOOI(Q), Radiation Monitoring System Alarm Response EAL - 6.2.1 Rev. 00 Page 2 of 2

HCGS EAL/RALTechnical Basis 6.0 Radiological Relea~es/Occurrences 6.2 Liquid Effluent Release ALERT - 6.2.2 IC Any Unplanned Release of Liquid Radioactivity to the Environment that Exceeds 200 Times the Radiological Technical Specifications for 15 minutes or longer EAL Valid Cooling Tower Blowdown Eftluent Radiation Monitor High Alarm Condition Sample analysis of liquid eftluent indicates concentration in excess of 200 times Technical Specification limits Release continues for :;:15 minutes after the alarm occurs OPERATIONAL CONDITION - All BASIS A Valid Cooling Tower Blowdown Eftluent Radiation Monitor High alarm condition corresponds to the Technical Specification Liquid Eftluent Release Limit. Despite this limit being well below the EAL threshold, exceeding this limit with a failure to terminate the discharge may be a precursor to an Unplanned Liquid Radiological Release in excess of200 times Technical Specifications that continues for greater than 15 minutes. The threshold for this EAL is NOT based on a specific offsite dose rate, but rather on the loss of plant control implied by a radiological release of this magnitude, that is not isolated in 15 minutes. The release duration was reduced from 60 minutes (UE) to 15 minutes in recognition of the increased severity ofa release of this magnitude. Valid is defined as the Cooling Tower Blowdown Eftluent Radiation Monitor High Alarm actuating specifically due to a Liquid Release exceeding the Technical Specification limit, thus precluding unwarranted event declaration as the result of spurious actuation. Unplanned is defined as any release for which a radioactive discharge permit was not prepared, or a release that exceeds the conditions on the applicable permit. It is not intended that the release be averaged over 15 minutes, but exceed 200 times the Technical Specification limit for 15 minutes or longer. In addition, it is intended that the event be declared as soon as it is determined that the release will exceed 200 times the limit for 15 minutes or longer. EAL- 6.2.2 Rev.00 Page 1 of 2

HCGS EAL!RALTechnical Basis Barrier Analysis NIA ESCALATION CRITERIA NIA DISCUSSION The Cooling Tower Blowdown Effluent Radiation Monitor (9RX506) monitors radioactivity in the cooling tower blowdown before it is discharged into the Delaware River and warns personnel of an excessive amount of radioactivity (greater than Technical Specification limits) being released to the environment. DEVIATION None REFERENCES NUMARC NESP-007, AAI.2 Off-Site Dose Calculation Manual, Section 1.0 - Liquid Effluents HC.RP-AR.SP-OOOI(Q), Radiation Monitoring System Alarm Response EAL - 6.2.2 Rev. 00 Page 2 of 2

HCGS EAL/RALTechnical Basis 6.0 Radiological Releases/Occurrences 6.3 In-Plant Radiation Occurrences UNUSUAL EVENT - 6.3.1 IC Unplanned Increase in Plant Radiation EAL Unplanned rise in radiation levels inside the Protected Area~ 1000 times normal as indicated by EITHER one of the following:

Permanent or portable Area Radiation Monitors
General Area Radiological Survey OPERATIONAL CONDITION - All BASIS An Unplanned increase (rise) in radiation levels within the Protected Area by a factor of 1000 times over normal represent a degradation in the control of radioactive material and a potential degradation in the level of safety of the plant. Unplanned is defined as those events or conditions which are not associated with a planned evolution, such that radiation levels are increasing in an uncontrolled manner. This condition specifically represents an uncontrolled rise in radiation levels within the Protected Area. Planned evolutions which cause elevated radiation levels do not warrant classification under this EAL.

Barrier Analysis NIA ESCALATION CRITERIA Emergency Classification will escalate to an Alert (6.3.2.a) when radiation levels rise to a level that would impede access to areas required for the safe shutdown of the plant. DISCUSSION Normal level is considered as the highest reading in the past 24-hours excluding current peak values. RM-11 computer trends, RMS strip charts, and/or SPDS can be used to confirm these values. EAL- 6.3.1 Rev. 00 Page 1 of 2

HCGS EAL/RALTechnical Basis Examples of a planned evolution that results in rising radiation levels within the Protected Area include, but are not limited to: Radiography Lifting of the Reactor Vessel Moisture Separator I Dryer during Refuel Operations Performance of a TIP trace Relocation of radioactive materials, including radioactive waste DEVIATION NUMARC IC AU2 includes unexpected increases in Airborne concentration in addition to plant radiation. The corresponding Hope Creek IC does not address Airborne concentration, since an increase in Airborne concentration is not addressed in the example EALs or the basis for the Unusual Event or Alert. Apparently, the Airborne concentration example EAL was deleted by NUMARC, but the corresponding IC was overlooked. "REFERENCES NUMARC NESP-007, AU2.4 EAL-*6.3.1 Rev. 00 Page 2 of 2

HCGS EAL/RALTechnical Basis 6.0 Radiological Releases/Occurrences 6.3 In-Plant Radiation Occurrences ALERT - 6.3.2.a 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 EAL Unplanned Dose Rates 2::, 2000 mRem/hr in ANY area of the plant which requires ACCESS to maintain plant safety functions (EXCLUDING the Main Control Room and CAS) as indicated by EITHER one of the following:

Permanent or portable Area Radiation Monitors
General Area Radiological Survey OPERATIONAL CONDITION - All BASIS An Unplanned Dose Rate of 2000 mRem/hr or greater in ANY area of the plant which requires ACCESS to maintain plant safety functions, warrants declaration of an Alert, due to the impaired ability to operate the required plant equipment. The 2000 mRem/hr is not intended to be above the pre-existing background, but includes the pre-existing background. Unplanned is defined as those events or conditions which are not associated with a planned evolution, such that radiation levels are rising in an uncontrolled manner. The Dose Rate threshold of 2000 mRem/hr was chosen based upon NC.NA-AP.ZZ-0024, Radiation Protection Program Administrative Dose Limits and Extension criteria which requires Senior Radiation Protection Supervisor approval prior to exceeding 2000 mRem/yr TEDE. Radiation levels could be indicated by either ARM or radiological survey.

Barrier Analysis NIA EAL- 6.3.2.a Rev. 00 Page 1 of 2

HCGS EALIRALTechnical Basis ESCALATION CRITERIA Emergency Classification will escalate to a Site Area Emergency when loss of control of radioactive materials causes significant offsite doses. DISCUSSION Emergency Coordinator judgment must be used, based on existing plant conditions, to dP~ermine areas that contain systems that are required to be operated manually, or require local surveillances to assure reliable support of safe plant operation for the conditions that exist. Areas having equipment that must be operated locally during an accident and areas along associated access routes that require HP coverage and continuous update of changing radiological conditions satisfy the definition of this condition. Areas of the Plant which require access following an accident to maintain plant safety functions include, but are not limited to: Reactor Core Isolation Cooling (RCIC) system areas Standby Liquid Control (SLC) system areas Residual Heat Removal (RHR) system areas Emergency Diesel Generators (EDGs) and adjacent Areas Service Water System (SWS) areas Station Auxiliary Cooling (SACS) system areas Areas covered in the HC.OP-EO.ZZ-300's (300 series EOPs) DEVIATION None REFERENCES NUMARC NESP-007, AA3.2 NC.NA-AP.ZZ-0024(Q), Radiation Protection Program EAL - 6.3 .2.a Rev. 00 Page 2 of 2

HCGS EALIRALTechnical Basis 6.0 Radiological Releases/Occurrences 6.3 In-Plant Radiation Occurrences ALERT - 6.3.2.b IC Release of Radioactive Material or increases in Radiation Levels within the facility that impedes operation of s) stems required to maintain safe operations or to establish or maintain Cold Shutdown EAL Unplanned Dose Rates~ 15 mRem/hr in EITHER one of the following:

Main Control Room
Security Central Alarm Station (CAS)

OPERATIONAL CONDITION - All BASIS An Unplanned Dose Rate of greater than or equal to 15 mRem/hr represents a condition which would jeopardize continuous occupancy of the Main Control Room or Security CAS, and warrants declaration of an Alert. It is the impaired ability to operate the plant that results in the actual or potential substantial degradation of the level of safety of the plant. In addition, unplanned rises in plant radiation levels represent a degradation in the control of radioactive materials and represent a degradation in the level of safety of the plant. Unplanned is defined as those events or conditions which are not associated with a planned evolution, such that radiation levels are rising in an uncontrolled manner. Barrier Analysis NIA ESCALATION CRITERIA Emergency Classification will escalate to a Site Area Emergency when loss of control of radioactive materials causes significant off-site doses. EAL - 6.3.2.b Rev. 00 Page 1 of 2

HCGS EALIRALTechnical Basis DISCUSSION The Control Room and Security Central Alarm Station general area radiation level threshold is set at 15 mRem/hr and was chosen because continuous occupancy is required. This is consistent with General Design Criteria 19, which addresses continuous occupancy of the Control Room for 30 days after an accident. The Security Secondary Alarm Station (SAS) was excluded because it is fully redundant to the Security CAS. For a radiological event, SAS could be evacuated, with all Security functions performed by the CAS. Events which require Main Control Room evacuation will be classified per ECG Section 8. DEVIATION None REFERENCES NUMARC NESP-007, AAJ.1 IOCFRSO NUREG/CR-4982 NRC Information Notice 08 EAL - G.3.2.b Rev. 00 Page 2 of 2

HCGS EAL/RALTechnical Basis 6.0 Radiological Releases/Occurrences 6.4 Irradiated Fuel Event UNUSUAL EVENT- 6.4.1.a IC Unplanned Increase in Plant Radiation EAL Uncontrolled water level drop in the Reactor Cavity as indicated by EITHER one of the following:

Visual Observation
Reactor Water Level Shutdown Range Indicator 1BBLl-R605 OPERATIONAL CONDITION - 5 BASIS An Uncontrolled lowering of Reactor Cavity Level during Refueling (Operational Condition 5) represents a condition which can result in rising radiation levels, due to the loss of radiation shielding, if the Reactor Cavity level drop can not be terminated. This event has a long lead time relative to potential for radiological release outside the site boundary, thus the impact to publiC health and safety is very low. Uncontrolled means that the level drop can not be terminated.

Determination of an uncontrolled level drop is made through either Visual Observation or indication in the Main Control Room. Visual Observation is the preferred method, whenever possible, however it is NOT intended that an individual must be dispatched for classification purposes, if the existing radiation level rise trend prevents personnel from accessing the Refuel Floor, or if cameras are available to remotely verify the condition. In the event visual observation is not available by any means, then Main Control Room indication should be used. Barrier Analysis NIA EAL - 6.4. l .a Rev. 00 Page I of 2

HCGS EAL/RALTechnical Basis ESCALATION CRITERIA Emergency Classification will escalate to an Alert as a result of uncovery of a fuel assembly and/or indication of high radiation levels on the refueling floor. DISCUSSION During Refueling operations, the RPV is flooded and RPV level indication is monitored on the shutdown instrument range. Limitations on evolutions on with a potential for draining the RPV are imposed when refueling is in progress. Lowering of RPV level may result in the loss of Shutdown Cooling ifRPV level continues to lower unchecked. This may result in the loss of decay heat removal from the fuel contained in the RPV. Technical Specifications requires at least 22 feet 2 inches of water be maintained over the top of the reactor pressure vessel flange while in Operating Condition 5 and either fuel assemblies are being handled or the fuel assemblies seated within the reactor vessel are irradiated. The Technical Specification minimum water level in the Reactor Vessel under these conditions is based on the minimum water level required to remove 99% of the assumed 10% iodine gap activity that would be released from the rupture of an irradiated fuel assembly. DEVIATION

1) NUMARC states that this EAL will be applicable in all modes of operation. In other than Operational Condition 5, the RPV head will be fully tensioned, and lowering of vessel level would be classified by EALs in Section 3.0, Fission Product Barriers, or Section 8.1, Loss ofHeat Removal Capability.

2) NUMARC IC AU2 includes unexpected increases in Airborne concentration in addition to plant radiation. The corresponding Hope Creek IC does not address Airborne concentration, since an increase in Airborne concentration is not addressed in the example EALs or the basis for the Unusual Event or Alert. Apparently, the Airborne concentration example EAL was deleted by NUMARC, but the corresponding IC was overlooked.

REFERENCES NUMARC NESP-007, AU2. I HC.OP-AB.ZZ.0142 (Q), Loss of Shutdown Cooling HC.OP-AB.ZZ-0144 (Q), Loss of Fuel Pool Inventory/Cooling HC.OP-AB.ZZ-0101 (Q), Irradiated Fuel Damage HC.OP-AB.ZZ-126 (Q), Abnormal Release of Gaseous Radioactivity HCGS Technical Specifications Section 3/4 9.8 EAL - 6.4.1.a Rev. 00 Page 2 of2

HCGS EAL/RALTechnical Basis 6.0 Radiological Releases/Occurrences 6.4 Irradiated Fuel Event UNUSUAL EVENT- 6.4.1.b IC Unplanned Increase in Plant Radiation EAL Uncontrolled water level drop in the Spent Fuel Pool as indicated by Valid Fuel Pool Low Level Alarm Condition Visual Observation OPERATIONAL CONDITION - All BASIS An Uncontrolled drop in Spent Fuel Pool Level represents a condition which can result in rising radiation levels, due to the loss of radiation shielding, ifthe Spent Fuel Pool level drop can not be terminated. This event has a long lead time relative to potential for radiological release outside the site boundary, thus the impact to public health and safety is very low. Uncontrolled means that the level drop can not be terminated. Determination of an uncontrolled level drop is made through receipt of the Spent Fuel Pool Low Level Alarm in the Main Control Room and Visual Observation. Barrier Analysis NIA ESCALATION CRITERIA Emergency Classification will escalate to an Alert as a result of uncovery of irradiated fuel as indicated by high radiation levels on the refueling. EAL - 6.4.1.b Rev. 00 Page I of 2

HCGS EALIRALTechnical Basis DISCUSSION Normal Spent Fuel Pool level is at 40' of water in the pool. This level provides approximately 25' of water above the top of fuel stored in pool, and 9' of water above fuel in transit. The low level alarm is set at 39' 9". This is above, but approaching the Technical Specification minimum required water level of 23 feet over the top of irradiated fuel assemblies seated in the spent fuel pool storage racks. The Technical Specification minimum water level in the Spent Fuel Pool is based on the minimum inventory and level required to remove 99% of the assumed 10% iodine gap activity that would be released from the rupture of an irradiated fuel assembly. To prevent accidental draining of the Spent Fuel Pool, no piping connections are made to the fuel pool below the normal water level. The spent fuel pool cooling water return lines are provided with vacuum breakers to prevent water from being siphoned out of the fuel pool should a break occur in one of these lines. The skimmer surge tanks receive the overflow from the spent fuel pool and serve as the suction source to the fuel pool cooling pumps. Lowering of level in the skimmer surge tank will result in isolation of the pool filter demineralizers. This will result in the loss of the fuel pool cooling pumps. Subsequent heating of the water in the spent fuel pool may occur depending on the heat load present. DEVIATION NUMARC IC AU2 includes unexpected increases in Airborne concentration in addition to plant radiation. The corresponding Hope Creek IC does not address Airborne concentration, since an increase in Airborne concentration is not addressed in the example EALs or the basis for the Unusual Event or Alert. Apparently, the Airborne concentration example EAL was deleted by NUMARC, but the corresponding IC was overlooked. REFERENCES NUMARC NESP-007, AU2.2 HC.OP-AR.ZZ-0014(Q), Annunciator Response Procedures, Window D3-A5 (D3834) HC.OP-AB.ZZ-0144 (Q), Loss of Fuel Pool Inventory/Cooling HC.OP-AB.ZZ-0101 (Q), Irradiated Fuel Damage HC.OP-AB.ZZ-126 (Q), Abnormal Release of Gaseous Radioactivity HCGS Technical Specifications Section 3/4 9.9 HCGS UFSAR, Section 9.2.2.2 EAL - 6.4. l.b Rev. 00 Page 2of2

HCGS EAL/RALTechnical Basis 6.0 Radiological Releases/Occurrences 6.4 Irradiated Fuel Event ALERT - 6.4.2.a IC Major Damage to Irradiated Fuel EAL Major Damage to Irradiated Fuel has occurred Valid High Alarm received from ANY one of the following RMS channels:

Refuel Floor Exhaust Channel A (9RX627)
Refuel Floor Exhaust Channel B (9RX628)
Refuel Floor Exhaust Channel C (9RX629)

OPERATIONAL CONDITION - All BASIS Damage to an irradiated fuel bundle that results in a High Refuel Floor Exhaust Radiation Monitor alarm warrants declaration of an Alert, due to the potential for an uncontrolled offsite release exceeding the Technical Specification limit. The intent of this EAL is to classify those events that result in the actual release of fission products from an irradiated Fuel Bundle, due to physical damage. Events that result in rising radiation levels due to shine, as a result of lowered shielding, but do not involv*! a release of fission products, should not be classified under this EAL, but should be classified EAL 6.3.2.d, when those conditions exist. Major Damage is defined as physical damage to an Irradiated Fuel Bundle that results from either dropping or physical contact with other components in the Fuel Pool, such that the magnitude of the damage specifically results in actuation of a Refuel Floor Exhaust High Radiation Alarm. Valid is defined as the High alarm occurring as a result of the damage to the irradiated fuel bundle which results in an actual release of fission products from the cladding. EAL - 6.4.2.a Rev. 00 Page I of2

HCGS EAL/RALTechnical Basis Barrier Analysis NIA ESCALATION CRITERIA Emergency Classification will escalate to a Site Area Emergency when loss of control of radioactive materials causes significant offsite doses. DISCUSSION The Refuel Floor Exhaust Rad Monitors are Process Monitors and are designed to detect a release of Fission Products to the Reactor Building atmosphere. Hence, they are included as part of the EAL threshold, to confirm the magnitude of damage to an irradiated fuel bundle. These monitors can also react as Area Radiation Monitors, in the event of rising radiation levels due to lowered shielding, as would occur during a loss of Fuel Pool inventory event. It is important to distinguish between the cause for rising radiation levels when classifying an event under this EAL. DEVIATION None REFERENCES NUMARC NESP-007, AA2. l HC.OP-SO.SM-000 I (Q), Isolation Systems Operation HC.OP-AB.ZZ-OI 16(Q), Containment Isolations and Recovery from an Isolation HC.RP-AR.SP-OOOI(Q), Radiation Monitoring System Alarm Response, Att. 54, 55, 56 HCGS Technical Specifications, 3.3.2 Table 3.3.2-2 HCGS-UFSAR, Section 11.5.2 NUREG/CR-4982 NRC Information Notice 08 EAL - 6.4.2.a Rev.00 Page 2of2

HCGS EALIRALTechnical Basis 6.0 Radiological Releases/Occurrences 6.4 Irradiated Fuel Event ALERT - 6.4.2.b/6.4.2.c IC Events that have or may result in uncovering Irradiated Fuel outside the Reactor Vessel EAL EITHER one of the following:

Unplanned rise on ANY one of the following Area Rad Monitors or by general area rad survey indicates~ 2000 mRem/hr:
Spent Fuel Storage Pool Area (9RX707)
New Fuel Criticality Storage Channel A (9RX612)
New Fuel Criticality Storage Channel B (9RX6 I 3)
Visual observation of Irradiated Fuel uncovered OPERATIONAL CONDITION - All BASIS An Unplanned Dose Rate of2000 mRem/hr as indicated on any of the Refuel Floor Area Radiation Monitors (ARMs) warrants declaration of an Alert, as dose rates of this magnitude could be the result of a loss of shielding of irradiated Fuel Bundles or possible damage to an irradiated Fuel Bundle. Offsite doses during these accidents would be well below the EPA Protective Action Guidelines and the classification as an Alert is therefore appropriate.

The intent of these EALs is to classify those events that result in rising dose rates on the Refuel Floor. Specifically, those events that result in rising radiation levels due to shine, as a result of lowered shielding, but do not involve a release of fission products should be classified under this EAL. Those events that result in physical damage to an irradiated fuel assembly and are accompanied by rising radiation levels should not be classified under this EAL, but should be classified EAL 6.3.2.c, when those conditions exist. Unplanned is defined as those events or conditions which are not associated with a planned evolution, such as lifting of the Reactor Vessel Internals, that results in radiation levels are rising in an uncontrolled manner. The Dose Rate threshold of 2000 mRem/hr was chosen based upon EAL - 6.4.2.b/6.4.2.c Rev.00 Page I of2

HCGS EAL/RALTechnical Basis NC.NA-AP.ZZ-0024, Radiation Protection Program Administrative Dose Limits and Extension criteria which requires Senior Radiation Protection Supervisor approval prior to exceeding 2000 mRem/yr TEDE. This value is low enough to ensure classification of an Alert before personnel access is severely hampered and high enough to allow any rise in normal radiation level, by a factor of I 000, to be classified as an Unusual Event per EAL 6.3. I .a. Radiation levels could be. indicated by either ARMs or radiological survey. Uncovered irradiated fuel will result in Onsite dose rates rising significantly. Barrier Analysis NIA ESCALATION CRITERIA Emergency Classification will escalate to a Site Area Emergency when loss of control of radioactive materials causes significant offsite doses. DISCUSSION The Refuel Floor ARMs are designed to detect rising radiation levels on the Refuel Floor. Hence, they are included as part of the EAL threshold, to determine the magnitude of a loss of shielding to irradiated Fuel Bundles.

Actual Damage to an irradiated fuel bundle will also cause a rise in these ARMs, however the Refuel Floor Exhaust Rad Monitors are specifically designed to detect the actual release of fission products to the atmosphere. It is important to distinguish between the possible causes for rising radiation levels when classifying an event under these EALs.

DEVIATION None REFERENCES NUMARC NESP-007, AA2.3, AA2.4 HCGS Technical Specifications, 3.3.7.1, Table 3.3.7.1-1 HC.RP-AR. SP-000 I (Q), Radiation Monitoring System Alarm Response, Att. 41, 42, 77 NUREG-1229, Source Term Estimation During Incident Response to Severe Nuclear Power Plant Accidents EPA 400-R-92-001, Manual of Protective Action Guides and Protective Actions For Nuclear Incidents NRC Information Notice 08 EAL - 6.4.2.b/6.4.2.c Rev. 00 Page 2 of2 J

HCGS EALIRALTechnical Basis 7.0 Electrical Power. i

                                                                         -~\
                                                                                                   \!

7.1 Loss of AC Power Ca~a:t>ilities _,,.----

                                                             \ .,       <£5 UNUSUAL EVENT- 7.1.1                                            l
                                                               ~.~~~~...,._.... ..:-...,. ......--

IC Loss of All Offsite Power to Vital Buses for greater than 15 minutes EAL Unplanned Loss of Power from Station Service Transformers 1AX501 AND 1BX501 to ALL 4.16 KV Vital Buses

   >15 minutes have elapsed OPERATIONAL CONDITION - All BASIS An Unplanned Loss of Power from Station Service Transformers 1AX501 AND IBX501 (Offsite Power Sources) to the 4.16 KV Vital Buses for greater than 15 minutes, reduces required plant redundancy and potentially degrades the level of safety by increasing plant vulnerability to a complete loss of all Vital AC power. Reliance on the EDGs to energize the Vital Buses represents a significantly abnormal condition.

The intent of the EAL is to classify an Unusual Event when the EDGs are being used to energize their respective Vital Buses, due to a loss of the offsite power sources. In the case where one or more EDGs are unavailable or fail to start for any reason, following the loss of the offsite power sources, an Unusual Event is warranted until only one Onsite or Offsite Power Source remains energized, such that the loss of this energized source would result in a complete loss of all 4 .16 KV Vital Power. 15 minutes was chosen to exclude transient or momentary power losses and to allow restoration of available sources. Unplanned is defined as the loss not being the result of planned or scheduled maintenance activities. Although no fission product barriers are directly affected by the loss of the offsite power sources to the Vital 4.16 KV buses, the heat addition to the Primary Containment combined with heat removal capability dependent on Emergency Diesel Generator operation, warrants classification as an Unusual Event, since it is potential precursor to more serious conditions. EAL - 7.1.1 Rev. 00 Page 1 of 3

HCGS EAL/RALTechnical Basis Barrier Analysis NIA ESCALATION CRITERIA Emergency Classification will escalate to an Alert based on a I,.oss of Offsite Power to Vital 4.16 KV buses coincident with Om.;te AC power being reduced to a single Vital 4.16 KV Bus. (Operational Conditions 1, 2, and 3); or having a Loss of all Offsite and Onsite AC power in Operational Conditions 4 or 5. DISCUSSION Hope Creek normally has three physically separate , independent 500 KV transmission lines, connecting the Hope Creek 500 KV Switchyard with the Offsite Power Distribution Network (PJM). The three sources are as follows: 500 KV Hope Creek - Salem Crosstie line. The Keeney Line, referred to as the 5015 line, is 30.1 mile tie to the Keeney Switching Station (located near Newark, Delaware), which feeds the 500 KV Switchyard Bus Section 3. The New Freedom Line, referred to as the 5023 line, is a 42. 9 mile tie to the New Freedom Switching Station (located northeast of Hope Creek in Camden County), which feeds the 500 KV Switchyard Bus Section 5. Power is distributed from the 500 KV Switchyard to a 13.8 KV ring bus. Station electrical loads are supplied from the 13.8 KV ring bus through 2 physically independent auxiliary power systems, via Station Service Transformers which supply Vital and Non-Vital Station Loads. Station Service Transformers 1AX501and1BX501 normally supply the 4.16 KV Vital Buses. The four 4.16 KV Vital Buses can be supplied by either 1AX501 or 1BX501. Two of the four Vital Buses are normally provided power from I AX501 with alternate power from 1BX501; the other two are normally supplied power from 1BX50 I with alternate power from 1AX501. Loss of the normal power supply to a 4.16 KV Vital Bus initiates a fast transfer (alternate feeder breaker closes) to the alternate source, provided power is available. Additionally, each 4.16 KV Vital Bus has an Emergency Diesel Generator which will automatically start and provide power to the bus in the event of a sustained loss of power to its associated Vital Bus. Additional automatic EDG starts are initiated on degraded power conditions on both 1AX50 I and I BX50 I, or under LOCA conditions (EDGs will not automatically provide power to the bus unless the bus has a sustained loss of power). EAL - 7.1.1 Rev. 00 P::gc 2 of 3

HCGS EALIRALTechnical Basis DEVIATION None REFERENCES NUMARC NESP-007, SUI HC.OP-AB.ZZ-0135 (Q), Station Blackout//Loss ofOffsite Power//Diesel Generator Malfunction HC.OP-EO.ZZ-0100 (Q)-FC, Reactor Scram HC.OP-EO.ZZ-0102 (Q)-FC, Primary Containment Control HCGS Technical Specifications 3/4.8, Electrical Power Systems EAL- 7. l.l Rev. 00 Page 3 of 3

HCGS EALIRALTechnical Basis 7.0 Electrical Power 7.1 Loss of AC Power Capabilities ALERT - 7.1.2.a IC AC power capability to Vital Buses reduced to a Single Power Source for greater than 15 minutes such that any additional single failure would result in a complete loss of all 4.16 KV Vital Buses EAL Loss of 4.16 KV Vital Bus Power Sources (Offsite and Onsite) which results in the availability of ONLY one 4.16 KV Vital Bus Power Source (Offsite or Onsite)

    > 15 minutes have elapsed OPERATIONAL CONDITION - I, 2, 3 BASIS A degradation of the six 4.16 KV Vital Bus Power Sources, which consist of the Offsite power sources (IAXSOI and IBXSOI) and the Onsite power sources (4 EDGs), available to the 4.16 KV Vital Buses, such that a loss of any additional single energized source would result in a complete loss of all 4.16 KV Vital Power, represents a significant challenge to plant safety and is classified under this EAL. These conditions could occur as a result of a Loss of the Offsite power sources with concurrent failure of all but one EDG to supply power to its Vital Bus, or due to a failure of all EDGs concurrent with the Offsite power sources reduced to a single source (even though all 4.16 KV Vital Buses may still be energized).

These conditions reduce redundancy and potentially degrade the level of safety by increasing plant vulnerability to a complete Loss of Vital AC power. The intent of this EAL is to classify an Alert in those conditions in which a loss of a single power source to the 4.16 KV Vital Buses would result in the loss of All 4.16 KV Vital power. Availability is defined as a power source that can be aligned to provide power to the bus within 15 minutes. This includes the power source, as well as, all required breakers needed to provide power. 15 minutes was chosen to exclude transient or momentary power losses. EAL - 7.1.2.a Rev. 00 Page 1 of 3

HCGS EAL/RALTechnical Basis Barrier Analysis NIA ESCALATION CRITERIA Emergency Classif.cation will escalate to a Site Area Emergency based on a Loss of Power to all 4.16 KV Vital Buses for> 15 minutes. DISCUSSION Hope Creek normally has three physically separate , independent 500 KV transmission lines, connecting the Hope Creek 500 KV Switchyard with the Offsite Power Distribution Network (PIM). The three sources are as follows: 500 KV Hope Creek - Salem Crosstie line. The Keeney Line, referred to as tht: 5015 line, is 30.1 mile tie to the Keeney Switching Station (located near Newark, Delaware), which feeds the 500 KV Switchyard Bus Section 3. The New Freedom Line, referred to as the 5023 line, is a 42. 9 mile tie to the New Freedom Switching Station (located northeast of Hope Creek in Camden County), which feeds the 500 KV Switchyard Bus Section 5. Power is distributed from the 500 KV Switchyard to a 13.8 KV ring bus. Station electrical loads are supplied from the 13.8 KV ring bus through 2 physically independent auxiliary power systems, via Station Service Transformers which supply Vital and Non-Vital Station Loads. Station Service Transformers I AX50 I and IBX50 I normally supply the 4.16 KV Vital Buses. The four 4 .16 KV Vital Buses can be supplied by either I AX501 or 1BX50 I. Two of the four Vital Buses are normally provided power from IAX501 with alternate power from 1BX50I; the other two are normally supplied power from 1BX501 with alternate power from I AX50 I. Loss of the normal power supply to a 4.16 KV Vital Bus initiates a fast transfer (alternate feeder breaker closes) to the alternate source, provided power is available. Additionally, each 4.16 KV Vital Bus has an Emergency Diesel Generator which will automatically start and provide power to the bus in the event of a sustained loss of power to its associated Vital Bus. Additional automatic EOG starts are initiated on degraded power conditions on both 1AX501 and IBX501, or under LOCA conditions (EDGs will not automatically provide power to the bus unless the bus has a sustained loss of power). DEVIATION None EAL - 7.1.2.a Rev. 00 Page 2 of 3

HCGS EAL/RALTechnical Basis REFERENCES NUMARC NESP-007, SAS HC.OP-AB.ZZ-0135 (Q), Station Blackout I Loss of Offsite Power I Diesel Generator Malfunction HC.OP-EO.ZZ-0100 (Q)-FC, Reactor Scram HC.OP-EO.ZZ-0102 (Q)-FC, Primary Containment Control HCGS Technical Specifications 3/4.8, Electrical Power Systems EAL - 7.1.2.a Rev. 00 Page 3 of 3

HCGS EALIRALTechnical Basis 7.0 Electrical Power 7.1 Loss of AC Power Capabilities ALERT - 7.1.2.b IC Loss of All Offsite Power and All Onsite AC Power to 4.16 KV Vital Buses during either Cold Shutdown or Refueling for greater than 15 minutes EAL ALL 4.16 KV Vital Buses are deenergized

   > 15 minutes have elapsed OPERATIONAL CONDITION - 4, 5, Defueled BASIS A Loss of ALL 4.16 KV Vital Buses that occurs while .the plant is in either Cold Shutdown or Refueling conditions, results in a compromise of plant systems. The intent of this EAL is to classify degraded AC power events that result in a Loss of Offsite power sources (1AX501 AND 1BX501} to the 4.16 KV Vital Buses, along with a Loss of Onsite power sources (EDGs). With the plant in Cold Shutdown or Refueling, the reduced decay heat, and lower Reactor Coolant temperatures and pressures, increases the time available to restore one of the Vital Buses before Fission Product Barriers are threatened relative to classification of this condition in Operational Conditions l, 2, or 3. Thus this condition is classified as an Alert. 15 minutes was chosen to exclude transient or momentary power losses.

Barrier Analysis NIA ESCALATION CRITERIA Emergency Classification will escalate to a Site Area Emergency based on Radiological Release (EAL Section 6.0), or on the long term inability to remove Decay Heat (EAL Section 8.0). EAL - 7.1.2.b Rev. 00 Page 1 of 3

HCGS EAL/RALTechnical Basis DISCUSSION Loss of all AC power to the Vital Buses compromises all plant safety systems requiring AC electric power including RHR, ECCS, Spent Fuel Pool Cooling and Service Water. Depending on the status of power supplies to non-vital buses, some Balance of Plant systems that would assist in maintaining plant conditions (i.e. RWCU, condensate, etc.) may be unavailable. Thus, the ability to remove decay heat and control containment parameters is severely challenged. During a Loss of all AC power to the Vital Buses, all Class IE System Instruments remain powered from Class IE Uninterruptable Power Supplies (UPS), which are powered by DC power via inverters. The I 25 VDC Battery Buses will continue to supply DC power from the batteries. Battery power is limited depending on the discharge rate and predischarge condition of the battery. The ability to restore power to AC buses may eventually be threatened as battery power (DC) is depleted due to the lack of DC (control power) for AC power circuit breakers. Normally, Hope Creek has three physically separate , independent 500 KV transmission lines, connecting the Hope Creek 500 KV Switchyard with the Offsite Power Distribution Network (PJM). The three sources are as follows: 500 KV Hope Creek - Salem Crosstie line. The Keeney Line, referred to as the 5015 line, is 30. l mile tie to the Keeney Switching Station (located near Newark, Delaware), which feeds the 500 KV Switchyard Bus Section 3. The New Freedom Line, referred to as the 5023 line, is a 42.9 mile tie to the New Freedom Switching Station (located northeast of Hope Creek in Camden County), which feeds the 500 KV Switchyard Bus Section 5. Power is distributed from the 500 KV Switchyard to a 13.8 KV ring bus. Station electrical loads are supplied from the 13.8 KV ring bus through 2 physically independent auxiliary power systems, via Station Service Transformers which supply Vital and Non-Vital Station Loads. Station Service Transformers l AX50 l and IBX50 l normally supply the 4.16 KV Vital Buses. The four 4 16 KV Vital Buses can be supplied by either I AX50 I or IBX501. Two of the four Vital Buses are normally provided power from 1AX50 l with alternate power from I BX501; the other two are normally supplied power from IBX501 with alternate power from 1AX501. Loss of the normal power supply to a 4.16 KV Vital Bus initiates a fast transfer (alternate feeder breaker closes) to the alternate source, provided power is available. Additionally, each 4.16 KV Vital Bus has an Emergency Diesel Generator which will automatically start and provide power to the bus in the event of a sustained loss of power to its associated Vital Bus. Additional automatic EOG starts are initiated on degraded power conditions on both I AX501 and 1BX50 I, or under LOCA conditions (EDGs will not automatically provide power to the bus unless the bus has a sustained loss of power). EAL - 7.1.2.b Rev. 00 Page 2 of 3

HCGS EAL!RALTechnical Basis DEVIATION None REFERENCES NUMARCNESP-007, SAi HC.OP-AB.ZZ-0135(Q), Station Blackout//Loss of Offsite Power//Diesel Generator Malfunction HC.OP-EO.ZZ-OIOO(Q)-FC, Reactor Scram HC.OP-EO.ZZ-0102(Q)-FC, Primary Containment Control HCGS Technical Specifications 3/4.8, Electrical Power Systems EAL - 7.1.2.b Rev.00 Page 3 of 3

HCGS EAL/RALTechnical Basis 7.0 Electrical Power 7.1 Loss of AC Power Capabilities SITE AREA EMERGENCY - 7.1.3 IC Loss of All Offsite Power and All Onsite AC Power to All Vital AC Buses during either Power Operation, Startup or Hot Shutdown for greater than 15 minutes EAL ALL 4.16 KV Vital Buses are deenergized

   > 15 minutes have elapsed OPERATIONAL CONDITION - I, 2, 3 BASIS A Loss of ALL 4.16 KV Vital Buses that occurs while _the plant is in either Power Operation, Startup or Hot Shutdown warrants declaration of a Site Area Emergency due to the compromise to all plant safety systems. The intent of this EAL is to classify degraded AC power events that result in a loss ofOffsite power source (IAX501 AND 1BX501) to the 4.16 KV Vital Buses, along with a Loss of Onsite power sources (EDGs). Declaration under this EAL should only occur for a loss of ALL 4.16 KV Vital Buses. Prolonged loss of Vital AC power may cause Core uncovery and the inability to remove heat from the containment. 15 minutes was chosen to exclude transient or momentary power losses.

Barrier Analysis Prolonged loss of AC power has the potential to cause a potential loss or loss of the Fission Product Barriers. ESCALATION CRITERIA Emergency Classification will be escalate to a General Emergency if the power loss is extended beyond 4 hours, or on loss of Fission Product Barriers per EAL Section 3.0. EAL- 7.1.3 Rev. 00 P:ige 1 of 3

HCGS EALIRALTechnical Basis DISCUSSION Loss of all AC power to the Vital Buses compromises all plant safety systems requiring AC electric power including RHR, ECCS, Spent Fuel Pool Cooling and Service Water. Depending on the status of power supplies to non-vital buses, some Balance of Plant systems that would assist in maintaining plant conditions (i.e. RWCU, condensate, etc.) may be unavailable. Thus, the ability to remove decay heat and control containment parameters is severely challenged. During a Loss of all AC power to the Vital Buses, all Class IE System Instruments remain powered from Class IE Uninterruptable Power Supplies (UPS), which are powered by DC power via inverters. The 125 VDC Battery Buses will continue to supply DC power from the batteries. Battery power is limited depending on the discharge rate and predischarge condition of the battery. The ability to restore power to AC buses may eventually be threatened as battery power (DC) is depleted due to the lack of DC (control power) for AC power circuit breakers. Normally, Hope Creek has three physically separate, independent 500 KV transmission lines, connecting the Hope Creek 500 KV Switchyard with the Offsite Power Distribution Network (PJM). The three sources are as follows: 500 KV Hope Creek - Salem Crosstie line. The Keeney Line, referred to as the 5015 line, is 30.1 mile tie to the Keeney Switching Station (located near Newark, Delaware), which feeds the 500 KV Switchyard Bus Section 3. The New Freedom Line, referred to as the 5023 line, is a 42.9 mile tie to the New Freedom Switching Station (located northeast of Hope Creek in Camden County), which feeds the 500 KV Switchyard Bus Section 5. Power is distributed from the 500 KV Switchyard to a 13.8 KV ring bus. Station electrical loads are supplied from the 13. 8 KV ring bus through 2 physically independent auxiliary power systems, via Station Service Transformers which supply Vital and Non-Vital Station Loads. Station Service Transformers lAXSOl and lBXSOl normally supply the 4.16 KV Vital Buses. The four 4.16 KV Vital Buses can be supplied by either lAXSOJ or lBXSOI. Two of the four Vital Buses are normally provided power from lAXSOl with alternate power from lBXSOl; the other two are normally supplied power from IBXSOJ with alternate power from 1AX501. Loss of the normal power supply to a 4.16 KV Vital Bus initiates a fast transfer (alternate feeder breaker closes) to the alternate source. Additionally, each 4. 16 KV Vital Bus has an Emergency Diesel Generator which will automatically start and provide power to the bus in the event of a sustained loss of power to its associated Vital Bus. Additional automatic EDG starts are initiated on degraded power conditions on both 1AXSO I and lBXSO I, or under LOCA conditions (EDGs will not automatically provide power to the bus unless the bus has a sustained lost of power). EAL- 7.1.3 Rev. 00 Page 2 o: 3

HCGS EAL/RALTechnical Basis Under a Loss of Vital AC Power condition, operation and control of plant systems is guided by the Station Blackout/Loss of Offsite Power/Diesel Generator Malfunction Abnormal Operating Procedure. Successful coping maintains the following key parameters within given acceptable limits:

1. Reactor water level > (T AF)

2. Suppression pool level low enough to prevent HPCI and/or RCIC stt,am exhaust line flooding

3. Reactor pressure high enough to maintain HPCI and RCIC operable

4. Containment pressure< design limit

5. Torus temperature <-*design limits (HPCI/RCIC lube oil temperature concern when suction aligned to suppression pool)

6. Drywell temperature below design limits RCIC and HPCI operability is dependent on the availability of 125/250 VDC power. The parameters listed above can be maintained as long as battery power remains available. Battery power is limited depending on the discharge rate and predischarge condition of the battery. The HCGS IPE assumes that the batteries will be available for four hours, even though the design battery depletion time is six hours. Additionally, the loss of ventilation to the HPCI and RCIC turbine areas may result in a system isolation due to elevated temperatures.

Other than HPCI and/or RCIC, additional inventory makeup may be possible by using the diesel driven fire pump to inject water (at low pressure), to the RPV via, the RHR/LPCI system. This may require RPV depressurization using the SRVs, which also require 125 VDC power. DEVIATION None REFERENCES NUMARC NESP-007, SS I HC.OP-AB.ZZ-0135 (Q), Station Blackout I Loss of Offsite Power I Diesel Generator Malfunction HC.OP-EO.ZZ-0100 (Q)-FC, Reactor Scram HC.OP-EO.ZZ-0102 (Q)-FC, Primary Containment Control HCGS Technical Specifications Section 3/4.8, Electrical Power Systems EAL - 7.1.3 Rev. 00 Page 3 of 3

HCGS EAL/RALTechnical Basis 7.0 Electrical Power 7.1 Loss of AC Power Capabilities GENERAL EMERGENCY- 7.1.4.a IC Prolonged Loss of All Offsite and Onsite AC Power to All Vital AC Buses EAL ALL 4.16 KV Vital Buses are deenergized Restoration of Power to at least one 4.16 KV Vital Bus within 4 hours is NOT likely OPERATIONAL CONDITION - 1, 2, 3 BASIS A Loss of ALL 4.16 KV Vital Buses for a prolonged period of time(> 4 Hours) represents a compromise to all plant safety systems. The intent of this EAL is to classify degraded AC power events that result in a Loss of offsite power source (IAX501AND1BX501) to the 4.16 KV Vital Buses, along with a Loss of Onsite power sources (EDGs) for greater than 4 hours. Prolonged loss of Vital AC power may cause Core uncovery and the inability to remove heat from the containment. 4 Hours is based on the assumptions of the Station Blackout Coping Studies for Hope Creek. Beyond the four hour window, Reactor injection capability may no longer be available, and degradation in core cooling will commence. However, a General Emergency should be declared before 4 hours if it can be determined that the power loss cannot be recovered within 4 hours, or if potential loss or loss of fission product barriers is imminent. Barrier Analysis Although not directly related to Fission Product Barriers, these events will eventually result in the loss of all three barriers if power cannot be restored. In addition, the extent of the loss of power will result in degraded monitoring capability. It is therefore important in such events to closely monitor the Fission Product Barriers and use judgment related to the IMMINENT Loss or Potential Loss of barriers as directed in EAL Section 3.0 EAL - 7.1.4.a Rev.00 Page 1 of 3

HCGS EAL/RALTechnical Basis ESCALATION CRITERIA NIA DISCUSSION 10 CFR 50.2 defines a station blackout (SBO) as complete loss of AC power to Vital AND Non-Vital buses. Loss of all AC power to the Vital Buses compromises all plant safety systems requiring AC electric power including RHR, ECCS, Spent Fuel Pool Cooling and Service Water. Depending on the status of power supplies to non-vital buses, some Balance of Plant systems that would assist in maintaining plant conditions (i.e. RWCU, condensate, etc.) may be unavailable. Thus, the ability to remove decay heat and control containment parameters is severely challenged. During a Loss of all AC power to the Vital Buses, all Class IE System Instruments remain powered from Class IE Uninterruptable Power Supplies (UPS), which are powered by DC power via inverters. The 125 VDC Battery Buses will continue to supply DC power from the batteries. Battery power is limited depending on the discharge rate and predischarge condition of the battery. The ability to restore power to AC buses may eventually be threatened as battery power (DC) is depleted due to the lack of DC (control power) for AC power circuit breakers. Under a Loss of Vital AC Power condition, operation and control of plant systems is guided by the Station Blackout//Loss of Offsite Power//Diesel Generator Malfunction Abnormal Operating Procedure. Successful coping maintains the following key parameters within given acceptable limits:

1. Reactor water level > (T AF)

2. Suppression pool level low enough to prevent HPCI and/or RCIC steam exhaust line flooding

3. Reactor pressure high enough to maintain HPCI and RCIC operable

4. Containment pressure < design limit

5. Torus temperature< design limits (HPCI/RCIC lube oil temperature concern when suction aligned to suppression pool)

6. Drywell temperature below design limits RCIC and HPCI operability is dependent on the availability of 125/250 VDC power. The parameters listed above can be maintained as long as battery power remains available. Battery power is limited depending on the discharge rate and predischarge condition of the battery. The HCGS IPE assumes (based on the Coping Study) that the batteries will be available for four hours, even though the design battery depletion time is six hours. Additionally, the loss of ventilation to the HPCI and RCIC turbine areas may result in a system isolation due to elevated temperatures.

Other than HPCI and/or RCIC, additional inventory makeup may be possible by using the diesel driven fire pump to inject water (at low pressure), to the RPV via, the RHR/LPCI system. This may require RPV depressurization using the SRVs, which also require 125 VDC power. EAL - 7.1.4.a Rev. 00 Page 2 of 3

HCGS EAL/RALTechnical Basis The likelihood of restoring at least one emergency bus should be based on a realistic appraisal of the situation since a delay in an upgrade decision based on only a chance of mitigating the event could result in a loss of valuable time in preparing and implementing public protective actions. In addition, under these conditions, fission product barrier monitoring capability may be degraded. Although it may be difficult to predict when power can be restored, it is necessary to give the Emergency Coordinator a reasonable idea of how quickly he may need to declare a General Emergency based on two major considerations: I. Are there any present indications that core cooling is already degraded to the point that loss or potential loss of fission product barriers is imminent?

2. If there are no present indications of such core cooling degradation, how likely is it that power can be restored in time to assure that a loss of two barriers with a potential loss of the third barrier can be prevented?

It is estimated that several hours are required to fully evacuate the l 0 mile EPZ. Taking into consideration the above factors, declaring a General Emergency leaves sufficient time for the offsite authorities to implement Protective Actions well before a radioactive release would occur while providing sufficient time for on-site and off-site mitigation activities to restore AC power. DEVIATION None REFERENCES NUMARC NESP-007, SG I Station Blackout Coping Studies HC.OP-EO.ZZ-0100 (Q)-FC, Reactor Scram HC.OP-EO.ZZ-0102 (Q)-FC, Primary Containment Control HC.OP-EO.ZZ-0104 (Q)-FC, Radioactive Release Control HC.OP-AB.ZZ-0135 (Q). Station Blackout I Loss ofOffsite Power I Diesel Generator Malfunction HCGS Technical Specifications Section 3/4.8, Electrical Power Sy:.;tems HCGS Individual Plant Evaluation, Section 3.1.1.4.6, 3.1.2.1.6 EAL - 7.1.4.a Rev. 00 Page 3 of 3

HCGS EAL/RALTechnical Basis 7.0 Electrical Power 7.1 Loss of AC Power Capabilities GENERAL EMERGENCY - 7.1.4.b IC Prolonged Loss of All Offsite and Onsite AC Power to All Vital AC Buses EAL ALL 4.16 KV Vital Buses are deenergized Loss of any 2 Fission Product Barriers has occurred or is Imminent OPERATIONAL CONDITION - 1, 2, 3 BASIS Loss of ALL 4.16 KV Vital Buses may result in Safety System Losses and Fission Product Barrier degradation.. Prolonged loss of Vital AC pow~r may cause Core uncovery and the inability to remove heat from the containment. Reactor injection capability may no longer be available, and degradation in core cooling will commence. Indication of continuing core cooling degradation must be based on Fission Product Barrier monitoring with emphasis on EC Judgment as it relates to imminent loss of Fission Product Barrier and because abilities to monitor the barriers is degraded. Imminent is defined as expected to occur within 2 hours. Barrier Analysis Although not directly related to Fission Product Barriers, these events will eventually n~sult in the loss of all three barriers if power cannot be restored. In addition, the extent of the loss of power will result in degraded monitoring capability. It is therefore important in such events to closely monitor the Fission Product Barriers and use judgment related to the IMMINENT Loss or Potential Loss of barriers as directed in EAL Section 3.0 ESCALATION CRITERIA NIA EAL - 7.1.4.b Rev. 00 Page 1 of 3

HCGS EALIRALTechnical Basis DISCUSSION 10 CFR 50.2 defines a station blackout (SBO) as complete loss of AC power to Vital AND Non-Vital buses. Loss of all AC power to the Vital Buses compromises all plant safety systems requiring AC electric power including RHR, ECCS, Spent Fuel Pool Cooling and Service Water. Depending on the status of power supplies to non-vital buses, some Balance of Plant systems that would assist in maintaining plant conditions (i.e. RWCU, condensate, etc.) may be unavailable. Thus, the ability to remove decay heat and control containment parameters is severely rJallenged. During a Loss of all AC power to the Vital Buses, all Class IE System Instruments remain powered from Class IE Uninterruptable Power Supplies (UPS), which are powered by DC power via inverters. The 125 VDC Battery Buses will continue to supply DC power from the batteries. Battery power is limited depending on the discharge rate and predischarge condition of the battery. The ability to restore power to AC buses may eventu?lly be threatened as battery power (DC) is depleted due to the lack of DC (control power) for AC power circuit breakers. Under a Loss of Vital AC Power condition, operation and control of plant systems is guided by the Station Blackout//Loss of Offsite Power//Diesel Generator Malfunction Abnormal Operating Procedure. Successful coping maintains the following key parameters within given acceptable limits:

1. Reactor water level > (T AF)

2. Suppression pool level low enough to prevent HPCI and/or RCIC steam exhaust line flooding

3. Reactor pressure high enough to maintain HPCI and RCIC operable

4. Containment pressure< design limit

5. Torus temperature< design limits (HPCI/RCIC lube oil temperature concern when suction aligned to suppression pool)

6. Drywell temperature below design limits RCIC and HPCI operability is dependent on the availability of 125/250 VDC power. The parameters listed above can be maintained as long as battery power remains available. Battery power is limited depending on the discharge rate and predischarge condition of the battery.

Additionally, the loss of ventilation to the HPCI and RCIC turbine areas may result in a system isolation due to elevated temperatures. Other than HPCI and/or RCIC, additional inventory makeup may be possible by using the diesel driven fire pump to inject water (at low pressure), to the RPV via, the RHR/LPCI system. This may require RPV depressurization using the SRVs, which also require 125 VDC power. The likelihood of loss of the second Barrier should be based on a realistic appraisal of the situation since a delay in an upgrade decision based on only a chance of mitigating the event could result in a loss of valuable time in preparing and implementing public protective actions. In addition, under these conditions, fission product barrier monitoring capability may be degraded. EAL - 7.1.4.b Rev. 00 Page 2 of 3

HCGS EAL/RALTechnical Basis Although it may be difficult to predict when power can be restored, and the loss may be mitigated, it is necessary to give the Emergency Coordinator a reasonable idea of how quickly he may need to declare a General Emergency based on these conditions. It is estimated that several hours are required to fully evacuate the 10 mile EPZ. Taking into consideration the above factors, declaring a General Emergency leaves sufficient time for the offsite authorities to implement Protective Actions well before a radioactive release would occur while providing sufficient tin:I! for on-site and off-site mitigation activities to restore AC power. DEVIATION None REFERENCES NUMARC NESP-007, SG l HC.OP-EO.ZZ-0100 (Q)-FC, Reactor Scram HC.OP-EO.ZZ-0102 (Q)-FC, Primary Containment Control HC.OP-EO.ZZ-0104 (Q)-FC, Radioactive Release Control HC.OP-AB.ZZ-0135 (Q), Station Blackout I Loss of Offsite Power I Diesel Generator Malfunction HCGS Technical Specifications Section 3/4.8, Electrical Power Systems HCGS Individual Plant Evaluation, Section 3. l.1.4.6, 3. l.2.1.6 EAL - 7.1.4.b Rev. 00

'age 3 of 3

HCGS EAL/RALTechnical Basis 7.0 Electrical Power 7.2 Loss of DC Power Capabilities UNUSUAL EVENT- 7.2.1 IC Unplanned Loss of All Vital 125 VDC Power during either Cold Shutdown or Refueling Mode for greater than 15 minutes EAL Unplanned degraded voltage condition for ALL Vital 125 VDC Buses, such that voltage is < 108 VDC

   > 15 minutes have elapsed OPERATIONAL CONDITION - 4, 5, Defueled BASIS An Unplanned degraded voltage condition(< 108 VDC) for ALL Vital 125VDC Buses for greater than 15 minutes with the unit in Operational Condition 4 or 5 compromises the ability to monitor and control plant functions. The minimum required voltage value is based on the minimum voltage required for Vital 125VDC bus operability following a battery discharge test per Technical Specification 4.8.2.1.b. Although continued equipment operation may occur with degraded voltage, this value signifies the minimum operable voltage allowed.

Unplanned is defined as the loss not being the result of planned or scheduled maintenance activities. 15 minutes was chosen to exclude transient or momentary power losses. Barrier Analysis NIA ESCALATION CRITERIA Emergency Classification will escalate if the condition effects the inability to maintain cold shutdown, based on Loss of Decay Heat Removal Capability EAL section 8.1. EAL- 7.2.1 Rev. 00 Page 1 cf 2

HCGS EALIRALTechnical Basis DISCUSSION Vital 125VDC provides control power to Engineered Safety Features actuation, diesel generator auxiliaries, plant alarm and indication circuits as well as the control power for the associated loads. If 125VDC power is lost for an extended period of time (greater than 15 minutes) critical plant functions such as 4.16 KV Breaker Controls, HPCI, RCIC, CS, and RHR. Pump controls required to maintain safe plant conditions may not operate, and core uncovtry with subsequent Reactor Coolant System (RCS) and Primary Containment failure might occur. Both the RCS and Primary Containment may already be Open for Refueling. In Operational Condition 4 or 5, a minimum of two of the four DC power channels are required by Technical Specifications, including either Channel A ( 1OD410) or Channel B (1 OD420). The loss of one of the required two 125VDC distribution systems would require that core alterations be suspended, that handling of irradiated fuel in the Secondary Containment and operations with a potential for draining the reactor vessel be stopped. The design limits of the 1E battery banks are as follows: 125 VDC Vital Power: CHANNEL Switchgear Battery CAPACITY A 10D410 1AD411 1800 AH at 8 hours B IOD420 IBD41 l 1800 AH at 8 hours c 10D430 1CD411 1800 AH at 8 hours D 10D440 1DD411 1800 AH at 8 hours DEVIATION None REFERENCES NUMARC NESP-007, SU7 HC.OP-EO.ZZ-0100 (Q)-FC, Reactor Scram HC.OP-AB.ZZ-0147 (Q), DC System Grounds HC.OP-AB.ZZ-0150 (Q), 125VDC System Malfunction HC.OP-AB.ZZ-0151 (Q), +or - 24 Volt DC Malfunction HC.OP-AB.ZZ-0135 (Q), Station Blackout//Loss ofOtTsite Power//Diesel Generator Malfunction HCGS Technical Specifications Section 3.8.2.2; 3.8.3.2 LCR 93-12, HCGS Technical Specifications Section 4.8.2.1 Revision Request EAL- 7.2.1 Rev. 00 Page 2 of 2

HCGS EAL/RALTechnical Basis 7.0 Electrical Power 7.2 Loss of DC Power Capabilities SITE AREA EMERGENCY- 7.2.3 IC Unplanned Loss of All Vital 125 VDC Power during either Power Operations, Startup or Hot Shutdown for greater than 15 minutes EAL Unplanned degraded voltage condition for ALL Vital 125 VDC Buses, such that voltage is < 108 VDC

    > 15 minutes have elapsed OPERATIONAL CONDITION - 1, 2, 3 BASIS An Unplanned degraded voltage condition {<108 VDC) for ALL Vital 125 VDC Buses for greater than 15 minutes with the unit in Operational Condition 1,2 or 3 compromises the ability to monitor and control plant functions. The minimum required voltage value is based on the minimum yoltage required for Vital 125 VDC bus operability following a battery discharge test per Technical Specification 4.8.2.1.b. Although continued equipment operation may occur with degraded voltage, this value signifies the minimum operable voltage allowed.

Unplanned is defined as the loss not being the result of planned or scheduled maintenance activities. 15 minutes was chosen to exclude transient or momentary power losses. Barrier Analysis NIA ESCALATION CRITERIA Emergency Classification will escalate based on other EALs indicating Radiological Release (EAL Section 6.0) or loss of Fission Product Barriers (EAL Section 3.0). EAL- 7.2.3 Rev. 00 Page 1 of 3

HCGS EAL/RALTechnical Basis DISCUSSION Vital 125VDC provides control power to Engineered Safety Features actuation, diesel generator auxiliaries, plant alann and indication circuits as well as the control power for the associated loads. If 125VDC power is lost for an e?(tended period of time (greater than 15 minutes) critical plant functions such as 4.16 KV Breaker Controls, HPCI, RCIC, CS, and RIIR. pump controls required to maintain safe plant conditions may not operate, and core uncovery with subsequent Reactor Coolant System (RCS) and Primary Containment failure might occur. Loss of ADS may create a loss of low pressure ECCS availability due to the potential inability to depressurize the reactor. In addition, loss of these buses will eventually lead to MSIV closure and reactor scram due to the loss of the Primary Containment Instrument Gas (PCIG). Subsequent to MSIV closure, much of the equipment noted above will be required for plant stabilization and shutdown. A sustained loss of 125VDC power will threaten the ability to remove heat from the reactor core and from the containment. SRVs will remain operable in the relief mode and the heat addition to the containment could result in a loss of the Primary Containment as a fission product release barrier. HPCI and RCIC also require Vital 250VDC power for system operability. Loss of Vital 250 VDC power will only render the associated system inoperable; it does not affect the operability of the systems listed/discussed above. Loss of all Vital 1E 125VDC power will also render these systems inoperable for automatic initiation, and from the Control Room due to loss of control power. The loss Vital lE 250VDC system requires that HPCI and/or RCIC be declared inoperable and the respective Technical Specification LCO be entered. Loss of these sources is therefore not included in this EAL. The design limits of the 1E battery banks are as follows: 125 VDC Vital Power: Channel Switchgear Battery CAPACITY A 100410 1AD41 l 1800 AH at 8 hours B 100420 IB0411 1800 AH at 8 hours c 100430 IC0411 1800 AH at 8 hours D 100440 IOD41 l 1800 AH at 8 hours In Operational Conditions 1, 2, or 3, the loss of any single channel 125VDC power source would require the channel to be restored within 2 hours or the unit placed in at least Hot Shutdown within the next 12 hours and in Cold Shutdown within the following 24 hours. EAL- 7.2.3 Rev. 00 Page 2 of 3

HCGS EAL!RALTechnical Basis DEVIATION None REFERENCES NUMARC NESP-007, SS3 HC.OP-EO.ZZ-0100 (Q)-FC, Reactor Scram HC.OP-EO.ZZ-0202 (Q)-FC, Emergency Depressurization HC.OP-AB.ZZ-0147 (Q), DC System Grounds HC.OP-AB.ZZ-0149 (Q), 250VDC System Malfunction HC.OP-AB.ZZ-0150 (Q), 125VDC System Malfunction HC.OP-AB.ZZ-0151 (Q), +or- 24 Volt DC Malfunction HC.OP-AB.ZZ-0135 (Q), Station Blackout//Loss ofOffsite Power//Diesel Generator Malfunction HCGS Technical Specifications Section 3/4.8.2.1, 3/4.8.3.1 LCR 93-12, HCGS Technical Specifications Section 4.8.2.1 Revision Request EAL-7.2.3 Rev. 00 Page 3 of 3

HCGS ~AL/RAL Technical Basis 8.0 System Malfunctions .-"1

                                                                                               ~~~                \~
                                                                         .........-:-*-:--"'~*~*~2*.**.* ~ ~1 8.1 Loss of Heat Removal CapabilitY                                                      \
                                                                                                                      \
                                                                                        /..-                           \

ALERT - 8.1.2 \ (' ~

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                                                          \ .            "-..Q__,)

IC Inability to Maintain the Plant in Cold Shutdown\ ~

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EAL Unplanned, Complete Loss of ALL Technical Specification required systems available to provide Decay Heat Removal functions EITHER one of the following occur:

RCS Temperature has risen to> 200 °F (Excluding a< 15 minute rise> 200 °F with a heat removal function restored)
An UNCONTROLLED temperature rise is RAPIDLY approaching 200 °F (with NO heat removal function restored)

OPERA TI ON AL CONDITION - 4, 5 BASIS Loss of Decay Heat Removal capabilities necessary to maintain Cold Shutdown conditions could potentially lead to core damage if corrective actions are not implemented. Declaration of an Alert is warranted when ALL Technical Specification required systems are not available to provide Decay Heat Removal functions and cannot be restored to prevent boiling in the core. The specification of an RCS temperature rise, rather than specific equipment failures, recognizes the potential for long heatup times providing adequate time for restoration of some form of alternate cooling. The statement "Unplanned, Complete Loss of ALL Technical Specification required systems available to provide Decay Heat Removal functions" is intended to represent a complete loss of functions available, or an inadequate ability, to provide core cooling during the Cold Shutdown and Refueling Modes, including alternate decay heat removal methods. This EAL allows for actions taken IAW OP-AB.ZZ-0142, Loss of Shutdown Cooling (Abnormal Operating EAL- 8.1.2 Rev. 00 Page 1 of 3

HCGS EALIRALTechnical Basis Procedure) to reestablish RHR in the Shutdown Cooling Mode or provide for an alternate methods of decay heat removal, with the intent of maintaining RCS temperature below 200°F. For loss of an in-service Decay Heat Removal system with other decay heat removal methods available, actions taken to provide for restoration of a decay heat removal function may require time to implement. If the event results in RCS temperature "momentarily" (for less than 15 minutes) rising above 200°F with heat removal capability restored, Emergency Coordinator judgment will be required to determine whether heat removal systems are adequate to prevent boiling in the core and restoration of RCS temperature control. Momentary (not to exceed 15 minutes) unplanned excursions above 200°F, when alternate decay heat removal capabilities exist, should not be classified under this EAL. NRC analysis has shown that specific sequences can result in core uncovery within 15 to 20 minutes and severe core damage within an hour after decay heat removal capability has been lost. Unplanned is defined as a condition that is not due to scheduled operations or maintenance activities, in which an RHR system is intentionally removed from service. Barrier Analysis NIA ESCALATION CRITERIA Emergency Classification will escalate to a Site Area Emergency based on inability to maintain RPV Water level above the Top of the Active Fuel, or r_ising Radiological Releases. DISCUSSION The Residual Heat Removal (RHR) system provides the normal method for decay heat removal operating in the Shutdown Cooling Mode. With RHR unavailable for shutdown cooling operation, (including the loss of SACS and/or service water which supply cooling water to the RHR heat exchangers), alternate decay heat removal system can be aligned to control decay heat. An unavailability of these systems, can result in a gradual rise in RCS temperature to the values specified in this EAL. The rate of rise in coolant temperature would be dependent on the amount of decay heat present. The threshold for this EAL is the RCS temperature transition value between Operational Conditions 4 and 3. Procedural guidance is provided to establish an alternate method of decay heat removal. These alternate methods include: aligning Reactor Water Cleanup system (RWCU), with maximum RACS aligned to the Non-Regenerative Heat Exchanger; aligning Condensate Transfer via the ECCS injection lines; aligning RPV Head Spray with RPV Water Level established above

+ 80"; maximizing Fuel Pool Cooling if the RPV head is removed and the reactor cavity flooded; using the "C" RHR pump crosstied to the "A" RHR loop.

EAL - 8.1.2 Rev. 00 Page 2 of 3

HCGS EAL/RALTechnical Basis If these alternate means are unavailable, or ineffective, decay heat removal must be accomplished by feed-and-bleed using ECCS systems and discharging steam to the Suppression Pool via the SRVs. DEVIATION None REFERENCES NUMARC NESP-007, SA3 NUMARC Questions and Answers, June 1993, "System Malfunction Question #6b" HC.OP-AB.ZZ-0142 (Q), Loss of Shutdown Cooling HC.OP-EO.ZZ-0101 (Q)-FC, Reactor Pressure Vessel (RPV) Control HC.OP-EO.ZZ-0102 (Q)-FC, Primary Containment Control Hope Creek Appendix A based on NED0-2121, Supplement A to BWR Owners Group Emergency Procedure Guidelines, Revision 4 HCGS Technical Specifications Sections 3/4.3, 3/4.4.9, 3/4.7.1, 3/4.7.2 EAL - 8.1.2 Rev.00 Page 3 of 3

HCGS EAL/RALTechnical Basis 8.0 System Malfunctions 8.1 Loss of Heat Removal Capability SITE AREA EMERGENCY-8.1.3.a IC Loss of Reactor Water Level that has or will Uncover Fuel in the Reactor Vessel EAL Reactor Water Level REACHES -161" (Top of Active Fuel) OPERATIONAL CONDITION - 4, 5 BASIS Reactor Water Level reaching -161" (Top of Active Fuel) indicates a loss of core submergence. Without core submergence, the integrity of the fuel clad barrier can no longer be assured, even with the reduced decay heat levels in Cold Shutdown and Refuel. This event is classified based on reaching the Reactor Water level threshold (instead of being able to restore and maintain above the threshold) due to the potentially severe consequences of a loss of core submergence. Since the design of the normal and emergency makeup systems should preclude this condition, an extreme challenge to their ability to provide core cooling by submergence has occurred. Additionally, ECCS availability and Containment Integrity requirements may be relaxed under these Operational Conditions, thus classification at the Site Area Emergency level is warranted. Barrier Analysis Fuel Clad Barrier has been potentially lost RCS Barrier has been lost. ESCALATION CRITERIA Emergency Classification will escalate to a General Emergency based on abnormal Radiological Releases. EAL - 8.1.3.a Rev.00 Page 1 of 2

HCGS EALIRALTechnical Basis DISCUSSION Core Submergence ensures adequate core cooling. When RPV water level decreases to below Top of Active Fuel (T AF) the ability to effectively remove decay heat can no longer be guaranteed and the Fuel Cladding Barrier can no longer be considered intact. Sustained partial or total core uncovery can result in clad damage and a significant release of fission products to the reactor coolant. Sustained core uncovery can also result in a breach of the reactor vessel, or an unisolated intersystem LOCA wi~h the RHR System. DEVIATION None REFEPJ:NCES NUMARC NESP-007, SSS HC.OP-EO.ZZ-0101 (Q)-FC, Reactor Pressure Vessel (RPV) Control HC.OP-EO.ZZ-0201 (Q)-FC, Alternate Level Control EAL - 8.1.3.a Rev.00 Page 2 of 2

HCGS EAL/RALTechnical Basis 8.0 System Malfunctions 8.1 Loss of Heat Removal Capability SITE AREA EMERGENCY-8.1.3.b IC Complete Loss of Functions Needed to Achieve Cold Shutdown Conditions EAL Loss of Main Condenser capabilities, as evidenced ~y an inability to remove Decay Heat from the Reactor Loss of Torus capabilities as evidenced by EITHER one of the following:

Entry into an Unsafe region of ANY of the following curves:
Heat Capacity Temperature Limit (HCTL) Curve
Heat Capacity Level Limit (HCLL) Curve
Pressure Suppression Pressure (PSP) Curve
SRV Tailpipe Level Limit Curve
Insufficient SR V capacity to reduce RPV pressure OPERATIONAL CONDITION - 1, 2, 3 BASIS A Complete Loss of decay heat removal systems required to ACHIEVE Cold Shutdown conditions from a Hot Shutdown condition, represents a significant challenge to the plant due to the failure of multiple systems designed for the protection of the public. Hence, declaration of a Site Area Emergency is warranted.

This EAL specifically includes a degradation of those plant systems required to ACHIEVE a Cold Shutdown condition. It does NOT include an inability to MAINTAIN a Cold Shutdown condition. The inability to MAINTAIN Cold Shutdown Conditions is specifically addressed by EAL 8.1.2. Hence, a Loss of RHR Shutdown Cooling is not included in this EAL. This EAL includes a loss ofService Water or SACS capabilities, based on the effect a loss of these systems has on the ability to maintain Torus capabilities with the Safe Region of the referenced EOP curves. Loss is defined as the systems being unavailable to perform their intended EAL - 8.1.3.b Rev. 00 Page 1 of 3

HCGS EALIRALTechnical Basis design function. Hence, in the case where the Main Condenser became isolated from the Reactor due to an MSIV Isolation, but the MSIV could be reopened by procedure, then a Loss of the Main Condenser capabilities has not occurred. Barrier Analysis NIA ESCALATION CRITERIA Emergency Classification will escalate to a General Emergency based on loss of Fission Product Barriers or Radiological Releases. DISCUSSION In this event, a loss of both the normal heat sink for the Reactor and an impending severe degradation of alternate heat removal capability to the Torus has occurred. Loss of the heat sink for the reactor when in a Hot Shutdown condition will limit the ability to maintain that Operational Condition, or to cooldown the reactor if required. The Main Condenser can be lost for a variety or reasons; loss of Circulating Water, loss of the Turbine Control and/or Bypass Valve functions, Main Steam Line isolation, etc. With the Main Condenser not available and without the RHR System lined up in Shutdown Cooling Mode, decay heat must be removed from the RCS by HPCI, RCIC or the SRVs and be absorbed in the Suppression Pool (SP). Loss of the pressure control ab_ility of the SRVs as indicated by the inability to reduce RPV pressure represents a loss of control of a major RCS parameter which could result in RPV overpressure conditions, or the inability to cooldown if Cold Shutdown is required. The HCTL curve is defined as the highest Torus temperature at which initiation ofRPV depressurization will not result in exceeding either the SP design temperature or the Primary Containment pressure limit before the rate of energy transfer from the RPV to the Primary Containment is beyond the capacity of the Containment Vent. The HCLL curve is defined as the higher of either the elevation of the Containment downcomer opening or the lowest Torus level at which initiation of RPV depressurization will not result in exceeding the HCTL. Violation of either curve would require an immediate emergency depressurization, thus ensuring that the immediately present thermal energy in the RCS has been transferred to the Primary Containment while maintaining the Containment within design limits. This represents a serious potential threat to the Primary Containment Barrier. EAL - 8.1.3.b Rev.00 Page 2 of *3

HCGS EALIRALTechnical Basis DEVIATION The NUMARC IC associated with EAL SS4 suggests that the IC should include a Complete Loss of Functions needed to achieve or maintain Hot Shutdown. The NUMARC basis includes both reactivity control and decay heat removal. At Hope Creek, as with all other BWRs, the operator action of placing the Reactor Mode Switch in the Shutdown position that results in Control Rod inserting into the cor .:! such that the Reactor will remain shutdown under all conditions without boron, places the Reactor in a Hot Shutdown condition. No additional actions are required to maintain the Reactor in this condition. Systems are required and additional operator actions are required to achieve Cold Shutdown conditions. Based on this, Hope Creek has modified the NUMARC IC for SS4 to apply specifically to a total loss of decay heat removal, since reactivity control concerns are addressed under the ATWS Section. This IC and EAL are consistent with the requirements for declaration of a Site Area Emergency. REFERENCES NUMARC NESP-007, SS4 HC.OP-EO.ZZ-0100 (Q)-FC, Reactor Scram HC.OP-EO.ZZ-0101 (Q)-FC, Reactor Pressure Vessel (RPV) Control HC.OP-EO.ZZ-0102 (Q)-FC, Primary Containment Control Hope Creek Appendix A based on NED0-2121, Surplement A to BWR Owners Group Emergency Procedure Guidelines, Revision 4 HCGS Technical Specifications 3/4.1.3, 3/4.1.5 EAL - 8.1.3.b Rev. 00 Page 3 of 3

HCGS EALIRALTechnical Basis 8.0 System Malfunctions 8.2 Loss of Overhead Annunciators UNUSUAL EVENT-8.2.1 IC Unplanned Loss of Most or All AnnunciatioQ or Indication in the Control Room for Greater Than 15 Minutes EAL Unplanned Loss of> 75% of Main Control Room Overhead Annunciators EITHER one of the following:

15 minutes have elapsed since the loss of OHAs
A significant transient is in progres~

OPERATIONAL CONDITION - 1, 2, 3 BASIS A Unplanned Loss of> 75% of all Main Control Room OHAs without a plant transient in Operational Conditions 1, 2 or 3 for greater than 15 minutes warrants a heightened awareness by Control Room Operators. Quantification of> 75% is left to the discretion of the SNSS and is considered approximately 75%. It is not intended that a detailed count be performed, but that a rough approximation be used to determine the severity of the loss. Either CRIDS or SPDS is available to provide compensatory (backup) indication. 15 minutes is used as a threshold to exclude transient or momentary power losses. The 15 minutes clock starts when the annunciators have been lost, or are determined to have been lost. If upon time of discovery it is determined that the annunciators have been lost for at least 15 minutes prior to discovery, classification should be made under this EAL regardless of time required for restoration. If it is determined that the annunciators had previously been lost for at least 15 minutes but the annunciators were available at the time of discovery, classification is not requ!red under this EAL but a review of the "After The Fact" RAL must be completed. Unplanned loss of annunciators excludes scheduled maintenance and testing activities. EAL- 8.2.1 Rev. 00 Page 1 of 2

HCGS EALIRALTechnical Basis Significant transients include response to automatic or manually initiated actions such as:

Reactor Scram
Load Rejection > 25% Power
ECCS Injection
Thermal Power oscillations of I 0%

Barrier Analysis NIA ESCALATION CRITERIA Emergency classification will be escalate to an Alert if a transient is in progress or if CRIDS and SPDS becomes unavailable and 15 minutes have elapsed since the loss of OHAs. DISCUSSION Without Control Room OHAs, there may be difficulty initially recognizing changing plant conditions as well as monitoring conditions associated with normal plant operations. SNSS judgment of the severity of the loss should also be based on the need to initiate increased or continuous plant equipment monitoring. Also, specific annunciator loss should be judged against those needed for by the operating staff for operation in abnormal and emergency operating procedures. Most alarm conditions for the annunciator system have CRIDS digital alarm points as well. By monitoring the CRIDS screens, most alarm conditions can be observed and responded to independent of the overhead annunciators. This EAL is not required in Operational Conditions 4 or 5 due to the limited number of safety systems required for operation. DEVIATION An EAL threshold for declaring an UE has been added if a significant transient is in progress when the loss of annunciators occurs as requested by the NJ-BNE. These two independent events occurring at the same time warrant an expeditious notification and not waiting 15 minutes for the Unusual Event declaration. REFERENCES NUMARC NESP-007, SU3 HC.OP.AB.ZZ-0143 (Q}, Loss of Overhead Annunciators I Loss ofCRIDS HC.OP.EO.ZZ-0100 (Q)-FC, Reactor Scram EAL- 8.2.1 Rev. 00 Page 2 of 2

HCGS EAL/RALTechnical Basis 8.0 System Malfunctions 8.2 Loss of Overhead Annunciators ALERT - 8.2.2.a IC Unplanned Loss of Most or All Control Room Annunci::ttors and a Significant Transient is in Progress or Compensatory Indicators are Unavailable EAL Unplanned Loss of> 75% of Main Control Room Overhead Annunciators A significant transient is in progress 15 minutes have elapsed since the loss of OHAs OPERATIONAL CONDITION - 1, 2, 3 BASIS An Unplanned Loss of> 75% of Main Control Room OHAs with a significant transient in progress significantly hampers operator response. Quantification of> 75% is left to the discretion of the SNSS and is considered approximately 75%. It is not intended that a detailed count be performed, but that a rough approximation be used to determine the severity of the loss. Significant transients include response to automatic or manually initiated actions such as:

Rea.;tor Scram
Load Rejection > 25% Power
ECCS Injection
Thermal Power oscillations of I 0%

15 minutes is used as a threshold to exclude transient or momentary power losses. The 15 minutes clock starts when the annunciators have been lost, or are determined to have been lost. If upon time of discovery it is determined that the annunciators have been lost for at least 15 minutes prior to discovery, classification must be made under this EAL regardless of time required for restoration. If it is determined that the annunciators were lost for at least I 5 minutes with the annunciators available at the time of discovery, classification is not required under this EAL but a EAL - 8.2.2.a Rev. 00 Page 1 of 2

HCGS EALIRALTechnical Basis review of the "After The Fact" RAL should be completed. Unplanned loss of annunciators excludes scheduled maintenance and testing activities. Barrier Analysis NIA E~CALATION CRITERIA Emergency Classification will escalate to a Site Area Emergency based on alternate indications are not available with a loss of Control Room OHAs and a significant plant transient in progress and a failure of both the Control Room Integrated Display System (CRIDS) and the Safety Parameter Display System (SPDS). DISCUSSION Without Control Room OHAs, it may be difficult to monitor col)ditions associated with normal plant operations. During a transient event such as those listed in the EAL, the difficulty becomes more acute. Loss of OHAs significantly reduces the ability of the operations staff to monitor and evaluate plant conditions. SNSS judgment of the severity of the loss should also be based on the need to initiate increased or continuous plant equipment monitoring. Most alarm conditions for the annunciator system have CRIDS digital alarm points as well. By monitoring the CRIDS screens, most alarm conditions can be observed and responded to, independent of the overhead annunciators. The SPDS also provides information and indication related to selected plant parameters during a plant transient. This EAL is not required in Operational Conditions 4 or 5 due to the limited number of safety systems required for operation. DEVIATION None REFERENCES NUMARC NESP-007, SA4 HC.OP.AB.ZZ-0143 (Q), Loss of Overhead Annunciators I Loss of CRIDS HC.OP.EO.ZZ-0100 (Q)-FC, Reactor Scram* EAL - 8.2.2.a Rev. 00 Page 2 of 2

HCGS EALIRALTechnical Basis 8.0 System Malfunctions 8.2 Loss of Overhead Annunciators ALERT - 8.2.2.b IC Unplanned Loss of Most or All Control Room Annunciators and a Significant Transient is in Progress or Compensatory Indicators are Unavailable EAL Unplanned Loss of> 75% of Main Control Room Overhead Annunciators BOTH of the following:

CRIDS
SPDS are NOT AVAILABLE 15 minutes have elapsed since the loss of OHAs OPERATIONAL CONDITION -1, 2, 3 BASIS An Unplanned Loss of> 75% of Main Control Room OHAs with loss of backup control room monitoring significantly hampers operator response. Quantification of> 75% is left to the discretion of the SNSS, and is considered approximately 75%. It is not intended that a detailed count be performed, but that a rough approximation be used to determine the severity of the loss.

15 minutes is used as a threshold to exclude transient or momentary power losses. The 15 minutes clock starts when the annunciators have been lost, or are determined to have been lost. If upon time of discovery it is determined that the annunciators have been lost for at least 15 minutes prior to discovery, classification must be made under this EAL regardless of time required for restoration. If it is determined that the annunciators were lost for at least 15 minutes with the annunciators available at the time of discovery, classification is not required under this EAL, but a review of the "After The Fact" RAL should be completed. Unplanned loss of annunciators excludes scheduled maintenance and testing activities. The fifteen minutes also allows for attempting to restore either the CRIDS or SPDS computer. EAL - 8.2.2.b Rev. 00 Pagel of 2

HCGS EALIRALTechnical Basis Barrier Analysis NIA ESCALATION CRITERIA Emergency Classification will escalate to a Site Area Emergency based on alternate indications are not available with a loss of Control Room OHAs and a significant plant transient in progress and a failure of both the Control Room Integrated Display System (CRIDS) and the Safety Parameter Display System (SPDS). DISCUSSION CRIDS is not essential for the safe shutdown or operation of the plant. However, with the loss of Control Room OHAs, the loss of CRIDS significantly reduces the ability of the operations staff to monitor and evaluate plant conditions. SNSS judgment of the severity of the loss should also be based on the need to initiate increased or continuous plant equipment monitoring. Most alarm conditions for the annunciator system have CRIDS digital alarm points as well. By monitoring the CRIDS screens, most alarm conditions can be observed and responded to, independent of the overhead annunciators. SPDS also provides information and indication related to selected plant parameters during a plant transient. Loss of this assessment tool may hamper operators attempt to comply with directions provided in EOPs or may limit the recognition of significant parameter values called out in the EOPs. It is not included in the threshold for this EAL because of the limited scope of the parameters it monitors. This EAL is not required in Operational Conditions 4 or 5 due to the limited number of safety systems required for operation. DEVIATION None REFERENCES NUMARC NESP-007, SA4 HC.OP.AB.ZZ-0143 (Q), Loss of Overhead Annunciators I Loss of CRIDS HC.OP.EO.ZZ-0100 (Q)-FC, Reactor Scram EAL - 8.2.2.b Rev. 00 Page 2 of 2

HCGS EALIRALTechnical Basis 8.0 System Malfunctions 8.2 Loss of Overhead Annunciators SITE AREA EMERGENCY - 8.2.3 IC Inability to Monitor a Significant Transient in Progress EAL Loss of> 75% of Main Control Room Overhead Annunciators A significant transient is in progress BOTH of the following:

CRIDS
SPDS are NOT AVAILABLE Main Control Room Indications are NOT AVAILABLE to monitor ANY of the following:
RCS Status
Reactivity Control
ECCS
Containment Parameters OPERATIONAL CONDITION - 1, 2, 3 BASIS A Loss of> 75% of Main Control Room OHAs with loss of backup control room monitoring, and while a significant transient is in progress represents a major loss of ability to properly respond to a transient condition. Quantification of > 75% is left to the discretion of the SNSS and is considered approximately 75%. It is not intended that a detailed count be performed, but that a rough approximation be used to determine the severity of the loss.

Loss of backup monitoring from both CRIDS and SPDS compounds the difficulty to monitor the progress of the transient. In addition, a loss of alternate Main Control Room Indications for any EAL- 8.2.3 Rev.00 Page 1 of 2

HCGS EAL/RALTechnical Basis one of the safety systems listed in the EAL must also occur. Significant transients include response to automatic or manually initiated actions such as:

Reactor Scram
Load Rejection > 25% Power
ECCS Injection
Thermal Powtr oscillations of I 0%

Barrier Analysis NIA ESCALATION CRITERIA Emergency Classification will escalate to a General Emergency based on either the Loss of Fission Product Barriers, increased plant radiation levels or radiological releases. DISCUSSION Without Control Room annunciators, it may be difficult to monitor conditions associated with normal plant operations. During a transient event such as those listed in the EAL, the difficulty becomes more acute. Compounding these, a concurrent loss of control room backup and alternate monitoring will further hinder operations staff decision making needed to respond to the transient. The Safety Parameter Display System (SPDS) also provides information and indication related to selected plant parameters during a plant transient. Loss of this assessment tool may hamper operators attempt to comply with directions provided in EOPs or may limit the recognition of significant parameter values called out in the EOPs. It is not included in the threshold for this EAL because of the limited scope of the parameters it monitors. This EAL is not required in Operational Conditions 4 or 5 due to the limited number of safety systems required for operation. DEVIATION None REFERENCES NUMARC NESP-007, SS6 HC.OP.AB.ZZ-0143 (Q), Loss of Overhead Annunciators I Loss of CRIDS HC.OP.EO.ZZ-0100 (Q)-FC, Reactor Scram EAL- 8.2.3 Rev. 00 Page 2 of 2

HCGS EALIRALTechnical Basis 8.0 System Malfunctions 8.3 Loss of Communications Capability UNUSUAL EVENT - 8.3.1.a IC Unplanned Loss of All Onsite or Offsite Communications Capabilities EAL Unplanned Loss of ALL ONSITE communications as evidenced by the loss of ALL of the following systems:

Direct Inward Dial System (DID)
Station Page System (Gaitronics)
Station Radio System OPERATIONAL CONDITION - All BASIS An Unplanned loss of communication ability significantly degrades the operating crew's ability to perform tasks necessary for plant operations and warrants declaration of an Unusual Event.

The loss of ALL ONSITE communications capability is more comprehensive than that addressed by 10CFRSO. 72.b. Unplanned is defined as the loss of communication capabilities not being the result of planned maintenance activities, where compensatory measures would be taken. Barrier Analysis NIA ESCALATION CRITERIA None DISCUSSION None EAL - 8.3.1.a Rev. 00 Page I of2

HCGS EAL/RALTechnical Basis DEVIATION None REFERENCES NUMARC NESP-007, SU6 EAL - 8.3.1.a Rev. 00 Page2of2

HCGS EALIRALTechnical Basis 8.0 System Malfunctions 8.3 Loss of Communications Capability UNUSUAL EVENT- 8.3.1.b . IC Unplanned Loss of All Onsite or Offsite Communications Capabilities EAL Unplanned Loss of ALL OFFSITE communications as evidenced by the loss of ALL of the following systems:

Direct Inward Dial System (DID)
Nuclear Emergency Telephone System (NETS)
ESSX (Centrex) Phone System OPERATIONAL CONDITION - All BASIS An Unplanned loss of communication ability significantly degrades the operating crew's ability to communicate with offsite authorities and warrants declaration of an Unusual Event. The loss of ALL OFFSITE communications capability is more comprehensive than that addressed by IOCFRS0.72.b. Unplanned is defined as the loss of communication capabilities not being the result of planned maintenance activities, where compensatory measures would be taken.

Barrier Analysis NIA ESCALA Tl ON CRITERIA None DISCUSSION None EAL - 8.3.1.b Rev.00 Page I of2

HCGS EAL/RALTechnical Basis DEVIATION None REFERENCES NUMARC NESP-007, SU6 EAL - 8.3. l.b Rev. 00 Page 2of2

HCGS EAL/RALTechnical Basis 8.0 System 1\-talfunctions 8.4 Control Room Evacuation ALERT - 8.4.2 IC Main Control Room Evacuation has been Initiated EAL Main Control Room Evacuation has been initiated OPERATIONAL CONDITION - All BASIS Main Control Room evacuation represents a serious plant situation since the degree of plant control at the Remote Shutdown Panel (RSP) is not as complete as from the Main Control Room. The intent of this EAL is to declare an Alert when the determination to evacuate the Main Control Room has been made based on environmental/personnel safety concerns, and the physical process of evacuating the Control Room per OP-AB.ZZ-0130, Control Room Evacuation has commenced. Barrier Analysis NIA ESCALATION CRITERIA Emergency Classification will escalate to a Site Area Emergency if control of the plant cannot be established within 15 minutes. DISCUSSION Main Control Room evacuation requires establishment of plant control from outside the Control Room at the RSP. Support from the Technical Support Center (TSC) is necessary. Establishing remote system control will bypass many protective trips and interlocks. In addition, most of the instrumentation and assessment tools available in the Main Control Room will not be available. Operator actions upon deciding that the Control Room should be evacuated include scramming the Reactor and closing the MSIVs. With these actions taken, all RCS inventory and pressure control can be accomplished at the RSP. EAL- 8.4.2 Rev. 00 Page 1 of 2

HCGS EALIRALTechnical Basis A fire in any one of the following fire zones has the potential to render redundant safe shutdown controls and instrumentation in the Main Control Room inoperable. Fire Zone Description 5202 Cable Spreading Room 5302 Control Equipment Room 5403 Control Equipment Room Mezzanine 5510 Main Control Room 5605 Class IE Panel Room 5620 1E Panel Room HY AC 5704 Diesel Area HY AC DEVIATION None REFERENCES NUMARC NESP-007, HAS HC.OP-AB.ZZ-0130 (Q), Control Room Evacuation EAL- 8.4.2 Rev. 00 Page 2 of 2

HCGS EAL/RALTechnical Basis 8.0 System Malfunctions 8.4 Control Room Evacuation SITE AREA EMERGENCY - 8.4.3 IC Main Control Room Evacuation has been Initiated and Plant Control cannot be established EAL Main Control Room Evacuation has been initiated Control of the plant CANNOT be established from outside the Main Control Room within 15 minutes OPERATIONAL CONDITION - All BASIS Failure to transfer and establish control of safety systems needed to maintain the Reactor in a safe shutdown condition and remove decay heat, could result in damage to the fission product barriers, and the ability to determine plant status may be lost. The 15 minute time limit for transfer of control is based on a reasonable time period for personnel to leave the Main Control Room, arrive at the Remote Shutdown Panel, and reestablish plant control to preclude core uncovery and/or core damage. The term "control of the plant" will require SNSS assessment to determine whether sufficient control has been established to maintain adequate core cooling. Barrier Analysis NIA ESCALATION CRITERIA Emergency Classification will escalate to a General Emergency based on Joss of fission product barriers, abnormal radiological releases, or Emergency Coordinator judgment. EAL- 8.4.3 Rev. 00 Page 1 of 2

HCGS EALIRALTechnical Basis DISCUSSION Most of the monitoring capability of the Remote Shutdown Panel is not enabled until control is transferred. During this transitional period the function of monitoring and/or controlling parameters necessary for plant safety may not be occurring and may result in a threat to plant safety. If the transitional period is prolonged, damage to plant systems and safety barriers may occur and worsen witl":out actions being taken to mitigate the consequences. Main Control Room evacuation requires establishment of plant control from outside the Control Room at the Remote Shutdown Panel (RSP). Support from the Technical Support Center (TSC) and Emergency Operations Facility (EOF) is necessary. A fire in any one of the following fire zones has the potential to render redundant safe shutdown controls and instrumentation in the Main Control Room inoperable. Fire Zone Description 5202 Cable Spreading Room 5302 Control Equipment Room 5403 Control Equipment Room Mezzanine 5510 Main Control Room 5605 Class 1E Panel Room 5620 1E Panel Room HY AC 5704 Diesel Area HY AC DEVIATION None REFERENCES NUMARC NESP-007, HS2 HC.OP-AB.ZZ-0130 (Q), Control Room Evacuation EAL - 8.4.3 Rev. 00 Page 2 of 2

HCGS EALIRALTechnical Basis 8.0 System Malfunctions 8.5 Technical Specifications UNUSUAL EVENT - 8.5.1 IC Inability to Reach Required Operational Condition within Technical Specification Limits EAL Plant is NOT brought to the REQUIRED Operational Condition within the Technical Specification required time limit OPERATIONAL CONDITION - I, 2, 3 BASIS Failure to place the unit in an Operational Condition in compliance with the Technical Specification {TIS) LCO Action Statement warrants declaration of an Unusual Event due to the . plant being outside the defined TIS safety envelope. Classification under this EAL is specific to an INABILITY ORF AIL URE to comply with the Operational Condition change requirements of those TIS LC Os that require the plant placed in a more *conservative Operational Condition. Classification should be made under this EAL for a failure to comply with ANY TIS required change in Operational Condition FROM the Operational Conditions in which this EAL applies (Operational Conditions 1,2 and 3). An Unusual Event is declared when the plant fails to comply with the Action Statement of an LCO, and NOT as the result of implementing the required action. Barrier Analysis NIA ESCALATION CRITERIA Emergency Classification will escalate based upon system malfunctions or other conditions covered in various other EAL sections. EAL - 8.5.1 Rev. 00 Page 1 of 2

HCGS EAL/RALTechnical Basis DISCUSSION A shutdown required by the TIS requires a one hour report under 10 CFR 50.72 (b) Non-emergency events. The plant is within its safety envelope when actions are completed within the allowable Action- Statement time in the TIS. If the times specified within the Action Statements are not met, the plant may be in an unsafe condition. The declaration is based on exceeding the LCO Action Statement time period from the POINT OF RECOGNITION and is not related to how long a plant condition may have existed. DEVIATION None REFERENCES NUMARC NESP-007, SU2 HCGS Technical Specifications 10CFR50.72 NUMARC Questions and Answers, June 1993, "General Question #7 11 NUMARC Questions and Answers, June 1993, "General Question #8 11 EAL- 8.5.1 Rev.00 Page 2 of 2

HCGS EAL/RALTechnical Basis 9.0 Hazards - Internal/External __,.--:-;_:~~~--(*--\.

9.1 Security Threats i.

                                                                                    /_,..,....                        \

UNUSUAL EVENT - 9.1.1 \ /C-, \

                                                                \           \.[::J             ,..,.,.,,,,..,..,..,.,.."'
                                                                  \                     -'~

IC Confirmed Security Event Which Indicates a Potential ~egr~"atiorMrrffie Level of Safety of the Plant ,_.... EAL Confirmed security threat directed toward the station as evidenced by ANY one of the following:

Credible threat of malicious acts or destructive device within the Protected Area, resulting in SCP-5 implementation
Credible intrusion or assault threat to the Protected Area, resulting in SCP-5 implementation
Attempted intrusion or assault to the Protected Area, resulting in SCP-7 or SCP-11 implementation
Malicious acts attempted or discovered within the Protected Area, resulting in SCP-I 0 implementation
Hostage/Extortion situation that threatens normal plant operations, resulting in SCP-8 implementation
Destructive device discovered within the Protected Area, resulting in SCP-10 implementation OPERATIONAL CONDITION - All BASIS Security evt:*nts classified under this EAL represent a potential degradation in the levelI of safety of the plant. The EAL threshold is satisfied if the event is identified as being directed toward the station. The intent of this EAL is to classify security events which threaten the Protected Area, but have not been determined to threaten Plant Vital Areas.

A confirmed security threat exists if physical evidence supporting the threat exists, if information independent from the actual threat exists, or if a specific group claims responsibility for the threat. The SNSS/EC should declare an Unusual Event upon consulting with Security to determine the validity of the entry conditions. EAL - 9.1.1 Rev.00 Page 1 of 2

HCGS EAL/RALTechnical Basis Security Contingency Procedure (SCP) numbers are referenced following each EAL threshold. Since some SCP numbers appear in more than one EAL, the on-duty PSE&G Security Supervisor will provide information concerning the specific event to aid in classification. Barrier Analysis NIA ESCALATION CRITERIA Emergency Classification will escalate to an Alert based upon an actual hostile intrusion or malicious acts within the Protected Area. DISCUSSION Security events which do not represent a potential degradation in the level of safety of the plant are reported under RAL Section 11.0, One Hour Non-Emergency Safeguards Event (10 CFR 73.71 or 10 CFR 50.72), and will not result in an Unusual Event declaration. The following is an index of Security Contingency Procedures referenced by this event:

SCP-5, "Security Threat"
SCP-7, "Internal Disturbance"
SCP-8, "Hostage Situation"
SCP-10, "Discovery of Destructive Devices or Evidence of Malicious Acts"
SCP-11, "Civil Disturbance" DEVIATION None REFERENCES

. NUMARC NESP-007, HU4. l, HU4.2 Safeguards Contingency Plan EAL - 9.1.1 Rev. 00 Page 2 of 2

HCGS EAL/RALTechnical Basis 9.0 Hazards - Internal/External 9.1 Security Threats ALERT- 9.1.2 IC Security Event in a Plant Protected Area EAL Confirmed hostile intrusion or malicious acts as evidenced by ANY one of the following:

Discovery of an intruder(s), armed and violent, within the Protected Area, resulting in SCP-6 implementation
Hostage held on-site in a non-vital area, resulting in SCP-8 implementation OPERATIONAL CONDITION - All BASIS Security events classified under this EAL represent an *escalated threat to the level of safety of the plant. The event is confirmed if physical evidence supporting the hostile intrusion or assault exists. The intent of this EAL is to classify security events which represent an actual intrusion into the Protected Area. The SNSS/EC should declare an Alert upon consulting with the Security to determine the validity of the entry conditions.

Security Contingency Procedure (SCP) numbers are referenced following each EAL threshold. Since some SCP numbers appear in more than one EAL, the on-duty PSE&G Security Supervisor will provide information concerning the specific event to aid in classification. Barrier Analysis NIA ESCALATION CRITERIA Emergency Classification will be escalate to a Site Area Emergency .based upon a hostile intrusion or act in Plant Vital Areas. EAL- 9.1.2 Rev. 00 Page! 1 of 2

HCGS EAL/RALTechnical Basis DISCUSSION The following is an index of Security Contingency Procedures referenced by this event:

SCP-6, "Discovery of Intruders or Attack"
SCP-8, "Hostage Situation" DEVIATION None REFERENCES NUMARC NESP-007, HA4. l, HA4.2 Safeguards Contingency Plan EAL - 9.1.2 Rev. 00 Page 2 of 2

HCGS EAL/RALTechnical Basis 9.0 Hazards - Internal/External 9.1 Security Threats SITE AREA EMERGENCY - 9.1.3 IC Security Event in a Plant Vital Area EAL Confirmed hostile intrusion or malicious acts in Plant Vital Areas as evidenced by :

Discovery of an intruder(s), armed and violent, within a Plant Vital Area, resulting in SCP-6 implementation
Malicious acts or destructive device discovered in a Plant Vital Area resulting in SCP-10 implementation OPERATIONAL CONDITION - All BASIS Security events classified under this EAL represent an escalated threat to plant safety above that contained in an Alert in that a hostile intrusion or assault has progressed from the Protected Area to a Plant Vital Area. These areas contain vital equipment which includes any equipment, system, device or material, the failure, destruction or release of could directly or indirectly endanger the public health and safety by exposure to radiation. Equipment or systems which would be required to function to protect health and safety following such failure, destruction or release are also considered vital.

Security Contingency Procedure (SCP) numbers are referenced following each EAL threshold. Since some SCP numbers appear in more than one EAL, the on-duty PSE&G Security Supervisor will provide information concerning the specific event to aid in classification. Barrier Analysis NIA ESCALATION CRITERIA Emergency Classification will escalate to a General Emergency based upon the actual loss of physical control of the Main Control Room or Remote Shutdown Panel. EAL - 9.1.3 Rev. 00 Page 1 of 2

HCGS EAL/RALTechnical Basis DISCUSSION Plant Vital Areas are within the Protected Area and are generally controlled by card key readers. A hostile intrusion into a Plant Vital Area could represent a situation that threatens the safety of plant personnel and the general public. The following is an index of the Security Contingency Procedure referenced by this event:

SCP-6, "Discovery of Intruders or Attack"
SCP-10, "Discovery of Destructive Device or Evidence of Malicious Acts" DEVIATION None REFERENCES NUMARC NESP-007, HSI.I, HSI.2 Safeguards Contingency Plan EAL - 9.1.3 Rev. 00 Page 2 of 2

HCGS EAL/RALTechnical Basis 9.0 Hazards - Internal/External 9.1 Security Threats GENERAL EMERGENCY - 9.1.4 IC Security Event Resulting in Loss of Ability to Reach and Maintain Cold Shutdown EAL Security event resulting in the actual loss of physical control of EITHER one of the following:

Main Control Room
Remote Shutdown Panel OPERA TI ON AL CONDITION - All BASIS Security events classified under this EAL represent conditions under which a hostile force has taken physical control of areas required to reach and ma,intain Cold Shutdown. Both the Main Control Room and Remote Shutdown Panel are included, since control of either could hamper the operating crew's ability to perform a safe plant shutdown. Actual loss of physical control is defined as the condition where licensed Control Room Operators can no longer take required action to operate the plant, including unauthorized transfer of plant control from the Main Control Room.

Barrier Analysis NIA ESCALATION CRITERIA NIA DISCUSSION Security threats which meet the threshold for declaration of a General Emergency are an actual loss of physical control of the Main Control Room or the Remote Shutdown Panel. This situation places the plant in a potentially unstable condition with high potential of multiple fission product barrier failures. EAL- 9.1.4 Rev. 00 Page 1 of 2

HCGS EALIRALTechnical Basis DEVIATION None REFERENCES NUMARC NESP-007, HGI.l, HGI.2 Safeguards Contingency Plan EAL - 9.1.4 Rev. 00 Page 2 of 2

HCGS EALIRALTechnical Basis 9.0 Hazards - Internal/External 9.2 Fire UNUSUAL EVENT - 9.2.1 IC Fire within the Protected Area Boundary Not Extinguished within 15 minutes of Detection EAL Valid Fire Alarm is received in the Main Control Room OR Report of a fire from personnel at the scene Fire is within ANY one of the following Plant Structures (EXCLUDING small fires that have NO potential to affect Safety Systems or Protected Area Permanent Plant Structures)

Reactor Building
Turbine Building
Control/Aux Building
Service Water Intake Structure
Service/Rad Waste Building
Low Level Radwaste Interim Storage Facility Fire is NOT extinguished within 15 minutes of EITHER one of the following:
Receipt of a Valid Fire Alarm
Report of a fire from the scene OPERA TI ON AL CONDITION - All BASIS Fires classified under this EAL include those of a magnitude and extent that may be a potential precursor to damage to Safety Systems, and hence have safety significance. This EAL includes Plant Vital Structures and also structures and areas that are contiguous to Plant Vital Structures, due to the potential for a fire to spread from a non-safety related structure to an adjoining safety related structure.

EAL- 9.2.1 Rev.00 Page 1 of 3

HCGS EAL/RALTechnical Basis A fire alarm received in the Main Control Room is considered to be Valid when the alarm is substantiated by the receipt of related independent alarms (fire, temperature, deluge, etc.) in the Main Control Room or by visual confirmation if only a single detector is alarming. This EAL EXCLUDES such items as fires in Plant Structures other than those listed in the EAL, waste-basket fires, and other small fires of no safety significance based on the judgment of the SNSS that NO potential to affect a Safety System exists. Emergency Coordinator judgment must be exercised to determine if a fire within a Plant Structure is of any safety significance. The 15 minute clock starts upon receipt ofa Valid Fire Alarm or report of a fire from personnel at the scene. 15 minutes was determined to be a reasonable time limit for small fires to be extinguished. A Safety System is defined as any system or component included within the Technical Specification. Fire is defined as combustion characterized by the generation heat and smoke. Sources of smoke such as overheated electrical equipment and slipping drive belts, for example, do not constitute fires. Observation of a flame is preferred but is NOT required if large quantities of smoke and heat are observed. Barrier Analysis NIA ESCALATION CRITERIA Emergency Classification will escalate to an Alert if the fire damages more than one plant Safety System or damages any Plant Vital Structures. DISCUSSION The presence of a fire within the specified areas must be evaluated to determine the potential impact on Safety Systems, even if initial reports are that the fire is effecting a non-safety related p~>rtion of the plant, but has the potential to spread. Excluded non-vital structures include: Circulating Water Structure Station Service Transformer and Switchyard Area Hope Creek Admin. Building Onsite Warehouses Onsite Trailers Main and Aux Guardhouse Nuclear Services Building Auxiliary Boiler House EAL-9.2.1 Rev.00 Page 2 of 3

HCGS EAL/RALTechnical Basis DEVIATION None REFERENCES ~CNESP-007,HlJ2 HCGS Fire & Medical Emergency Response; HC.FP-EO.ZZ-0001 (Z) NUMARC Questions and Answers, June 1993, "Hazards Question #7" EAL- 9.2.l Rev. 00 Page 3 of 3

HCGS EAL/RALTechnical Basis 9.0 Hazards - Internal/External 9.2 Fire ALERT- 9.2.2 IC Fire Affecting the Operability of Plant Safety Systems Required to Establish or Maintain Safe Shutdown EAL Fire within ANY one of the following Plant Vital Structures:

Reactor Building
Control/Aux Building
Service Water Intake Structure
Service/Rad Waste Building The Fire is of a magnitude that it SPECIFICALLY results in Damage to ANY one of the following:
TWO OR MORE subsystems of a Safety System
MORE THAN ONE Safety System
Any Plant Vital Structure which renders the structure incapable of performing its Design Function Damaged Safety System(s) or Plant Vital Structure is required for the present Operational Condition OPERA TI ON AL CONDITION - All BASIS The primary concern in this EAL is the magnitude of the fire and the effects on Safety Systems required for the present Operational Condition. Specific system degradation is addressed in the System Malfunction EALs. A detailed assessment of system damage is not required prior to classification. The term "Damage" is defined as evidence that the fire has caused component malfunction (pump trip, breaker trip, etc.) or a report of visible scorching, blistering or other deformation that may have resulted in the equipment/structure being INOPERABLE or otherwise EAL- 9.2.2 Rev. 00 Page 1 of 3

HCGS EALIRALTechnical Basis incapable of performing it's design function. A Safety System is defined as any system required to maintain safe operation or to establish or maintain Cold Shutdown. In those cases where it is believed that the fire may have caused damage to Safety Systems, then an Alert declaration is warranted, since the full extent of the damage may not be known. For Plant Vital Structure damage, classification is required under this EAL ifthe structure houses or otherwise supports Safety Systems required for the present Operational Condition. For example, a fire that has been confirmed to be localized to a single piece of equipment, like a 4.16 KV Breaker, with no potential to spread to adjacent equipment, does not warrant classification as an Alert. In the event, however, that the fire has spread or is believed to be spreading to other 4.16 KV Breakers for component(s) required for the present operating condition, then an Alert is warranted. Fire is defined as combustion characterized by the generation heat and smoke. Sources of smoke such as overheated electrical equipment and slipping drive belts, for example, do not constitute fires. Observation of a flame is preferred, but is NOT required if large quantities of smoke and heat are observed. Barrier Analysis NIA ESCALATION CRITERIA Emergency Classification will escalate based on further damage to plant safety systems, loss of fission product barriers, or abnormal radiological releases. The EC may use Emergency Coordinator Discretion and escalate the classification to SAE based on the nature of the fire. DISCUSSION No lengthy and time consuming assessment of damage is required prior to classification. In this EAL, no attempt is made to quantify the magnitude of the damage to any Safety System:, but instead an attempt is made to identify any damage in order to quantify the magnitude and extent of the fire. In short, if the fire is big enough that it has damaged MORE THAN ONE Safety System, or more than one subsystem of a Safety System, then the fire is big enough to justify an Alert declaration. Damage to Plant Vital Structures must be to the extent that EC judgment must be used to determine ifthe structure is still capable of performing its design function. Electrical failures (such as shorts, grounds, arcing, etc.) should be evaluated for the possibility of a fire. Any security aspects of this event should be considered under EAL sections covering Security Events. DEVIATION EAL-9.2.2 Rev. 00 Page 2 of 3

HCGS EALIRALTechnical Basis None REFERENCES NU?v1ARCNESP-007,HA2 HCGS Fire & Medical Emergency Response; HC.FP-EO.ZZ-OOOI(Z) HCGS Technical Specifications Section 3/4 7-6, Control Room Emergency Filtration System NU?v1ARC Questions and Answers, June 1993, "Hazards Question #7" EAL- 9.2.2 Rev. 00 Page 3 of 3

HCGS EALIRALTechnical Basis 9.0 Hazards - InternaVExternal 9.3 Explosion UNUSUAL EVENT - 9.3.1 IC Natural and Destructive Phenomena Affecting the Protected Area EAL Confirmed Explosion within the Protected Area Report of visible damage to Plant equipment or Protected Area Permanent Plant Structures OPERATIONAL CONDITION - All BASIS Occurrence of these event within the Protected Area, that cause visible damage to plant equipment or Protected Area Permanent Plant Structures warrant declaration as an Unusual Event under this EAL. Confirmed Explosions outside the Protected Area should not be classified under this EAL. No attempt should be made to assess the magnitude of the damage. The confirmed occurrence of the explosion with a report of damage (deformation/scorching) is sufficient for declaration. A confirmed explosion is defined as visual evidence that a rapid, unconfined combustion, or a catastrophic failure of pressurized equipment that imparts energy of sufficient force to damage or potentially damage Permanent Plant Structures, systems or components. Barrier Analysis NIA ESCALATION CRITERIA Emergency Classification will escalate to Alert if the explosion damages more than one Safety Systems or damages any Plant Vital Structure. EAL- 9.3.1 Rev. 00 Page 1 of 2

HCGS EAL/RALTechnical Basis DISCUSSION Electrical failures (such as shorts, grounds, arcing, etc.) should not be considered an explosion; however, they should be evaluated for the possibility of a fire. Any security aspects of this event should be considered under EAL sections covering Security Events. DEVIATION None REFERENCES NUMARC NESP-007, HUI .5 HCGS Fire & Medical Emergency Response; HC.FP-EO.ZZ-OOOl(Z) EAL - 9.3.1 Rev. 00 Page 2 of 2

HCGS EALIRALTechnical Basis 9.0 Hazards - Internal/External 9.3 Explosion ALERT - 9.3.2 IC Explosion Affecting the Operability of Plant Safety Systems Required to Establish or Maintain Safe Shutdown EAL Confirmed Explosion within ANY one of the following Plant Vital Structures:

Reactor Building
Control/Aux Building
Service Water Intake Structure e . Service/Rad Waste Building The Explosion is of a magnitude that it SPECIFICALLY results in Damage to ANY one of the following:
TWO OR MORE subsystems of a Safety System
MORE THAN ONE Safety System
Any Plant Vital Structure which renders the structure incapable of performing its Design Function Damaged Safety System(s) or Plant Vital Structure is required for the present Operational Condition OPERA TI ON AL CONDITION - All BASIS The primary concern in this EAL is the magnitude of the explosion and the effects on safety systems required for the present Operational Condition. Specific system degradation is addressed in the System Malfunction EALs. A detailed assessment of system damage is not required prior to classification. The term "Damage" is defined as evidence that the explosion has caused component malfunction (pump trip, breaker trip, etc.) that may have resulted in the equipment/structure being INOPERABLE or otherwise incapable of performing it's design function. A Safety System is defined as any system required to maintain safe operation or to EAL - 9.3.2 Rev. 00 Page 1 of 2

HCGS EALIRALTechnical Basis establish or maintain Cold Shutdown . In those cases where it is believed that the explosion may have caused damage to Safety Systems, then an Alert declaration is warranted, since the full extent of the damage may not be known. For Plant Vital Structure damage, classification is required under this EAL if the structure houses or otherwise supports safety systems required for the present Operational Condition. A confirmed explosion is defined as visual evidence that a rapid, unconfined combustion, or a catastrophic failure of pressurized equipment that imparts energy of sufficient force to damage or potentially damage Permanent Plant Structures, systems or components. Barrier Analysis NIA ESCALATION CRITERIA Emergency Classification will escalate based on further damage to plant safety systems, loss of fission product barriers, or abnormal radiological releases. The EC may use Emergency Coordinator Discretion and escalate the classification to SAE based on the nature of the explosion. DISCUSSION No lengthy and time consuming assessment of damage is required prior to classification. In this EAL, no attempt is made to quantify the magnitude of the damage to any safety system but instead an attempt is made to identify any damage in order to quantify the magnitude and extent of the explosion. In short, if the explosion is big enough that it has damaged MORE THAN ONE safety system, or more than one subsystem of a Safety System, then the explosion is big enough to justify an Alert declaration. Damage to Plant Vital Structures must be to the extent that EC judgment must be used to determine if the structure is still capable of performing its design function. Electrical failures (such as shorts, grounds, arcing, etc.) should not be considered an explosion; however, they should be evaluated for the possibility of a fire. Any security aspects of this event should be considered under EAL sections covering Security Events. DEVIATION None REFERENCES NUMARC NESP-007, HA2 HCGS Fire & Medical Emergency Response; HC.FP-EO.ZZ-OOOl(Z) EAL- 9.3.2 Rev.00 Page 2 of 2

HCGS EALIRALTechnical Basis 9.0 Hazards - Internal/External 9.4 Toxic/Flammable Gases UNUSUAL EVENT- 9.4.1.a IC Release of Toxic or Flammable Gases Deemed Detrimental to Safe Operation of the Plant EAL Notification by Local, County, or State Officials for the potential need to EVACUATE non-essential personnel due to an Offsite Toxic Gas release SNSS deems evacuation of non-essential personnel is required OPERATIONAL CONDITION - All BASIS Notification by Local, County, or State Officials for the potential need to EVACUATE non-essential personnel due to an Offsite Toxic Gas release, along with SNSS concurrence that such action is appropriate warrants declaration of an Unusual Event, since a release that has occurred offsite, may have an impact on routine plant operations. An offsite event (such as a tanker accident or a barge accident) may place the Protected Area within the evacuatian area. The evacuation is determined from the DOT Evacuation Tables for Selected Hazardous Materials in the DOT Emergency Response Guide for Hazardous Materials. A Toxic Gas is considered to be any substance that is dangerous to life or limb by reason of inhalation or skin contact. A Toxic Gas release is considered to be a threat to plant personnel if concentrations are high enough to endanger the health of those personnel. Barrier Analysis NIA EAL - 9.4.1.a Rev. 00 Page 1 of 2

HCGS EALIRALTechnical Basis. ESCALATION CRITERIA Emergency Classification will escalate to an Alert if the Toxic Gas enters either a Plant Vital Area or an area contiguous to a Plant Vital Area. DISCUSSION None DEVIATION None REFERENCES NUMARC NESP-007, HU3.l and HU3.2 HC.OP-AB.ZZ-0129 (Q), High Radiation, Smoke, or Toxic Gases in the Control Room Air Supply HCGS Technical Specifications Section 3/4 7-6, Control Room Emergency Filtration System EAL - Q.4.1.a Rev. 00 Page 2 of 2

HCGS EAL/RALTechnical Basis 9.0 Hazards - Internal/External 9.4 Toxic/Flammable Gases UNUSUAL EVENT- 9.4.1.b IC Release of Toxic or Flammable Gases Deemed Detrimental to Safe Operation of the Plant EAL Uncontrolled Toxic Gas release within the Protected Area in ANY area which does not normally require an atmospheric survey or Respiratory Protection for entry Routine Plant Operations are IMPEDED based on EITHER one of the following:

Access restrictions caused by the uncontrolled release
Personnel injuries have occurred as a result of the release OPERATIONAL CONDITION - All BASIS An uncontrolled Toxic Gas release within the Protected Area, in high enough concentrations, will adversely affect the health and safety of plant personnel, along with the safe operation of the plant. This EAL specifically addresses those areas within the Protected Area that do not normally require an atmospheric survey or Respiratory Protection for entry, since the atmosphere in an area that does require an atmospheric survey or Respiratory Protection does not meet the intent of this EAL.

Releases classified under this EAL include those that originate both onsite and offsite. A Toxic Gas is considered to be any substance that is dangerous to life or limb by reason of inhalation or skin contact. Uncontrolled Toxic Gas releases are considered to be those releases that cannot be isolated I confined to a single compartment or area, or are not as the result of a designed plant safety feature. For example, an uncontrolled release of chlorine/ammonia into the Turbine Building warrants declaration of an Unusual Event. A Cardox discharge inside any area that contains this safety feature (i.e. Diesel Bays) does not warrant Unusual Event declaration, unless personnel injuries have occurred as a direct result of the discharge. EAL - 9.4.1.b Rev. 00 Page 1 of 2

HCGS EAL/RALTechnical Basis A Toxic Gas release is considered to be IMPEDING normal plant operations if concentrations are high enough to restrict routine operator movements. Access restrictions includes those conditions where access is only possible with appropriate personnel protection equipment, since this equipment restricts normal vision and mobility. Barrier Analysis NIA ESCALATION CRITERIA Emergency Classification will escalate to an Alert if the Flammable Gas enter either a Plant Vital Area or an area contiguous to a Plant Vital Area. DISCUSSION This EAL should not be construed to include confined spaces that must be ventilated prior to entry or situations involving Site Protection personnel who are using respiratory equipment during the performance of their duties unless it also affects personnel not involved with Site Protection activates. These areas include the Drywell (when inerted) and ALL Confined Spaces. In addition, those situations that require personnel to wear respiratory protection equipment as the result of airborne contamination as required by Radiation Protection personnel do not meet the intent of this EAL. An offsite event (such as a tanker accident or a barge accident) may place the Protected Area within the evacuation area. The evacuation is determined from the DOT Evacuation Tables for Selected Hazardous Materials in the DOT Emergency Response Guide for Hazardous Materials. DEVIATION None REFERENCES NUMARC NESP-007, HU3. l and HU3.2 HC.OP-AB.ZZ-0129 (Q), High Radiation, Smoke, or Toxic Gases in the Control Room Air Supply HCGS Technical Specifications Section 3/4 7-6, Control Room Emergency Filtration System EAL - 9.4.1.b Rev. 00 Page 2 of 2

HCGS EAL/RALTechnical Basis 9.0 Hazards - Internal/External 9.4 Toxic/Flammable Gases UNUSUAL EVENT - 9.4.1.c IC Release of Toxic or Flammable Gases Deemed Detrimental to Safe Operation of the Plant EAL Uncontrolled Flammable Gas release within the Protected Area that RESULTS in Flammable Gas concentrations EXCEEDING 25% of the LEL Routine Plant Operations are IMPEDED based on EITHER one of the following:

Access restrictions caused by the uncontrolled release
Personnel injuries have occurred as a result of the release OPERATIONAL CONDITION - All BASIS An uncontrolled Flammable Gas release within the Protected Area, in high enough concentrations, will adversely affect the health and safety of plant personnel, along with the safe operation of the plant. This EAL specifically addresses those conditions where a Flammable Gas concentration EXCEEDING 25% of the LEL (Lower Explosive Limit) exists anywhere within the Protected Area. Releases classified under this EAL include those that originate both onsite and offsite.

A Flammable Gas is considered to be any substance that can result in an ignition, sustained burn or detonation. Uncontrolled Flammable Gas releases are considered to be those releases that cannot be isolated I confined to a single compartment or area. For example, an uncontrolled release of hydrogen into the Turbine Building in concentration exceeding 25 % of the LEL warrants declaration of an Unusual Event. In comparison, a controlled release of Hydrogen during Generator purging or Hydrogen Tank trailer purging does not warrant event declaration, as these evolutions are controlled. EAL - 9.4.1.c Rev. 00 Pagt! 1 of 2

HCGS EAL/RALTechnical Basis includes those conditions where access is only possible with appropriate personnel protection equipment, since this equipment restricts normal vision and mobility. Barrier Analysis NIA ESCALATION CRITERIA Emergency Classification will escalate to an Alert if the Flammable Gas enter either a Plant Vital Area or an area contiguous to a Plant Vital Area. DISCUSSION For Hydrogen Gas, the explosive limit is 4%. Hence, a threshold of25% of the LEL equates to 1% Hydrogen. This EAL should not be construed to include those controlled evolutions that may discharge a Flammable Gas within the Protected Area, but present no danger to plant safety, since the evolution is planned and controlled. An offsite event (such as a tanker accident or a barge accident) may place the Protected Area within the evacuation area. The evacuation is determined from the DOT Evacuation Tables for Selected Hazardous Materials in the DOT Emergency Response Guide for Hazardous Materials. DEVIATION None REFERENCES NUMARC NESP-007, HU3. land HU3.2 HC.OP-AB.ZZ-0129 (Q), High Radiation, Smoke, or Toxic Gases in the Control Room Air Supply HCGS Technical Specifications Section 3/4 7-6, Control Room Emergency Filtration System EAL - 9.4.1.c Rev. 00 Page 2 of 2

HCGS EALIRALTechnical Basis 9.0 Hazards - Internal/External 9.4 Toxic/Flammable Gases ALERT - 9.4.2.a IC Release of Toxic or Flammable Gases Within a Facility Structure Which Jeopardizes Operation of Systems Required to Maintain Safe Ope1 ations or to Establish or Maintain Cold Shutdown Conditions EAL Uncontrolled Toxic Gas release within ANY one of the following Plant Structures

Reactor Building
Turbine Building
Control/Aux Building
Service Water Intake Structure
Service/Rad Waste Building Toxic Gas concentrations result in ANY one of the.following:
An IDLH atmosphere
Plant personnel report severe adverse health reactions, including burning eyes, nose, throat, or dizziness
The Threshold Limit Value (TLV) being EXCEEDED Plant personnel are unable to perform actions necessary to complete a Safe Shutdown of the plant without appropriate personnel protection equipment OPERATIONAL CONDITION - All BASIS An uncontrolled Toxic Gas release entering any of the plant structures listed in the EAL, that threatens the ability of plant personnel to perform actions required for safe shutdown of the plant, warrants declaration of an Alert. The EAL threshold includes those conditions that present a significant challenge to plant personnel. This EAL specifically addresses only those plant EAL - 9.4.2.a Rev. 00 Page 1 of 3

HCGS EAL!RALTechnical Basis structures that either contain safe shutdown equipment or are contiguous to those areas. Release classified under this EAL include those that originate both onsite and offsite. A Toxic Gas is considered to be any substance that is dangerous to life or limb by reason of inhalation or skin contact. Uncontrolled Toxic Gas releases are considered to be those releases that can not be isolated I confined to a single compartment or area, or are not as the result of a designed plant safety feature. For example, an uncontrolled release of chlorine/ammonia into the Turbine Building that directly effects plant personnel, warrants declaration of an Alert. A Cardox discharge inside any area that contains this safety feature (i.e. Diesel Bays) does not warrant Alert declaration, unless personnel injuries have occurred as a direct result of the discharge. Barrier Analysis NIA ESCALATION CRITERIA Emergency Classification will escalated based on further damage to plant safety systems, loss of fission product barriers, or abnormal radiological releases. The EC may use Emergency Coordinator Discretion and escalate the classification to SAE based on the nature of the toxic gas release. DISCUSSION Access is considered impeded if the Toxic Gas concentrations are life threatening, i.e. require the use of personnel protective equipment. Use of protective equipment also limits the mobility and vision. The cause or magnitude of the gas concentration is not the major concern in this EAL, but rather that access required to an area that may be impeded. An IDLH atmosphere is any atmosphere that is determined to be Immediately Dangerous to Life and Health. This EAL should not be construed to include confined spaces that must be ventilated prior to entry or situations involving Site Protection personnel who are using respiratory equipment during the performance of their duties unless it also affects personnel not involved with Site Protection activities. These areas include the Drywell (when inerted) and ALL Confined Spaces. In addition, those situations that require personnel to wear respiratory protection equipment as the result of airborne contamination as required by Radiation Protection personnel do not meet the intent of this EAL. An offsite event (such as a tanker accident or a barge accident) may place the Protected Area within the evacuation area. The evacuation is determined from the DOT Evacuation Tables for Selected Hazardous Materials in the DOT Emergency Response Guide for Hazardous Materials. EAL - 9.4.2.a Rev. 00 Page 2 of 3

HCGS EAL/RALTechnical Basis DEVIATION None REFERENCES NUMARC NESP-007, HA3. l and HA3.2 HC.OP-AB.ZZ-0129 (Q), High Radiation, Smoke, or Toxic Gases in the Control Room Air Supply HCGS Technical Specifications Section 3/4 7-6, Control Room Emergency Filtration System EAL - 9.4.2.a Rev.00 Page 3 of 3

HCGS EAL/RALTechnical Basis 9.0 Hazards - Internal/External 9.4 Toxic/Flammable Gases ALERT - 9.4.2.b IC Release of Toxic or Flammable Gases Within a Facility Structure Which Jeopardizes Operation of Systems Required to Maintain Safe Operations or to Establish or Maintain Cold Shutdown Conditions EAL Uncontrolled Flammable Gas release within ANY one of the following Plant Structures

Reactor Building
Turbine Building
Control/Aux Building

Service Water Intake Structure
Service/Rad Waste Building Flammable Gas concentrations EXCEED 50% of the LEL Plant personnel are unable to perform actions necessary to complete a Safe Shutdown of the plant without appropriate personnel protection equipment OPERATIONAL CONDITION - All BASIS An uncontrolled Flammable Gas release entering any of the plant structures listed in the EAL, that threatens the ability of plant personnel to perform actions required for safe shutdown of the plant, warrants declaration of an Alert. The EAL threshold includes those conditions that present a significant challenge to plant personnel. This EAL specifically addresses only those plant structures that either contain safe shutdown equipment or are contiguous to those areas. Release classified under this EAL include those that originate both onsite and offsite. A Flammable Gas is considered to be any substance that is capable of being easily ignited or burning quickly.

Uncontrolled Flammable Gas releases are considered to be those releases that can not be EAL - Q.4.2.b Rev. 00 Page 1 of 2

HCGS EAL/RALTechnical Basis isolated I confined to a single compartment or area, or are not as the result of a designed plant safety feature. For example, an uncontrolled release of hydrogen into the Turbine Building in concentration exceeding 50% of the LEL (Lower Explosive Limit) warrants declaration of an Alert. In comparison, a controlled release of Hydrogen during Generator purging does not warrant event declaration, as this evolution is controlled. Barrier Analysis NIA ESCALATION CRITERIA Emergency Classification will escalated based on subsequent damage to plant safety systems, loss of fission product barriers, or abnormal radiological releases. The EC may use discretion and escalate the classification to SAE based on the nature of the flammablegas release. DISCUSSION For Hydrogen Gas, the explosive limit is 4%. Hence, a threshold of 50% of the LEL equates to 2% Hydrogen. This EAL should not be construed to include those controlled evolutions that may discharge a Flammable Gas within the Protected Area, but present no danger to plant safety, since the evolution is planned and controlled. An offsite event (such as a tanker accident or a barge accident) may place the Protected Area within the evacuation area. The evacuation is determined from the DOT Evacuation Tables for Selected Hazardous Materials in the DOT Emergency Response Guide for Hazardous Materials. DEVIATION None REFERENCES NUMARC NESP-007, HA3. l and HA3.2 HC.OP-AB.ZZ-0129 (Q), High Radiation, Smoke, or Toxic Gases in the Control Room Air Supply HCGS Technical Specifications Section 3/4 7-6, Control Room Emergency Filtration System EAL - 9.4.2.b Rev. 00 Page 2 of 2

HCGS EALIRALTechnical Basis 9.0 Hazards - Internal/External 9.5 Seismic Event UNUSUAL EVENT - 9.5.1.a I 9.5.1.b IC Natural and Destructive Phenomena Affecting the Protected Area EAL EITHER one of the following conditions:

Seismic Event felt by personnel within the Protected Area
Valid Actuation of the Seismic Trigger (>O.Olg) has occurred as verified by the SMA-3 Event Indicator (flag) being WHITE on Panel 10-C-673 in the Upper Relay Room OPERATIONAL CONDITION - All BASIS The condition that the Seismic Event has been felt by personnel within the Protected Area, or a Valid Actuation of the Seismic Trigger indicates that a Seismic Event of a magnitude greater than O.Olg has occurred. This threshold warrants declaration of an Unusual Event.
Valid is defined as the Seismic Trigger actuation being the direct result of a Seismic Event.

Classification should be based on an actuation of the Seismic Trigger as verified in the Upper Relay Room. Additional information can be obtained by contacting the National Earthquake Center in Denver, Colorado at (303) 273-8500. However, it is important to realize that it will take the Earthquake Center approximately 30 minutes to provide the requested information. The time required to obtain this additional information should not result in a delay of event classification for a valid actuation. Barrier Analysis NIA EAL - 9.5.1.a/ 9.5.1.b Rev. 00 Page 1 of 2

HCGS EAL/RALTechnical Basis ESCALATION CRITERIA Emergency Classification will escalate to an Alert if the a subsequent seismic event occurred in excess of the Operating Basis Earthquake level (O. lg). DISCUSSION An earthquake of a magnitude equivalent to O.Olg is not expected to affect the capability of plant safety functions. This threshold value is well below the Operating Basis Earthquake (QBE) level of O.lg. An approximate relationship between acceleration and magnitude is as follows: An Acceleration of: is approx. equal to a Richter Scale Magnitude of: O.Olg 4.0 0.02g 4.5 0.lg 5.5 0.2g 6.5 DEVIATION None REFERENCES NUMARC NESP-007, HUl.l HC.OP-AB.ZZ-0139 (Q), Acts ofNature HCGS Technical Specification Section 3/4.3.7.2, Seismic Monitoring Instrumentation HC.OP-SO.SG-0001 (Z), Seismic Instrumentation System Operation HC.OP-AR.ZZ-0011 (Q), Overhead Annunciator Window Box C6 EAL - 9.5.1.a/ 9.5.1.b Rev. 00 Page 2 of 2

HCGS EAL/RALTechnical Basis 9.0 Hazards - Internal/External 9.5 Seismic Event ALERT - 9.5.2 IC Natural and Destructive Phenomena Affecting the Plant Vital Area EAL Seismic Event felt by personnel within the Protected Area Valid Actuation of the Seismic Trigger (>O.Olg) has occurred as verified by the SMA-3 Event Indicator (flag) being WHITE on Panel 10-C-673 in the Upper Relay Room Valid Actuation of the Seismic Switch (>0.1 g) has occurred as verified by EITHER one of the following:

Valid Actuation of Main Control Room Overhead Annunciator C6-C4
AMBER Alarm light on the Seismic Switch Power Supply Drawer is lit on Panel 10-C-673 in the Upper Relay Room OPERATIONAL CONDITION - All BASIS A Valid Actuation of the Seismic Switch indicates that a Seismic Event of a magnitude greater than O. lg (Operating Basis Earthquake) has occurred. The Salem SNSS must be informed of this information immediately. At this level, plant safety systems are designed to remain functional and within design stress and deformation limits. Thus, an earthquake of this magnitude is not expected to affect the capability of plant safety functions required to shut down the plant and place it in a cold shutdown condition.

This threshold warrants declaration of an Alert. Valid is defined as the Seismic Switch actuation being the direct result of a Seismic Event. The condition that the Seismic Event has been felt by personnel within the Protected Area or along with Seismic Trigger actuation provides further confirmation that an event has occurred. Classification should be based on a Valid actuation of the Seismic Switch as verified in the Upper Relay Room. Additional information can be obtained EAL- 9.5.2 Rev.00 Page 1 of 2

HCGS EALIRALTechnical Basis by contacting the National Earthquake Center in Denver, Colorado at (303) 273-8500. However, it is important to realize that it will take the Earthquake Center approximately 30 minutes to provide the requested information. The time required to obtain this additional information should not result in a delay of event classification for a valid actuation. Barrier Analysis NIA ESCALATION CRITERIA Emergency Classification will escalate if the seismic event caused additional damage to plant safety systems, loss of fission product barriers, or abnonnal radiological releases. The EC may use Emergency Coordinator Discretion and escalate the classification to SAE based on the nature of the event. DISCUSSION Seismic Event annunciation on panel 1OC673 would alert operators to this event and the active seismic monitoring instrumentation would begin to monitor the event. This threshold value associated with this EAL is well below the Design Basis Earthquake of 0.2g that is the maximum seismic event that is expected to occur based on local geological and seismological factors. An approximate relationship between acceleration and magnitude is as follows: Acceleration: Richter Scale Magnitude (approximate): O.Olg 4.0 0.02g 4.5 O.lg 5.5 0.2g 6.5 DEVIATION None REFERENCES NUMARC NESP-0007, HAI. I HC.OP-AB.ZZ-0139 (Q), Acts of Nature HCGS Technical Specification Section 3/4.3.7.2, Seismic Monitoring Instrumentation HC.OP-SO.SG-0001 (Z), Seismic Instrumentation System Operation HC.OP-AR.ZZ-0011 (Q), Overhead Annunciator Window Box C6 EAL- 9.5.2 Rev. 00 Page 2 of 2

HCGS EAL/RALTechnical Basis 9.0 Hazards - Internal/External 9.6 High Winds UNUSUAL EVENT - 9.6.1.a IC Natural and Destructive Phenomena Affecting the Protected Area EAL Report of a Tornado TOUCHING DOWN within the Protected Area OPERATIONAL CONDITION - All BASIS A tornado touching down within the Protected Area is an observed event with the potential to cause damage to structures containing systems or functions necessary for safe shutdown of the plant. As such, a tornado represents a potential degradation in the level of safety of the plant. Verification of the tornado should be by direct observation and report by plant personnel. Barrier Analysis NIA ESCALATION CRITERIA Emergency Classification will escalate to an Alert ifthe tornado causes damage to Plant Vital Structures or affects the operability of Technical Specification required equipment DISCUSSION The National Weather Service can be contacted for further information about existing or projected Adverse Weather Conditions: Wilmington (302) 573-6142 Mount Holly (609) 261-6604 Mount Holly (609) 261-6602 DEVIATION None EAL - 9.6.1.a Rev. 00 Page 1 of 2

HCGS EAL/RALTechnical Basis REFERENCES NUMARC NESP-007, HUl.2 and HUI.7 HC.OP-AB.ZZ-0139 (Q), Acts of Nature HCGS Technical Specification Section 3/4, 3.7.3, Meteorological Monitoring Instrumentation HCGS UFSAR Sections 2.3, 3.3.1 EAL - 9.6.1.a Rev.DO Page 2 of 2

HCGS EAL/RALTechnical Basis 9.0 Hazards - Internal/External 9.6 High Winds UNUSUAL EVENT - 9.6.1.b IC Natural and Destructive Phenomena Affecting the Protected Area EAL Sustained wind speeds > 75 MPH for 15 minutes at ANY elevation of the Met Tower OPERATIONAL CONDITION - All BASIS Sustained wind speeds in excess of 75 MPH are of sufficient velocity to have the potential to cause damage to Plant Vital Areas. These conditions are indicative of unstable weather conditions and represent a potential degradation in the level of safety of the plant. The wind speed threshold is well below the structure design basis of I 08 mph, and is set at the value used to characterize Hurricane force winds. Sustained wind speed means wi.nds in excess of the threshold value for greater than 15 minutes. Barrier Analysis NIA ESCALATION CRITERIA Emergency Classification will escalate to an Alert if the high winds cause damage to Plant Vital Structures or affects the operability of Technical Specification required equipment. DISCUSSION Verification of sustained wind speed will be by observation of meteorological tower data. The Wind Speed indication from the Met Tower instrumentation is full scale at I 00 mph. EAL - 9.6.1.b Rev.00 Page 1 of 2

HCGS EAL/RALTechnical Basis The National Weather Service can be contacted for further information about existing or projected Adverse Weather Conditions: Wilmington (302) 573-6142 Mount Holly (609) 261-6604 Mount Holly (609) 261-6602 -DEVIATION None REFERENCES NUMARC NESP-007, HUl.2 and HUl.7 HC.OP-AB.ZZ-0139 (Q), Acts of Nature HCGS Technical Specification Section 3/4, 3.7.3, Meteorological Monitoring Instrumentation HCGS UFSAR Sections 2.3, 3.3.1 EAL - 9.6.1.b Rev. 00 Page 2 of 2

HCGS EALIRALTechnical Basis 9.0 Hazards - Internal/External 9.6 High Winds . ALERT- 9.6.2 IC Natural an1 Destructive Phenomena Affecting the Plant Vital Area EAL EITHER one of the following:

Report ofa Tornado TOUCHING DOWN within the Protected Area
Sustained wind speeds> 75 MPH for 15 minutes, from at ANY elevation of the Met Tower The Wind Speed is of a magnitude that it SPECIFICALLY results in Damage to ANY one of the following:
TWO OR MORE subsystems of a Safety System
MORE THAN ONE Safety System
Rendering ANY of the following structures incapable of performing its Design Function:
Reactor Building
Control/Aux Building
Service Water Intake Structure
Service/Radwaste Building Damaged Safety System(s) or Plant Vital Structure is required for the present Operating Condition OPERA TI ON AL CONDITION - All BASIS The primary concern in this EAL is the magnitude of the high winds and the effects on safety functions required for the present Operational Condition. Specific system degradation is addressed in the System Malfunction EALs. A detailed assessment of system damage is not required prior to classification. The term "Damage" is defined as evidence that the high winds has caused component malfunction (pump trip, breaker trip, etc.) or a report of visible scorching, blistering or other deformation that may have resulted in the equipment/structure being EAL- 9.6.2 Rev. 00 Page 1 of 3

HCGS EAL!RALTechnical Basis INOPERABLE or otherwise incapable of performing it's design function. A Safety System is defined as any system required to maintain safe operation or to establish or maintain Cold Shutdown. In those cases where it is believeJ that the high winds may have caused damage to Safety Systems, then an Alert declaration is warranted, since the full extent of the damage may not be known. For Plant Vital Structure damage, classification is required under this EAL ifthe structure houses or otherwise supports safety systems required for the present Operational Condition. It is not intended that a lengthy engineering analysis be performed to determine if damage has affected structural design but EC judgment must determine whether to exclude minor exterior damage which does not affect the structural design capability. Sustained wind speed means winds in excess of the threshold value for greater than 15 minutes. Barrier Analysis NIA ESCALATION CRITERIA Emergency Classification will escalate based on further damage to plant Safety Systems, loss of fission product barriers, or abnormal radiological releases. The EC may use discretion and escalate the classification to SAE based on the nature of the winds. DISCUSSION The wind speed threshold is well below the structure design basis of I 08 mph, and is set at the value used to characterize Hurricane force winds. The Wind Speed indication from the Met Tower instrumentation is full scale at 100 mph. The National Weather Service can be contacted for further information about existing or projected Adverse Weather Conditions: Wilmington (302) 573-6142 Mount Holly ( 609) 261-6604 Mount Holly (609) 261-6602 DEVIATION None EAL-9.6.2 Rev.00 Page 2 of 3

HCGS EAL/RALTechnical Basis REFERENCES NUMARC NESP-007, HAI.2 and HAI.3 HC.OP-AB.ZZ-0139 (Q), Acts ofNature HCGS Technical Specification Section 3/4, 3.7.3, Meteorological Monitoring Instrumentation HCGS UFSAR Sections 2.3, 3.3.1 EAL - 9.6.2 Rev. 00 Page 3 of 3

HCGS EALIRALTechnical Basis 9.0 Hazards - Internal/External

9. 7 Flooding UNUSUAL EVENT- 9.7.1 IC Internal Flooding in Excess of Sump Handling Capability Affecting Safety Related Areas of the Plant EAL Visual Observation of Uncontrolled Flooding that confirms ANY one of the following:

Reactor Building Floor Levels above the Maximum Normal Floor Level (> l ")

referenced in EOP 103, Secondary Containment Control

Receipt of a SSWS Pump Room Flooded Alarm
Greater than 2" of water in ANY area that contains a Safety System(s), not included above OPERATIONAL CONDITION - All BASIS Uncontrolled flooding in the areas listed in the EAL represents the potential to directly impact continued safe operation of the plant. This EAL specifically addresses those areas of the plant where uncontrolled flooding presents a challenge to Safety System(s). Visual Observation of the flooding should occur prior to classification to validate any alarm conditions. Uncontrolled flooding is defined as event or condition that does not result from a controlled evolution.

Events classified under this EAL, for example, include the effects of flooding from system malfunctions, component failures, or repair activity failures (such as a failed freeze seal). Those events th:lt result in the flooding of an area as the direct result of a planned evolution, such as system draining in preparation for an equipment outage, do not warrant event classification, unless the draining can not be successfully terminated. Safety System is defined as any system or component included in the Technical Specification. Barrier Analysis NIA EAL- 9.7.1 Rev. 00 Page 1 of 2

HCGS EALIRALTechnical Basis ESCALATION CRITERIA Emergency Classification will escalate to an Alert if the flooding results in damage to equipment required for the present Operational Condition. DISCUSSION For the purpose of implementing this EAL, levels in the Reactor Building that would require classification under this EAL are defined as the Maximum Normal Floor Level in the EOPs. Exceeding this level in any of the Reactor Building areas would require running all available sump pumps. If level in these areas cannot be lowered to below the I " level, then systems discharging into this area are to be isolated, except for systems required to:

Ensure adequate core cooling

        ~       Shutdown the reactor

Protect primary containment integrity
Suppress a fire DEVIATION None REFERENCES NUMARC NESP-007, HUI.7 HC.OP-EO.ZZ-0103 (Q)-FC, Reactor Building Control HCGS Technical Specifications Section 3/4 7-3, Flood Protection EAL- 9.7.l Rev. 00 Page 2 of 2

HCGS EAL/RALTechnical Basis 9.0 Hazards - Internal/External

9. 7 Flooding ALERT- 9.7.2 IC Internal Flooding Affecting the Operability of Plant Safety Systems Required to Establish or Maintain Safe Shutdown EAL Visual Observation of Flooding within ANY one of the following Plant Vital Structures:

   ~   Reactor Building

Control/Aux Building
Service Water Intake Structure

   ,. Service/Rad Waste Building The Flooding is of a magnitude that it SPECIFICALLY results in Damage to ANY one of the following:

TWO OR MORE subsystems of a Safety System
MORE THAN ONE Safety System
Any of the above listed Plant Vital Structures which renders the structure incapable of performing its Design Function Damaged Safet)' System(s) or Plant Vital Structure is required for the present Operational Condition OPERATIONAL CONDITION - All BASIS The primary concern in this EAL is the magnitude of the internal flooding and the effects on Safety Systems required for the present Operational Condition. Specific system degradation is addressed in the System Malfunction EALs. A detailed assessment of system damage is not required prior to classification. The term "Damage" is defined as evidence that the internal flooding has caused component malfunction (pump trip, breaker trip, etc.) that may have resulted in the equipment/structure being INOPERABLE or otherwise incapable of performing it's design function. A Safety System is defined as any system required to maintain safe operation or to EAL- 9.7.2 Rev.00 Page 1 of 2

HCGS EAL/RALTechnical Basis establish or maintain cold shutdown. In those cases where it is believed that the internal flooding may have caused damage to Safety Systems, then an Alert declaration is warranted, since the full extent of the damage may not be known. For Plant Vital Structure damage, classification is required under this EAL ifthe structure houses or otherwise supports safety systems required for the present Operational Condition. Barrier Analysis NIA ESCALATION CRITERIA Emergency Classification will escalate based on damage to plant systems, loss of fission product barriers, or abnormal radiological releases. The EC may use discretion and escalate the classification to SAE based on the nature of the flooding. DISCUSSION Degraded system performance or observation of potential for damage that could degrade system performance is used as the indicator that the safety system operability was actually affected. A report of damage should not be interpreted as mandating a lengthy and timely assessment prior to justification; there is no inference in this EAL that the actual magnitude of damage be qualified or quantified. DEVIATION None REFERENCES NUMARC NESP-007, HAI.7 HCGS Technical Specifications EAL- 9.7.2 Rev. 00 Page 2 of 2

HCGS EAL/RALTechnical Basis 9.0 Hazards - InternaVExternal 9.8 Turbine Failure I Vehicle Crash I Missile Impact UNUSUAL EVENT - 9.8.1.a IC Natural and Destructive Phenomena Affecting Certain Structures W'.thin the Protected Area EAL Catastrophic damage to the Main Turbine as evidenced by EITHER one of the following:

Main Turbine casing penetration
Main Turbine/Generator Damage potentially releasing Lube Oil or Hydrogen Gas to the Turbine Building OPERATIONAL CONDITION - All BASIS Main Turbine failure of sufficient magnitude to cause damage to the turbine casing or generator seals increases the potential for leakage of combustible/explosive gases and of combustible liquids to the Turbine Building, warrants declaration of an Unusual Event. The presence of H 2 gas in sufficient quantities may present a flammable/explosive hazard. Turbine Lube Oil may also be

. present which may contribute to the flammability hazard. Barrier Analysis NIA ESCALATION CRITERIA Emergency Classification will escalate to an Alert based upon damage done by missiles generated by the failure. DISCUSSION Turbine rotating component failures may also result in other direct damage to plant systems and components. Damage may rupture the Turbine Lubricating Oil System, which would release flammable liquids to the Turbine Building. Potential rupture of the Main Condenser and condenser tubes may cause flooding in the lower levels of the Turbine Building. This damage should be readily observable. EAL - 9.8.1.a Rev. 00 Page 1 of 2

HCGS EAL/RALTechnical Basis Escape of hydrogen gas from the generator due to a loss of seal oil pumps or turbine lube oil without a turbine rotating component failure should not be classified under this event. DEVIATION None REFERENCES NUMARC NESP-007, HUI.6 EAL - 9.8.1.a Rev. 00 Page 2 of 2

HCGS EALIRALTechnical Basis 9.0 Hazards - Internal/External 9.8 Turbine Failure I Vehicle Crash I Missile Impact UNUSUAL EVENT - 9.8.1.b IC Natural and Destructive Phenomena Affecting Certain Structures Within the Protected Area EAL Vehicle Crash I Missile Impact with or within ANY one of the following Plant Structures:

Reactor Building
Turbine Building
Control/Aux Building
Service Water Intake Structure
Service/Radwaste Building OPERATIONAL CONDITION - All BASIS A Vehicle Crash I Missile Impact with or within a listed Plant Structure represents a potential challenge to plant safety. Events classified under this EAL include those of a magnitude and extent that may be a potential precursor to damage to Safety Systems, and hence has safety significance. Vehicle Crash includes Aircraft, Helicopters, Ships, Barges, or any other vehicle types of sufficient momentum to potentially damage the structure. Missile Impact includes flying objects from both offsite and onsite, rotating equipment or turbine failure causing turbine casing penetration.

Barrier Analysis

      *None ESCALATION CRITERIA Emergency Classification will escalate to Alert if the vehicle crash or missile impact causes damage to Plant Vital Structures.

EAL - 9.8.1.b Rev. 00 Page 1 of 2

HCGS EALIRALTechnical Basis DISCUSSION Any security aspects of this event should be considered under ECG Section 9.1, Security Events. DEVIATION None REFERENCES NUMARC NESP-007, HUl.4 NUMARC Questions and Answers, June 1993, "Hazards Question #6" EAL - 9.8.1.b Rev. 00 Page 2 of 2

                                                                 *HCGS EALIRALTechnical Basis 9.0 Hazards - Internal/External 9.8 Turbine Failure I Vehicle Crash I Missile Impact ALERT - 9.8.2 IC       Natural and Destructive Phenomena Affecting Certain Structures Within the Plant Vital Area EAL Vehicle Crash I Missile Impact with or within ANY one of the following Plant Vital Structures:

Reactor Building
Control/Aux Building
Service Water Intake Structure
Service/Rad Waste Building The Vehicle Crash I Missile Impact is of a magnitude that it SPECIFICALLY results

. in Damage to ANY one of the following:

TWO OR MORE subsystems of a Safety System
MORE THAN ONE Safety System
Any of the above Plant Vital Structures which renders the structure incapable of performing its Design Function Damaged Safety System(s) or Plant Vital Structure is required for the present Operational Condition OPERATIONAL CONDITION - All BASIS The primary concern in this EAL is the magnitude of the vehicle crashes I missile impact and the effects on safety systems required for the present Operational Condition. Specific system degradation is addressed in the System Malfunction EALs. A detailed assessment of system damage is not required prior to classification. The term "Damage" is defined as evidence that the vehicle crashes I missile impact has caused component malfunction (pump trip, breaker trip, etc.)

EAL- 9.8.2 Rev. 00 Page 1 of 2

HCGS EAL/RALTechnical Basis that may have resulted in the equipment/structure being INOPERABLE or otherwise incapable of performing it's design function. A Safety System is defined as any system required to maintain safe operation or to establish or maintain cold shutdown. In those cases where it is believed that the vehicle crashes I missile impact may have caused damage to Safety Systems, then an Alert declaration is warranted, since the full extent of the damage may not be known. For Plant Vital Structure damage, classification is required under this EAL if the struci:ure houses or otherwise supports safety systems required for the present Operational Condition. Barrier Analysis NIA ESCALATION CRITERIA Emergency Classification will escalate based on further damage to plant safety systems, fission product barriers, or abnormal radiation releases in other EAL sections. The EC may use discretion and escalate the classification based on the nature of the damage. DISCUSSION. This EAL is intended to address the threat to safety related equipment imposed by vehicle of missile impacts. No attempt should be made to assess the magnitude of damage to Safety Systems or Plant Vital Structures prior to classification. The evidence of damage is sufficient for declaration. DEVIATION None REFERENCES NUMARC NESP-007, HAl.5 and HAl.6 NUMARC Questions and Answers, June 1993, "Hazards Question #6" EAL - 9.8.2 Rev. 00 Page 2 of 2

HCGS EALIRALTechnical Basis 9.0 Hazards - InternaVExternal 9.9 River Level UNUSUAL EVENT- 9.9.1.a IC Natural and Destructive Phenomena Affecting the Protected Area EAL River Level > 99.5' OPERATIONAL CONDITION - All I BASIS This EAL indicates river level conditions that can threaten the level of safety at the plant due to flooding. Barrier Analysis NIA ESCALATION CRITERIA Emergency Classification will escalate based on damage to plant safety systems, loss of fission product barriers, or abnormal radiological releases in other EAL sections. DISCUSSION River level greater than 99.5' (+10.5' MSL) is indication of impending site flood conditions. Flood protection measures are required by Hope Creek Technical Specifications and procedure at 95.0' (+6.0' MSL). At this river level precautionary actions are taken, including filling outside tanks and ensuring that perimeter flood doors are closed. These actions ensure that the facility flood protection features are in place prior to a river level which would necessitate their use. The high river level threshold is at the river level that would require a plant shutdown. Hope Creek Technical Specification actions required by a river level of> 99.5' includes placing the plant in at least Hot Shutdown within the next 12 hours and in Cold Shutdown within the next 24 hours. EAL - 9.9.1.a Rev. 00 Page 1 of 2

HCGS EAL/RALTechnical Basis This circumstance is a Common Site Unusual Event. ~alem Technical Specifications actions require that perimeter flood doors are closed. This is based on the river level at which facility flood protection features provide protection to safety related equipment. The National Weather Service can be contacted for further information about existing or projected Adverse Weather Conditions: Mount Holly ( 609) 261-6604 Mount Holly (609) 261-6602 DEVIATION None REFERENCES NUMARC NESP-007, HUI.7 HC.OP-AB.ZZ-0139 (Q), Acts of Nature HCGS Technical Specification Section 3/4, 3.7.3, 3/4.7.1.3, 3/5.7.3 HCGS UFSAR, Section 2.4, Figure 2.4-3 SGS UFSAR, Section 2.4.11.2, Figure 3.4-1 . EAL - 9.9.1.a Rev. 00 Page 2 of 2

HCGS EALIRALTechnical Basis 9.0 Hazards - Internal/External 9.9 River Level UNUSUAL EVENT- 9.9.1.b IC Natural and Destructive Phenomena Affecting the Protected Area EAL River Level < 80.0' OPERATIONAL CONDITION - All I BASIS This EAL indicates a river level condition that is one foot lower than the historical low water level of 81.0' (-8.0' MSL) (December 31, 1962) and is higher than the Service Water pumps design level. Barrier Analysis NIA ESCALATION CRITERIA This event will be escalated based on damage to plant safety systems (Service Water pumps, Diesels, Cooling Water pumps, etc.) in the High Winds section, Heat Removal Capabilities, loss of Fission Product Barriers, or abnormal Radiological Releases section. DISCUSSION River level less than 80.0' (-9.0' MSL) is indication of approaching loss of the Ultimate Heat Sink. This EAL threshold is set to correspond to river conditions that provide adequate early notification of approaching loss of the Ultimate Heat Sink that could jeopardize the level of safety of the plant due to potential loss of Service Water Intake (Ultimate Heat Sink). The National Weather Service can be contacted for further information about existing or projected Adverse Weather Conditions: Mount Holly (609) 261-6604 Mount Holly (609) 261-6602 EAL - 9.9.1.b Rev. 00 Page 1 of 2

HCGS EAL/RALTechnical Basis DEVIATION None REFERENCES NUMARC NESP-007, HUI.7 HC.OP-AB.ZZ-Ul39 (Q), Acts of Nature HCGS Technical Specification Section 3/4, 3.7.3, 3/4.7.1.3, 3/5.7.3 SI.OP-AB.ZZ-0001 (Q) S2.0P-AB.ZZ-0001 (Q) HC Operability Determination 961001148 HCGS UFSAR, Section 2.4, Figure 2.4-3 SGS UFSAR, Section 2.4.11.2, Figure 3.4-1 EAL - 9.9.1.b Rev. 00 Page 2 of 2

HCGS EALIRALTechnical Basis 11.0 Reportable Action Levels

                                                                               -*----::~;.:*v *#-1 11.1 Technical Specifications                                       f REPORTABLE ACTION LEVEL- 11.1.1.a IC      INITIATION OF ANY UNIT SHUTDOWN REQUIRED BY THE TECHNICAL SPECIFICATIONS [10CFR50.72(b)(l)(i)(A)]

RAL Unit shutdown is INITIATED to comply with Technical Specifications OPERATIONAL CONDITION - 1, 2 BASIS Unit shutdown initiated to comply with Technical Specification requires a one hour report in accordance with 10CFRSO.72(b)(1 )(i)(A). This RAL is intended to capture those events for which a Technical Specification required shutdown is initiated Thus, this RAL ensures that the NRC is provided with early warning of safety significant conditions serious enough to warrant a plant shutdown. Unit shutdown INITIATED is defined as the performance of any action(s) to start reducing reactor power to achieve a Hot Shutdown condition. This includes any means, such as, control rod insertion or reducing recirculation flow. A reduction of power for some other purpose, not constituting initiation of a shutdown required by Technical Specifications, is not reportable under this RAL. This 'includes reducing power only for the purpose of repairing a component. For example: The plant has seven days to fix a component or be shut down. If the plant shuts down (not required by TIS yet), the component is fixed, and the plant returns to power prior to the end of the seven day period, it need not be reported IAW I OCFRSO. 72. REFERENCES 10CFR50.72(b )(I )(i)(A) NUREG 1022, Rev. 1, 2nd Draft 0 Page I of I RAL - 11.1.1.a Rev.00

HCGS EAL/RALTechnical Basis 11.0 Reportable Action Levels 11.l Technical Specifications REPORT ABLE ACTION LEVEL - 11.1.1.b IC EXCEEDING ANY TECHNICAL SPECIFICATION SAFETY LIMIT [1 OCFR50.36( c)(l )] RAL Exceeding ANY one of the following Technical Specification Safety Limits:

T/S 2.1.1, THERMAL POWER, Low Pressure or Low Flow
T/S 2.1.2, THERMAL POWER, High Pressure and High Flow
T/S 2.1.3, REACTOR COOLANT SYSTEM PRESSURE o T/S 2.1.4, REACTOR VESSEL WATER LEVEL OPERATIONAL CONDITION - I, 2, 3, 4, 5 (as applicable in T/S)

BASIS This RAL addresses those conditions requiring a one hour report in accordance with IOCFR50.36(c)(l) which states that exceeding a Technical Specification Safety Limit requires a shutdown by Technical Specification. Exceeding a Safety Limit in Technical Specification Section 2.1, in the Operational Condition that the Safety Limit is applicable, shall be reported under this RAL. REFERENCES IOCFR50.36(c)(I) T/S 6.7 0 Page I of I RAL-11.1.1.b Rev. 00

HCGS EALIRALTechnical Basis 11.0 Reportable Action Levels 11.1 Technical Specifications REPORT ABLE ACTION LEVEL - 11.1.1.c IC ANY DEVIATION FROM TIS OR LICENSE CONDITION PURSUANT TO 10CFR50.54(x) [I OCFRSO. 72(b)(I )(i)(B)] RAL Action required because no action consistent with Technical Specifications or license can provide adequate or equivalent protection in an emergency (see NC.NA-AP.ZZ-0005(Q) for guidance on deviation from written procedures) OPERA TI ON AL CONDITION - All BASIS This RAL addresses those conditions that require a one hour report in accordance with IOCFR50.72(b)(l)(i)(B). IOCFR50.54(x) generally permits licensees to take reasonable action in .an emergency even though the action departs from license conditions or plant Technical Specifications if,

1) the action is immediately needed to protect the public health and safety, including site personnel, AND

2) NO action consistent with the license conditions and Technical Specifications is immediately apparent that can provide adequate or equivalent protection.

Such action requires, at a minimum, prior approval by a licensed Senior Reactor Operator who is a member of the Operating Shift. Refer to NC.NA-AP.ZZ-OOOS(Q), Station Operating Practices, for more information concerning the use of I OCFR50.54(x). REFERENCES 10CFR50.54(x) I OCFR50.54(y) I OCFRSO. 72(b)(I )(i)(B) NC.NA-AP.ZZ-OOOS(Q) NUREG 1022, Rev. I, 2nd Draft. Page I of I RAL - 11.1.1.c Rev. 00

HCGS EALIRALTechnical Basis 11.0 Reportable Action Levels 11.1 Technical Specifications REPORTABLE ACTION LEVEL - 11.1.3.a IC VIOLATION OF THE REQUIREMENTS CONTAINED IN THE OPERATING LICENSE [HCGS Operating License, Sections 2.F] RAL Violation of ANY one of the requirements contained in Section 2.C (Items 3 through 13) of the Operating License EXCEPT as otherwise provided in the Technical Specifications or Environmental Protection Plan OPERATIONAL CONDITION - All BASIS This RAL addresses the conditions for a twenty-four hour report in accordance with Item 2.F of the Hope Creek Facility Operating License. REFERENCES HCGS Facility License and Technical SpecificationsD Page I of I RAL - 11.1.3.a Rev. 00

HCGS EAL/RALTechnical Basis 11.0 Reportable Action Levels 11.1 Technical Specifications REPORTABLE ACTION LEVEL - 11.1.3.b IC ANY EVENT REQUIRING AN ENGINEERING EVALUATION BY TECHNICAL SPECIFICATIONS OR COMMITMENT [TIS 3.4.6.1, 3.4.4, 3.7.5] RAL Any of the TIS LCOs for RCS Pressure/ Temperature (TIS 3.4.6.1) are exceeded thereby requiring an Engineering Evaluation. OPERATIONAL CONDITION - All BASIS NOTE: This event may be reportable to the NRC based on other RALs or EALs. Refer to any other RAL or EAL reporting requirements that are applicable and implement those notifications in parallel with initiating an Engineering Evaluation. Conditions reported under this RAL require an Engineering Evaluation of the effects of the condition on plant materials and future operation. This RAL ensures that timely internal notification is initiated to implement the evaluations. REFERENCES HCGS Technical Specification 3.4.6.1 0 Page I of I RAL - 11.1.3.b Rev. 00

HCGS EAL/RALTechnical Basis 11.0 Reportable Action Levels 11.1 Technical Specifications

REPORTABLE ACTION LEVEL- 11.1.3.c IC ANY EVENT REQUIRING AN ENGINEERING EVALUATION BY TECHNICAL SPECIFICATIONS OR COMMITMENT

[f/S 3.4.6.1, 3.4.4, 3.7.5] RAL The conductivity, chloride concentration or pH in the RCS is in excess of its specified limits per T/S 3.4.4, Action Statement C.1 thereby requiring an Engineering Evaluation. OPERATIONAL CONDITION - All BASIS NOTE: This event may be reportable to the NRC based on other RALs or EALs. Refer to any other RAL or EAL reporting requirements that are applicable and implement those notifications in parallel with initiating an Engineering Evaluation. Conditions reported under this RAL require an Engineering Evaluation of the effects of the condition on plant materials and future operation. This RAL ensures that timely internal notification is initiated to implement the evaluations. REFERENCES HCGS Technical Specification 3.4.4 Page I of I RAL - 11.1.3.c Rev. 00

HCGS EAL/RALTechnical Basis 11.0 Reportable Action Levels 11.1 Technical Specifications REPORTABLE ACTION LEVEL- 11.1.3.d IC ANY EVENT REQUIRING AN ENGINEERING EVALUATION BY TECHNICAL SPECIFICATIONS OR COMMITMENT [T/S 3.4.6.1, 3.4.4, 3. 7.5] RAL One or more snubbers are found to be INOPERABLE and have been replaced or restored to an OPERABLE status, an Engineering Evaluation shall be performed per TIS 4.7.5.g OPERATIONAL CONDITION - All BASIS NOTE: This event may be reportable to the NRC based on other RALs or EALs. Refer to any other RAL or EAL reporting requirements that are applicable and implement those notifications in parallel with initiating an Engineering Evaluation. Conditions reported under this RAL require an engineering evaluation of the effects of the condition on plant materials and future operation. This RAL ensures that timely internal notification is initiated to implement the evaluations. REFERENCES HCGS Technical Specification 3.7.5

~*.

Page I of I il RAL - 11.1.3.d Rev. 00

HCGS EAL/RALTechnical Basis 11.0 Reportable Action Levels 11.2 Design Basis I Unanalyzed Condition REPORT ABLE ACTION LEVEL - 11.2.1 IC ANY EVENT OR CONDITION DURING OPERATION THAT RES UL TS IN THE CONDITION OF THE PLANT BEING SERIOUSLY DEGRADED [ 10CFRSO.72(b)( 1)(ii)] RAL As judged by the SNSS/EDO, an event or condition found during plant operations that results in ANY one of the following:

The condition of the plant, including its principal safety barriers, being seriously degraded.
The plant being in an unanalyzed condition that significantly compromises plant safety.
The plant being in a condition outside the design basis of the plant.
The plant being in a condition not covered by normal/abnormal or emergency operating procedures.

OPERATIONAL CONDITION - 1, 2 BASIS Reporting at the component, system, and structure level is required per the above condition. The condition of the plant, including its principal safety barriers, being seriously degraded includes material (e.g., metallurgical or chemical) problems that cause abnormal degradation of the principal safety barriers, (Fuel Clad, RCS, Containment). Examples include:

Fuel clad failure in reactor or spent fuel pool that exceed expected values, are unique or wide spread, are caused by unexpected factors and involve a release of significant quantities of fission products.
Cracks and breaks in RCS piping, reactor vessel or major RCS components.
Significant welding or material defects in the RCS.
Serious temperature or pressure transients.
Loss of relief/safety valve functions.
Loss of containment integrity including excessive containment leakage, loss of containment isolation valve function, loss of containment cooling.

Page I of2 RAL - 11.2.1 Rev. 00

HCGS EAL/RALTechnical Basis The plant being in an unanalyzed condition that significantly compromises plant safety refers to conditions potentially affecting a system, structure or component which are more than of a minor safety significance. It is not intended that this Action level (RAL) apply to minor variation in parameters or to problems concerning single pieces of equipment. The NRC understand that PSE&G will use engineering judgement and experience to determine if an unanalyzed condition exist. If when applying engineering judgement there is doubt as to whether to report or not the NRC recommends that the licensee make the report. The plant being in a condition that is outside design bases would include errors found in the actual design of structures, systems or components which perform safety functions. It would not include minor infractions such as:

Cases of technical inoperability where a component is declared inoperable because a surveillance is overdue.
Case where LCO allowed outage time is slightly exceeded.
Example of conditions that would be reportable under this RAL include:
Discovery that an ECCS design does not meet single failure criteria
Discovery that required high energy line break restraints were not installed.
One train of a safety systems has been incapable of performing its design function for an extended time.

REFERENCES 10CFRSO.72(b)( 1)(ii) NUREG 1022, Rev. 1, 2nd Draft. Page 2 of2 RAL - 11.2.1 Rev. 00

HCGS EAL/RALTechnical Basis 11.0 Reportable Action Levels 11.2 Design Basis I Unanalyzed Condition REPORTABLE ACTION LEVEL- 11.2.2.a IC ANY EVENT FOUND WHILE SHUTDOWN THAT, HAD IT BEEN FOUND WIBLE THE REACTOR WAS IN OPERATION, WOULD HA VE SERIOUSLY DEGRADED THE PLANT OR RESULTED IN BEING IN AN UNANALYZED CONDITION [ 10CFR50.72(b)(2)(i)] RAL Any event, found while the reactor is shutdown, that, had it been found during operation, would have resulted in the plant, including its principal safety barriers being in EITHER one of the following conditions:

seriously degraded
In an unanalyzed condition that significantly compromises plant safety OPERATIONAL CONDITION - 3, 4, 5, Defueled BASIS See RAL 11.2.1 for more information concerning the two plant conditions described in the above RAL.

REFERENCES 10CFR50.72(b )(2)(i) NUREG 1022, Rev. 1, 2nd Draft D Page I of I RAL - 11.2.2.a Rev. 00

HCGS EALIRALTechnical Basis 11.0 Reportable Action Levels 11.2 Design Basis I Unanalyzed Condition REPORT ABLE ACTION LEVEL - 11.2.2.b IC . EVENT/CONDITION THAT ALONE COULD HAVE PREVENTED CERTAIN SAFETY FUNCTIONS [10CFRS0.72 (b)(2) (iii)] RAL Any event or condition that alone could have prevented the fulfillrrient of the safety function of structures or systems that are needed to perform ANY one of the following:

Control the release of radioactive material
Shutdown the reactor and maintain it in a safe shutdown condition
Remove residual heat
Mitigate the consequences of an accident OPERA TI ON AL CONDITION - All BASIS The intent of this RAL is to require reporting of events or conditions that could have prevented safety systems or structures from performing their safety functions (actually or potentially) regardless of when the failure was discovered, whether the system was needed at the time, or whether an alternate system or means was available to perform the safety function.

The phrase "alone could have prevented" means the event or condition was, or would be, sufficient by itself to prevent the performance of the safety function(s) of a system or structure (i.e. no additional single failure is assumed or needed to prevent the function). This. RAL covers an event or condition where structures, components or trains of a Safety System could have failed to perform their intended functions because of:

One or more personnel errors including procedure violations or inadequate maintenance.
Design analysis, fabrication, equipment qualification, construction, or procedural deficiencies.
Equipment failure, if the failure constitutes a condition where there is reasonable doubt that the redundant train or channel is operable.

Note: For systems with 3 or more trains the failure of 2:2 trains should be reported if the functional capability of overall system is/was jeopardized. Page I of2 RAL - 11.2.2.b Rev. 00

HCGS EAL/RALTechnical Basis For a single train safety system, loss of the single train would prevent the fulfillment of the safety function of that system and is therefore reportable even though the plant technical specifications may allow such a condition to exist for a limited time. Individual component failure need not be reported under this RAL if redundant equipment in the same system was operable and available to perform the required safety function. REFERENCES 10CFR50.72 (b)(2) (iii) NUREG 1022, Rev. 1, 2nd Draft Page 2 of2 RAL - 11.2.2.b Rev. 00

HCGS EALIRALTechnical Basis 11.0 Reportable Action Levels 11.2 Design Basis I Unanalyzed Condition REPORT ABLE ACTION LEVEL - 11.2.2.c IC PRESENCE OF A LOOSE PART IN THE REACTOR COOLANT SYSTEM [Reg. Guide 1.133] RAL Presence of a loose part in the RCS is Confirmed OPERATIONAL CONDITION - All BASIS This RAL addresses the conditions requiring a prompt notification with written followup report of operating information in accordance with Regulatory Guides I. 133 and I. 16. Presence of a loose part maybe indicated by an overhead alarm and can be monitored both visually and audibly on the on the Loose Parts Monitor (LPM). The presence of a loose part (i.e., disengaged and drifting) in the primary coolant system can be indication of degraded reactor safety resulting from failure or weakening of a safety restraint component. Loose parts may also come from an item left in the RCS during refueling, or maintenance and can contribute to component damage and material wear by frequently impacting on other parts of the system. In addition, loose parts can pose a serious threat to flow blockage which could lead to localized cladding failure or control rod jamming. Confirmed indicates that an evaluation of a loose parts alarm has determined that the alarm is due to a loose part and not due to detector failure or other plant events. REFERENCES Reg. Guide I. I 6 Reg. Guide 1. I 33 Page I of I RAL - 11.2.2.c Rev. 00

HCGS EALIRALTechnical Basis 11.0 Reportable Action Levels 11.3 Engineered Safety Features (ESF) REPORT ABLE ACTION LEVEL - 11.3 .1 IC ANY EVENT THAT RES ULTS OR SHOULD HAVE RES ULTED IN ECCS DISCHARGE INTO THE RCS AS THE RESULT OF A VALID SIGNAL [ 10CFRSO.72(b)(1 )(iv)] RAL Valid ECCS Actuation, Manual or Automatic, has or should have occurred ECCS Actuation results or should have resulted in discharge to the vessel OPERATIONAL CONDITION - All BASIS NRC experience has shown that events that involve ECCS discharge to the vessel are generally more serious than ESF actuations without discharge to the vessel and thus warrant a one-hour report. Those events that result in either automatic or manual actuation ofECCS or would have resulted in actuation of the ECCS if some component had not failed or an operator action had not been taken are reportable. For example, if a valid ECCS signal was generated by plant conditions and the operator put all ECCS pumps in pull-to-lock position, although no ECCS discharge to the vessel occurred, the event is reportable. A valid signal refers to an intentional manual actuation or actual plant conditions or parameters satisfying the requirements for ECCS initiation. Excluded from this reporting requirement would be those instances in which instruments drift, spurious signals, human error or other invalid signal causes action (e.g. jarring a cabinet, an error in the use of jumpers or lifted leads, error in actuation of controls or switches, or equipment failures). IF the ECCS discharges or should have discharged into the RPV as result of an INVALID signal, THEN a report under this RAL is not required, however RAL 11.3.2 (ESF Actuation) should be reviewed for applicability.

Page I of2 RAL- 11.3.1 Rev. 00

HCGS EAL/RALTechnical Basis REFERENCES NC.NA-AP.ZZ-OOOO(Q), Action Request Process HCGS UFSAR 10CFRSO. 72(b )(1 )(iv) 10CFRS0.73 NUREG 1022, Rev. 1, 2nd Draft Page 2 of2 RAL - 11.3.1 Rev. 00

HCGS EAL/RALTechnical Basis 11.0 Reportable Action Levels 11.3 Engineered Safety Features (ESF) REPORTABLE ACTION LEVEL- 11.3.2 IC ACTUATION OF ENGINEERED SAFETY FEATURE (INCLUDING THE REACTOR PROTECTION SYSTEM) EXCEPT PREPLANNED [IOCFR50.72(b){2)(ii)] RAL Any event or condition that results in manual or automatic actuation of any Engineered Safety Feature (ESF), except as part of a preplanned sequence during reactor operation or testing, including the Reactor Protection System (RPS) ESF/RPS Actuation is determined to be reportable IAW NC.NA-AP.ZZ-OOOO(Q), Action Request Process. OPERATIONAL CONDITION - All BASIS This RAL addresses those conditions requiring a four hour report in accordance with 10CFR50.72(b)(2)(ii). All ESF actuations, including those of the RPS, are reportable regardless of the plant Operational Condition or power level, the significance of the structure, system, or component that initiated the event, or whether initiated manually or automatically. The fact that the safety analysis assumes that an ESF system will actuate automatically under certain plant conditions does not preclude the need to report such actuations. The following exceptions apply: Actuations that result from and are part of the preplanned sequence during testing or reactor operation. This implies that the procedural step indicates the specific ESF RPS actuation that will be generated, and Control Room personnel are aware of the specific signal generation before its occurrence or indication in the Main Control Room. However, if the ESF actuates during the planned operation or test in such a way that it is not part of the planned procedure, such as at a wrong step, that event is reportable. Page I of2 RAL- 11.3.2 Rev. 00

HCGS EALIRALTechnical Basis

2. Invalid actuations that occur when a system has been properly removed from service if all requirements of plant procedures for removing equipment from service have been met.

This would include required clearance documentation, equipment and control board tagging, and properly positioned valves and power supply breakers. NC.NA-AP.ZZ-OOOO(Q}, Action Request Process, Attachment 6, provides additional guidance on the reportability and reporting requirements for such events. REFERENCES NC.NA-AP.ZZ-OOOO(Q}, Action Request Process HCGS UFSAR 10CFRSO.72(b)( 1)(ii) IOCFRS0.73 NUREG 1022, Rev. I, 2nd Draft Page2of2 RAL - 11.3.2 Rev. 00

HCGS EAL/RALTechnical Basis 11.0 Reportable Action Levels 11.4 Personnel Safety I Overexposure REPORT ABLE ACTION LEVEL - 11.4.1 IC ANY INCIDENT OR EVENT INVOLVING BYPRODUCT, SOURCE, OR SPECIAL NUCLEAR MATERIAL CAUSING ANY OF THE LISTED RESULTS [10CFR20.2202(a)] RAL PERSONNEL OVEREXPOSURE or potential for overexposure as indicated by ANY one of the following:

TEDE exposure ~ 25 Rem
LDE exposure ~ 75 Rem
SDE exposure ~ 250 Rem
Release of radioactive material inside or outside of a Restricted Area so that, had an individual been present for 24 hours, the individual could have received ~ 5 times the.

occupational ALI (Annual Limit of Intake) which would usually equate to~ 25 Rem CEDE. This DOES NOT apply to areas where personnel are NOT nonnally stationed during routine operations OPERA TI ON AL CONDITION - All BASIS This RAL addresses those conditions requiring an immediate report IAW 10CFR20.2202(a). Annual Limits on Intake (ALI) are discussed in Appendix B of 10CFR20. Tenns: TEDE = Total Effective Dose Equivalent (integrated dose that consists of the sum of the external dose equivalent (DDE) and committed effective dose equivalent (CEDE). LDE Lens Dose Equivalent (dose equivalent to the eye) SDE Shallow Dose Equivalent (dose equivalent to the skin or extremities) CEDE= Committed Effective Dose Equivalent ALI = Annual Limit of Intake Page I of2 RAL- 11.4.1 Rev.00

HCGS EALIRALTechnical Basis REFERENCES IOCFR20.2202(a) 10CFR20, App. B Page 2 of2 RAL- 11.4.1 Rev. 00

HCGS EAL/RALTechnical Basis 11.0 Reportable Action Levels 11.4 Personnel Safety I Overexposure REPORT ABLE ACTION LEVEL - 11.4.2.a IC ANY INCIDENT OR EVENT INVOLVING LOSS OF CONTROL OF LICENSED MATERIAL CAUSING ANY OF THE LISTED RESULTS [10CFR20.2202(b)] RAL PERSONNEL OVEREXPOSURE or potential for overexposure, as indicated by ANY one of the following:

TEDE exposure > 5 Rem
LDE exposure > 15 Rem
SDE exposure > 50 Rem
Release of radioactive material inside or outside of a Restricted Area so that had an individual been present for 24 hours the individual could have received > 1 times the ~

occupational ALI (Annual Limit oflntake) which would usually equate to> 5 Rem CEDE. This DOES NOT apply to areas where personnel are NOT normally stationed during routine operations. OPERATIONAL CONDITION - ALL BASIS This RAL addresses those conditions requiring a 24 hour report IAW 10CFR20.2202(b). Annual Limits on Intake (ALI) are discussed in Appendix B of 10CFR20. However, because events that result in acute personnel overexposure may result in media interest or notifications to other government agencies, the RAL will result in a 4 hour report IAW 10CFRSO.72(b)(2)(vi). Terms: (The below listed terms are defined in RAL 11.4.l) TEDE = Total Effective Dose Equivalent LDE = Lens Dose Equivalent SDE = Shallow Dose Equivalent CEDE= Committed Effective Dose Equivalent ALI = Annual Limit oflntake Page I of2 RAL - 11.4.2.a Rev. 00

HCGS EAL/RALTechnical Basis REFERENCES 10CFR20.2202(b) 10CFR20, App. B 10CFR50. 72(b)(2)(vi) Page 2 of2 RAL - 11.4.2.a Rev. 00

HCGS EAL/RALTechnical Basis 11.0 Reportable Action Levels 11.4 Persminel Safety I Overexposure REPORTABLE ACTION LEVEL- 11.4.2.b IC ONSITE FATALITY [10CFR50.72(b)(2)(vi)] RAL Any fatality has occurred within the Owner Controlled Area (OCA) OPERATIONAL CONDITION - All BASIS The above condition is reportable because an "Onsite" fatality will most likely involve notification of other government agencies and may involve the media. Other government agencies and the media often rely on the NRC for an independent explanation of the safety implication of events at nuclear power plants; therefore, timely NRC notification is required. In this RAL, the normal definition of ONSITE which pertains to the PROTECTED AREA is expanded to include the entire OWNER CONTROLLED AREA (OCA) due to anticipated media interest in any fatality of an individual working at the site (i.e., Artificial Island). REFERENCES 10CFR50.72(b)(2)(vi) NUREG 1022, Rev. 1, 2nd Draft Page I of I RAL - 11.4.2.b Rev. 00

HCGS EALIRALTechnical Basis 11.0 Reportable Action Levels 11.4 Personnel Safety I Overexposure REPORTABLE ACTION LEVEL - 11.4.2.c IC RADIOACTIVELY CONT AMINA TED PERSON TRANSPORTED FROM THE SITE TO AN OFFSITE MEDICAL FACILITY FOR TREATMENT [10CFR50.72(b)(2)(v)] RAL Transportation of a radioactively contaminated or potentially contaminated individual from the site to an offsite medical facility for treatment. OPERATIONAL CONDITION - All BASIS This RAL addresses those conditions requiring a four hour report in accordance with I OCFRSO. 72(b )(2)(v). Transportation of a radioactively contaminated individual to an offsite medical facility has the potential for spreading the cor.tamination to individuals and institutions that are not trained or prepared to deal with radioactive materials. The NRC requires notification of any event with the potential to contaminate Unrestricted Areas in the public domain. A potentially contaminated individual means a person who, due to injuries or first aid treatments cannot be adequately surveyed for contamination prior to transport to an offsite medical facility. REFERENCES 10CFRSO.72(b)(2)(v) NUREG I 022, Rev. I, 2nd Draft Page I of I RAL - 11.4.2.c Rev. 00

HCGS EAL/RALTechnical Basis 11.0 Reportable Action Levels 11.4 Personnel Safety I Overexposure REPORTABLE ACTION LEVEL- 11.4.3.a IC SIGNIFICANT FITNESS FOR DUTY EVENTS [IOCFR26. 73] RAL Any event that is determined to be reportable by the Medical Review Officer (MRO) or designee IAW PSE&G's Fitness for Duty Program (NC.NA-AP.ZZ-0042(Q)) The reportable details of the event are made available to the SNSS by the l\1RO or designee. OPERATIONAL CONDITION - All BASIS NC.NA-AP.ZZ-0042(Q) provides the guidance to determine reportability of Significant Fitness for Duty events which requires a 24 hour report IAW fOCFR26. 73. Only the Medical Review Officer or designee may determine reportability of these events for PSE&G, unless the event has safeguards significance, in which case the determination to report is made by Security. REFERENCES NC.NA-AP .ZZ-0042(Q) IOCFR26.73 Page I of I RAL - 11.4.3.a Rev. 00

HCGS EAL/RALTechnical Basis 11.0 Reportable Action Levels 11.4 Personnel Safety I Overexposure REPORTABLE ACTION LEVEL- I L4.3.b IC FITNESS FOR DUTY PROGRAM: FALSE POSITIVE DUE TO ADMINISTRATIVE ERROR (BLIND TEST BY LAB) [10CFR26, APP. A, 2.8(e)(5)] RAL The occurrence of a false positive error on a blind lab performance test specimen under I OCFR26 as determined by the Medical Review Officer (MRO) IAW PSE&G's Fitness for Duty Program (NC.NA-AP.ZZ-0042(Q)) The reportable details of the event are made available to the SNSS by the :MRO or designee. OPERATIONAL CONDITION - All BASIS NC.NA-AP .ZZ-0042(Q) provides the guidance to determine reportability of administrative errors occurring in the lab testing program which requires a 24 hour report IAW I OCFR26. Blind Quality Control proficiency monitoring ofDIIBS LABS are performed on a regular basis. Any occurrence of a false positive error which, after investigation by the MRO, is determined to be the result of an administrative error (clerical, sample mix-up, etc.) is reportable to the NRC. Only the Medical Review Officer or designee may determine reportability of these events for PSE&G. REFERENCES NC.NA-AP.ZZ-0042(Q) 10CFR26, Appendix A 2.8(e)(5) Page I of I RAL - 11.4.3.b Rev. 00

HCGS EALIRALTechnical Basis 11.0 Reportable Action Levels 11.5 Environmental REPORTABLE ACTION LEVEL - 11.5.2.a IC SPILL/DISCHARGE OF ANY NON-RADIOACTIVE HAZARDOUS SUBSTANCE [10CFR50.72(b)(2)(vi); N.J.A.C. 7:1E] RAL Spill/discharge of an industrial chemical or petroleum product outside of a plant structure within the Owner Controlled Area (OCA) that results in EITHER one of the following:

Spill I discharge that has passed through the engineered fill and into the ground water as confirmed by Licensing
Spill I discharge that CANNOT be cleaned up within 1 hour and no contact with groundwater is suspected NOTE:

This event MAY require IMMEDIATE (15 minute) notifications. DO NOT delay implementation of Attachment 16. OPERATIONAL CONDITION - All BASIS This RAL addresses the conditions requiring reports IAW PSE&G's DPCC/DCR Plan. The intent of this RAL is to direct IMMEDIATE implementation of ECG Attachment 16, which will provide further direction on reportability based upon the nature of the Spill/Discharge as well as the expertise of Environmental Licensing personnel concerning requirements. REFERENCES 10CFR50. 72(b )(2)(vi) N.J.A.C. 7:1E DPCC/DCR Plan, Part III Page I of I RAL - 11.5.2.a Rev. 00

HCGS EAL/RALTechnical Basis 11.0 Reportable Action Levels 11.5 Environmental REPORTABLE ACTION LEVEL - 11.5.2.b IC SPILL/DISCHARGE OF ANY NON-RADIOACTIVE HAZARDOUS SUBSTANCE INTO OR UPON THE RIVER [10CFR50. 72(b)(2) (vi); N.J.A.C. 7: lE] RAL EITHER one of the following events occur:

Observation of a spill/discharge of an industrial chemical or pt!troleum product from on-site into the Delaware River or into a storm drain
Observation of an oil slick on the Delaware River from any source.

NOTE: This event MAY require IMMEDIATE (15 minute) notifications. DO NOT delay implementation of Attachment 16. OPERATIONAL CONDITION - All BASIS Thi.s RAL addresses the conditions requiring reports IAW PSE&G's DPCC/DCR Plan. The intent of this RAL is to direct IMMEDIATE implementation of ECG Attachment 16, which will provide further direction on reportability based upon the nature of the Spill/Discharge as well as the expertise of Environmental Licensing personnel concerning requirements. REFERENCES 10CFR50.72(b)(2) (vi) N.J.A.C.7:1E DPCC/DCR Plan, Part III Page I of I RAL - 11.5.2.b Rev. 00

HCGS EALIRALTechnical Basis 11.0 Reportable Action Levels 11.5 Environmental REPORTABLE ACTION LEVEL - 11.5.2.c IC UNUSUAL OR IMPORT ANT ENVIRONMENT AL EVENTS [ENVIRONMENTAL PROTECTION PLAN SECTION 4.1] RAL As judged by the SNSS/EDO ANY one of the following events has occurred:

Unusually large fish kill
Protected aquatic species impinge on Circulating or Service Water intake screens (e.g.,

sea turtle, sturgeon) as reported by Site personnel

Any occurrence of an unusual or important event that indicates or could result in significant environmental impact casually related to plant operation; such as the following:
Onsite plant or animal disease outbreaks
Mortality or unusual occurrence of any species protected by the Endangered Species Act of 1973
Increase in nuisance organisms or conditions
Excessive bird impactation
NJPDES Permit violations

     *Excessive opacity (smoke)

OPERATIONAL CONDITION - All BASIS This RAL addresses those conditions requiring reports in accordance with the Environmental Protection Plan. Final determination or reportability will be made by Environmental Licensing as a result of implementing Attachment 15. REFERENCES HCGS Technical Specifications, ENVIRONMENT AL PROTECTION PLAN Page l of l RAL - 11.5.2.d Rev. 00

HCGS EALIRALTechnical Basis 11.0 Reportable Action Levels 11.6 After-the-Fact REPORTABLE ACTIONLEVEL-11.6.1 IC EMERGENCY CONDITIONS DISCOVERED AFTER-THE-FACT RAL Discovery of events or conditions that had previously occurred (event was NOT ongoing at the time of discovery) which EXCEEDED an Emergency Action Level (EAL) and was NOT declared as an emergency There are currently NO adverse consequences in progress as a result of the event OPERATIONAL CONDITION - All BASIS In the event a condition is discovered to have previously occurred or existed that exceeded an Emergency Action Level threshold, but that no emergency was declared and the basis for the Emergency Classification no longer exists at the time of discovery, then a one hour report is required. This situation might arise due to a condition existing without detection by operating personnel. The NRC does not consider actual declaration of the emergency classification to be necessary in these circumstances. REFERENCES Hope Creek ECG Introduction Section NUREG 1022, Rev. 1, 2nd Draft, Pg. 20 Page I of I RAL- 11.6.1 Rev. 00

HCGS EAL/RALTechnical Basis 11.0 Reportable Action Levels

11. 7 Security I Emergency Response Capabilities REPORTABLE ACTION LEVEL- 11.7.1.a IC SAFEGUARDS EVENTS THAT ARE DETERMINED TO BE NON-EMERGENCIES, BUT ARE REPORTABLE TO THE NRC WITHIN ONE HOUR

[10CFR73. 71 (b )(1 )] RAL Any Non-Emergency safeguards event that is reportable in accordance with 10CFR73. 71 as determined by Security (SCP-15) OPERATIONAL CONDITION - All BASIS This RAL addresses those conditions requiring a one hour report in accordance with 10CFR73. 71(b)(l). These non-emergency events are outlined in Security Contingency Procedure #15. The on-duty PSE&G Security Supervisor should provide information concerning the specific event. REFERENCES 10CFR73. 71 (b )(1) SCP-15 Page I of I RAL- 11.7.1.a Rev. 00

HCGS EAL/RALTechnical Basis 11.0 Reportable Action Levels 11.7 Security I Emergency Response Capabilities REPORTABLE ACTION LEVEL- 11.7.1.b IC MAJOR LOSS. OF EMERGENCY ASSESSMENT CAP ABILITY, OFFSITE RESPONSE CAPABILITY, OR COMMUNICATIONS CAPABILITY [IOCFRSO. 72(b )(I )(v)] RAL SNSS/EC determines that an event(s) (excluding a scheduled test or preplanned maintenance activity) has occurred that would impair the ability to deal with an accident or emergency as indicated by the Loss of ANY one of the following:

Nuclear Emergency Telecommunications System (NETS) for> 1 hr
ENS for> 1 hr in the Control Room, TSC, or EOF (NIA if reported by the NRC)
More than seven Offsite Sirens for> 1 hr
Use of the EOF for> 8 hrs
All Meteorological data (Hope Creek AND Salem) for> 8 hrs
Site access due to Acts of Nature (snow, flood, etc.)

OPERATIONAL CONDITION - All BASIS NOTE: IF losses are part of a scheduled test or preplanned maintenance activity AND WHEN compensatory actions have been taken, THEN NO report is required. This RAL addresses conditions that are COMMON to both Hope Creek and Salem and may be reported to the NRC by EITHER station as a Common Site Event. I. Loss of the NETS or ENS for > l hour directly affects the ability to promptly notify and communicate with the NRC and/or offsite officials. IF a total loss of communications capabilities has occurred, THEN REFER to ECG Section 8.2. IF notified by the NRC Operations Officer of an inoperable ENS line, THEN NO further notification is necessary. Page I of2 RAL - 11.7.1.b Rev. 00

HCGS EALIRALTechnical Basis

2. Loss of Offsite Sirens (>10%) represents a loss of ability to promptly notify a large portion of the population, and warrants an immediate notification. There are 71 offsite sirens in the Plume EPZ and therefore a loss of~ 8 is a > 10% loss which represents a loss of Offsite Response Capability.

3. Use of the EOF may be vital in responding to an emergency. Loss of use of this facility or its supporting equipment, or ability to staff represents a significant loss of emergency response capability.

4. Loss of meteorological data for an extended period of time limits the ability to predict radiological conditions during an emergency situation. An extended loss warrants notification of the loss of this capability.

5. Limited site access may affect the ability to staff the site personnel and/or emergency response facilities, and the ability of off-site agencies to implement emergency plan requirements.

WHEN site reaction to anticipated conditions is commenced, THEN notification should be made, if possible. REFERENCES 10CFR50.72(b)( 1)(v) NUREG-1022, Rev. 1, 2nd Draft Page 2 of2 RAL-11.7.1.b Rev. 00 L

HCGS EAL/RALTechnical Basis 11.0 Reportable Action Levels

11. 7 Security I Emergency Response Capabilities REPORTABLE ACTION LEVEL- 11.7.l.c IC MAJOR LOSS OF EMERGENCY ASSESSMENT CAPABILITY, OFFSITE RESPONSE CAPABILITY, OR COMMUNICATIONS CAPABILITY

[10CFR50. 72(b)(l )(v)] RAL SNSS/EC determines that an event(s) (excluding a scheduled test or preplanned maintenance activity) has occurred that would impair the ability to deal with an accident or emergency as indicated hy the Loss of ANY one of the following:

Use of the TSC for > 8 hrs
SPY, NPV, or FRVS vent effluent radiation monitors for > 8 hrs
SPDS OR CRIDS for > 8 hrs
EROS, NRC phone line, Modem for >I hr (NIA if reported hy thecNRC)
More than 75 % of the OH As for < 15 minutes
Concurrent multiple accident or emergency condition indicators which in the judgment of the SNSS significantly impairs assessment capabilities OPERATIONAL CONDITION - All BASIS NOTE: IF losses are part of a scheduled test or preplanned maintenance activity AND WHEN compensatory actions have been taken, THEN NO report is required.

1. Use of the TSC may be vital in responding to an emergency. Loss of use of this facility, or its supporting equipment, or ability to staff represents a significant loss of emergency response capability.

2. Loss of ALL effluent radiation monitors on ANY one of the plant vents for an extended period of time limits the ability to predict radiological conditions during an emergency situation. An extended loss warrants notification of the loss of this capability.

Page I of 2 RAL - 11.7.1.c Rev. 00

HCGS EAL/RALTechnical Basis

3. Loss of SPDS or CRIDS for > 8 hours is considered an event that significantly impairs safety assessments capabilities.

4. Loss of ERDS, NRC phone line, Modem need to be reported to the NRC, so the NRC can have them repaired.

5. Loss of OHAs for a short period of time ( < 15 minutes) is considered a loss of emergency assessment capability.

IF OHAs are lost or were lost for 2. 15 minutes, THEN REFER to ECG Section 8.2.

6. Concurrent multiple accident or emergency condition indicators which in the judgment of the SNSS significantly impairs assessment capabilities is specific to Hope Creek in this RAL.

IF the loss of assessment capability is COMMON to both Hope Creek and Salem, THEN REFER to RAL 11.7.1.b. REFERENCES 10CFR50. 72(b )(I )(v) NUREG-1022, Rev. 1, 2nd Draft Page 2 of2 RAL- 11.7.1.c Rev. 00

HCGS EALIRALTechnical Basis 11.0 Reportable Action Levels 11.8 Public Interest REPORT ABLE ACTION LEVEL - 11.8.2.a IC UNUSUAL CONDITIONS WARRANTING A NEWS RELEASE OR NOTIFICATION OF GOVERNMENT AGENCIES [IOCFR50.72(b)(2)(vi)] RAL SNSS/EDO judges that an event or situation has occurred that is related to ANY one of the following:

The health and safety of the public
The health and safety of onsite personnel
Protection of the environment EITHER one of the following:
A news release is planned
Notification to a Local, State or Federal agency has been or will be made OPERATIONAL CONDITION - All BASIS Events that require the NRC to respond due to media or public interest, or other government agency involvement are reportable to the NRC. Examples of the events would include, but not be limited to:
release of contaminated tools or equipment to public areas
non-routine releases of radioactive effluents .
inadvertent operation of the offsite siren system
state agency contacted due to fish kill
toxic material release from the site PSE&G generally does not have to report media and government interaction or notify the NRC of every press release issued unless they are related to, or are perceived by the public or media to be related to, the radiological health and safety of the public or onsite personnel, or protection of the environment.

Page I of2 RAL - 11.8.2.a Rev. 00

HCGS EALIRALTechnical Basis REFERENCF.S 10CFR50. 72(b)(2)(vi) NUREG 1022, Rev. 1, 2nd Draft Page 2 of2 RAL - 11.8.2.a Rev. 00

HCGS EAL/RALTechnical Basis 11.0 Reportable Action Levels 11.8 Public Interest REPORTABLE ACTION LE'vEL- 11.8.2.b IC UNUSUAL CONDITIONS DIRECTLY AFFECTING LOWER ALLOWAYS CREEK TOWNSHIP (LACI) [LAC - M.O.U.] RAL As judged by the SNSS/EDO, events which are the responsibility of PSE&G which have or may result in EITHER one of the following:

Anticipated unusual movement of equipment or personnel which may significantly affect local traffic patterns
Onsite events which involve alarms, sirens or other noise which may be heard off-site OPERATIONAL CONDITION - All BASIS This RAL addresses conditions that are otherwise not reportable to the NRC, but are considered to warrant a prompt report IAW the Lower Alloways Creek Township Memorandum of Understanding (M.O.U.) with PSE&G because they are oflocal interest only.

IF an NRC report is required by any other EAL or RAL, TIIEN REFER to that section of the ECG for action required which will ensure that LAC Township is notified appropriately. PSE&G shall notify LAC Township as soon as sufficient details are available, but in no case should this time frame exceed twelve hours. Sufficient details are those needed to convey a general understanding of the condition or event to a lay public. Four hours is specified in this RAL (rather than the twelve allowed) as a reasonable time period for taking the actions required and well within the agreed time frame of the M.O.U. REFERENCES LAC-M.O.U. Page I of I RAL - 11.8.2.b Rev. 00

HCGS EAL/RALTechnical Basis 11.0 Reportable Action Levels 11.9 Accidental Criticality I Special Nuclear Material I Rad Material Shipments - Releases REPORTABLE ACTION LEVEL - 11.9.1.a IC UNPLANNED I ACCIDENTAL CRITICALITY [10CFR70.52(a)] RAL Any unplanned or accidental criticality OPERATIONAL CONDITION - All i BASIS This RAL is intended to provide immediate notification to the NRC for events which constitute a "loss" of Reactivity Control due to errors in calculations, or mis-operation. This condition can be detected from the Control Room using available Nuclear Instrumentation by observation of a sustained positive period on the Source Range Monitors (SRMs). Increases in neutron population due to subcritical multiplication can be expected during Core Alterations and should not be classified using this RAL. REFERENCES 10CFR70.52(a) Page I of I RAL - 11.9.1.a Rev. 00

HCGS EAL/RALTechnical Basis 11.0 Reportable Action Levels 11.9 Accidental Criticality I Special Nuclear Material I Rad Material Shipments - Releases REPORT ABLE ACTION LEVEL - 11. 9 .1.b IC LOSS AND INVESTIGATION OF THE LOSS OF SPECIAL NUCLEAR MATERIALS/ SPENT FUEL [10CFR73.27(c), 10CFR73.7l(a)] RAL ANY one of the following events occur involving Special Nuclear Material (SNM) or Spent Fuel:

Shipment of formula quantities of strategic SNM (SSNM) or Spent Fuel that is lost or unaccounted for after the estimated time of arrival
A lost or unaccounted for shipment of SSNM or Spent Fuel has been recovered or accounted for
Results of a trace investigation of lost or unaccounted for SSNM shipment are received OPERATIONAL CONDITION - All BASIS This RAL addresses those conditions requiring a one hour report IAW 10CFR73.27(c) or 10CFR73.7l(a).

Strategic Special Nuclear Material (SSNM) means uranium-235 (contained in uranium enriched to 20 percent or more in uranium-235 isotope), uranium-233, or plutonium. Formula quantity means 5000 grams SSNM in any combination, computed by the formula, grams SSNM =(grams contained U-235) + 2.5 (grams U-233 +grams plutonium) 10CFR73.7l(a)(l) requires a one hour report of a shipment loss, and on recovery of a lost shipment. IO CFR 73.27(c) requires an immediate trace investigation oflost or unaccounted for shipments and reporting in accordance with I OCFR 73. 71. Page I of2 RAL - 11.9.1.b Rev. 00

HCGS EAL/RALTechnical Basis REFERENCES 10CFR73.27(c) 10CFR73.71(a) Page 2 of2 RAL - 11.9.1.b Rev. 00

HCGS EAL/RALTechnical Basis 11.0 Reportable Action Levels 11.9 Accidental Criticality I Special Nuclear Material I Rad Material Shipments - Releases REPORTABLE ACTION LEVEL - 11.9.1.c IC THEFT OR LOSS OF LICENSED MATERJAL [IOCFR20.220l(a)(I)(i)] RAL Lost, stolen or missing licensed material ~ 1000 times the quantity specified in 10CFR20 Appendix C, in such circumstances that it appears that an exposure could result to persons in Unrestricted Areas. OPERATIONAL CONDITION - All BASIS This RAL addresses those conditions requiring an immediate report IAW 10CFR20.220 I (a)( I )(i). Licensed material means source material, special nuclear material (SNM), or by-product material received, possessed, used, or transferred under a general or specific license issued by the NRC pursuant to the regulations in I OCFR20. Unrestricted Areas are any areas beyond the Minimum Exclusion Area (MEA). (outside the Owner Controlled Area (OCA) boundary) REFERENCES 10CFR20.220 l (a)( l )(i) Page I of I RAL-11.9.1.c Rev. 00

HCGS EAL/RALTechnical Basis 11.0 Reportable Action Levels 11.9 Accidental Criticality I Special Nuclear Material I Rad Material Shipments - Releases REPORTABLE ACTION LEVEL- 11.9.1.d IC RECEIPT OF SSNM MATERIAL [10CFR73.27(b)] RAL Receipt of shipment of Strategic Special Nuclear Material (SSNM) OPERATIONAL CONDITION - All BASIS This RAL addresses, in part, those conditions requiring an immediate report in accordance with 10CFR73.27(b). Strategic Special Nuclear Material (SSNM) is uranium 235 (contained in uranium enriched to 20% or more in the U-235 isotope), U-233 or plutonium. REFERENCES 10CFR73.27(b) 10CFR70.4 Page I of I RAL - 11.9.1.d Rev. 00

HCGS EAL/RALTechnical Basis 11.0 Reportable Action Levels 11.9 Accidental Criticality I Special Nuclear Material I Rad Material Shipments - Releases REPORTABLE ACTION LEVEL- 11.9.1.e IC EXCESSIVE CONTAMINATION ANDI OR RADIATION LEVELS ON A PACKAGE [IOCFR20. l 906(d)] RAL Receipt survey indicates that package contamination I radiation levels equal or exceeds ANY one of the followinf

2200 dpm/100 cm
200 mR/hr on contact
10 mR/hr at 3 feet OPERA TI ONAL CONDITION - All BASIS This RAL addresses those conditions requiring an immediate report IAW I OCFR20. l 906(d).

This requirement refers to values provided in IOCFR71.87(i)(l) for contamination and to IOCFR71.47 for radiation levels. The RAL contamination level is based on the limit, adjusted for the standard swipe area used at Hope Creek. 10CFR71.87(i)(2) allows contamination levels of 10 times the above limits for Exclusive Use Shipments. Exclusive Use means the sole use of a conveyance by a single consignor and for which loading and unloading are carried out with the direction of the consignor or consignee. REFERENCES I OCFR20. l 906(d) 10CFR71.4 IOCFR71.47 I OCFR71.87(i)(l )/(2) Page I of I RAL - 11.9. l.e Rev. 00

HCGS EAL/RALTechnical Basis 11.0 Reportable Action Levels 11.9 Accidental Criticality I Special Nuclear Material I Rad Material Shipments - Releases REPORTABLE ACTION LEVEL- 11.9.2.a IC ACCIDENT OCCURRING DURING TRANSPORTATION OF LICENSED MATERIAL [IOCFR71.5(a)(l)(v)] RAL Accidents during the transportation of radioactive material which are reported to PSE&G as the shipper that involve (or potentially involve) damage to the cargo OPERATIONAL CONDITION - All BASIS IOCFR71.5(a)(l)(v) refers to 49CFR171.15/16 for transportation of licensed material accident reporting. Note: Vehicle breakdowns or delays enroute may also be reported by the driver, but are not reportable to the NRC unless an accident is involved (cargo damage). Radioactive Material means any item, gas, liquid, flowable solid, or material with radioactivity levels in excess of the limits for unconditional release found in Section 5.12.1. ofNC.NA-AP.ZZ-0024(Q), Radiation Protection Program. REFERENCES 10CFR71.5(a)(1 )(v) 49CFR171.15/16 NC.NA-AP.ZZ-0024(Q) Page I of I RAL - 11.9.2.a Rev.00

HCGS EAL/RALTechnical Basis 11.0 Reportable Action Levels 11.9 Accidental Criticality I Special Nuclear Material I Rad Material Shipments - Re~eases REPORTABLE ACTION LEVEL- 11.9.2.b IC CONTAMINATION OUTSIDE OF THE RADIOLOGICALLY CONTROLLED AREA [ 10CFRSO.72(b)(2)(vi)] RAL Discovery of a Contaminated Area OUTSIDE of the RCA with removable activity EITHER one of the following:

Location of Contaminated Area is such that a contaminated person or material may have left the Protected Area
Location of Contaminated Area is OUTSIDE of Plant Structures AND Size of Contaminated Area is LARGE (>I 00 ft 2)

OPERATIONAL CONDITION - All BASIS The purpose of the RAL is to ensure that the NRC is made aware of issues that may cause heightened public or government concern related to the radiological health and safety of the public or onsite personnel or protection of the environment. These RAL contamination levels are based on the likelihood that a news release and/or notifications to government agencies may need to be made for these conditions. Examples of conditions that would require classification under this RAL would include:

Release of contaminated tools, equipment, trash, vehicles, personnel to areas outside the Protected Area.
Unusual or abnormal release of radioactive effiuents. Unusual or abnormal can be considered a release that has the potential to generate public, media, or other government agency attention.

Radiological effluent releases that are >2 times Technical Specifications limits are classified in accordance with ECG Section 6. Page I of2 RAL - 11.9.2.b Rev. 00

HCGS EAL/RALTechnical Basis REFERENCES Commitment #: EP95-002 10CFR50.72(b)(2)(vi) NUREG 1022, Rev. 1, 2nd Draft Page 2 of2 it* RAL - 11.9.2.b Rev. 00

HCGS EAL/RALTechnical Basis 11.0 Reportable Action Levels 11.10 Voluntary Notifications REPORT ABLE ACTION LEVEL - 11.10.2 IC EVENTS/CONDITIONS WARRANT VOLUNTARY/COURTESY NRC NOTIFICATIONS [10CFRS0.72 - VOLUNTARY REPORT] RAL In the judgment of the SNSS, notification to the NRC is warranted NO other EALs or RALs appear to be applicable OPERATIONAL CONDITION - All BASIS Hope Creek may make voluntary or courtesy Emergency Notification System (ENS) notifications about events or conditions the NRC may be interested in. This is true when it is unique to our facility, but especially when it appears to have generic implications. The NRC responds to any voluntary notification of an event or condition as its safety significance warrants, regardless of our classification of the reporting requirement. IF it is determined later that the event IS reportable, THEN the SNSS can change the ENS notification to a required notification under the appropriate 10 CFR 50.72 reporting criterion. Hope Creek may continue with plant operation provided there is a reasonable expectation that the equipment in question is OPERABLE. WHEN this reasonable expectation no longer exists, OR significant doubts begin to arise, THEN the equipment should be considered INOPERABLE and appropriate actions, including required reporting, should be taken. In some cases, such as discovery of an existing, but previously unrecognized condition, it may be necessary to undertake an evaluation in order to determine if an event or condition is reportable. If so, the guidance provided in Generic Letter 91-18, which applies primarily to operability determinations, is appropriate for reportability determinations as well. This guidance indicates that Page I of2 RAL - 11.10.2 Rev. 00

HCGS EALIRALTechnical Basis an evaluation should generally proceed on a schedule commensurate with the safety significance of the question. REFERENCES Commitment #: EP95-001 Hope Creek ECG Introduction Section NUREG 1022, Rev. 1, 2nd Draft NRC Generic Letter 91-18 Page 2 of2 RAL - 11.10.2 Rev. 00

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ECG T.O.C. Pg. l of 4 l CC:;Tf-)QL CQPV #HOP~ CREEK EVENT CLASSIFICATION GUIDE I. T,t.BLE OF CONTENTS/SIGNATURE PAGE I ~ Ct.ZJ_,, l j January 21, 1997 I

PAGES DATE

~'MUN_,_                 --~LE REV#

T.O.C. Table of Contents/Signature Page 00 4 Ol/21197 Introduction and Usage 00 9 Ol/2 l/97 11 Glossary of Acronyms & Abbreviations 00 5 0 l/2 l/97 - 1.0 Fuel Clad Challenge 00 Ol/21/97 2.0 RCS Challenge 00 01/21197 3.0 Fission Product Barriers (Table) 00 Ol/21197 4.0 EC Discretion 00 01/2 l/97 5.0 Failure to SCRAM 00 l 01/21/97 6.0 Radiological Releases/Occurrences 6.1 Gaseous Effluent Release 00 4 Ol/21197 6.2 Liquid Effluent Release 00 l 01/21197 6.3 In - Plant Radiation Occurrences 00 1 Ol/21/97 6.4 Irradiated Fuel Event 00 2 Ol/21197 7.0 Electrical Power 7.1 Loss of AC Power Capabilities 00 2 0 l/21/97 7.2 Loss of DC Power Capabilities 00 1 Ol/21197 8.0 System Malfunctions 8.1 Loss of Heat Removal Capability 00 01/21/97 8.2 Loss of Overhead Annunciators 00 01/21/97 8.3 Loss of Communications Capability 00 Ol/21197 8.4 Loss of Control Room Habitability 00 1 0 l/2 l/97 8.5 Technical Specifications 00 l Ol/21197 9.0 Hazards - Internal/External 9.1 Security Threats 00 01/21/97 9.2 Fire 00 1 01/21197 9.3 Explosion 00 1 0 l/2 l/97 9.4 Toxic/Flammable Gases 00 2 01/21/97 9.5 Seismic Event 00 01/21/97 9.6 High Winds 00 01/21/97 9.7 Flooding 00 1 0 l/21/97 9.8 Turbine FailureNehicle Crash/ 00 1 01/21/97 Missile Impact 9.9 River Level 00 01121/97 HCGS Rev. 00

ECG T.O.C. Pg. 2of4 HOPE CREEK EVENT CLASSIFICATION GUIDE TABLE OF CONTENTS/SIGNATURE PAGE January 21, 1997 SECTION TITLE REV# PAGES DATE 10.0 Reserved for future use 11.0 Reportable Action Levels (RALs) 11.1 Technical Specifications 00 2 01/21/97 11.2 Design Basis/ Unanalyzed Condition 00 2 01/21197 11.3 Engineered Safety Features (ESF) 00 1 01/21197 11.4 Personnel Safety/Overexposure 00 2 01121/97 11.5 Environmental 00 01/21/97 11.6 After-the-Fact 00 01/21/97

11. 7 Security/Emergency Response 00 01/21/97 Capabilities 11.8 Public Interest 00 1 01/21/97

11. 9 Accidental Criticality/ 00 2 01/21/97 Special Nuclear Material I Rad Material Shipments - Releases 11.10 Voluntary Notifications 00 1 01/21/97 WC Hope Creek ECG Charts (Located in ERFs) 00 2 01/21/97 HCGS Rev.00

ECG T.O.C. Pg. 3of4 HOPE CREEK EVENT CLASSIFICATION GUIDE TABLE OF CONTENTS/SIGNATURE PAGE January 21, 1997 ATTACHMENT TITLE REV# PAGES DATE 1 UNUSUAL EVENT 00 9 01/21/97 2 ALERT 00 4 01/21/97 3 SITE AREA EMERGENCY 00 5 01/21/97 4 GENERAL EMERGENCY 00 7 01/21/97 5 NRC Data Sheet Completion Reference 00 7 01/21/97 6 Primary Communicator Log 00 8 01/21/97 7 Primary Communicator Log (GE) 00 7 01/21/97 8 Secondary Communicator Log 00 8 01/21/97 9 Non-Emergency Notifications Reference 00 3 01/21/97 10 l Hr Report - NRC Regional Office 00 3 01/21/97 11 1 Hr Report (Common Site) Security Safeguards 00 3 01/21/97 12 l Hr Report - NRC Operations 00 3 01/21/97 13 4 Hr Report - Contaminated Events 00 7 01/21/97 Outside Of The RCA 14 4 Hr Report - NRC Operations 00 3 01/21/97 15 Environmental Protection Plan 00 3 01/21/97 16 Spill I Discharge Reporting 00 7 01/21/97 17 4 Hr Report - Fatality or Medical 00 4 01/21/97 Emergency 18 4 Hr Report - Radiological 00 4 01/21197 Transportation Accident 19 24 Hr Report - Fitness For Duty (FFD) 00 3 01/21/97 Program Events 20 24 Hour Report - NRC Regional Office 00 3 01121197 21 Reportable Event LAC/Memorandum 00 2 01/21/97 Of Understanding (M.O.U.) 22 TIS Required Engineering Evaluation 00 2 01121197 23 Reserved 24 UNUSUAL EVENT (Common Site) 00 10 01121/97 25 1 Hr Report (Common Site) Major Loss Of 00 3 01/21/97 Emergency Assessment, Offsite Response OR Communications Capability HCGS Rev.00

ECG T.O.C. Pg. 4of4 SIGNATURE PAGE J-?-11 Prepared By: / 13 y7711t-Date Section/Attachments Revised: _ _,_fl~L-_._L____________ 0(-)1-97 (List Non-Editorial Only - Section/Attachments) Date

Reviewed By: -~~-:.;;..;*;;~#&,..c........~o:.....=.......----

Station Qualified Reviewer ReviewedBy: _ _ _ _ __::::*=~~~.c:;;t:::::::=========------~

                        ~anm;,nManager 1-J r--72 I

ergency Preparedness Manager Date ReviewedBy: _ _ _ _ _ _ _ _ ___.fj_;_~(+/--=---------- Director - QA/Nuclear Safety Review Date (If Applicable) SORC Review and Station Approvals Mtg. No. Hope Creek Chairman 1- 1-s-- Date rr Effective Date of this Revision: 01/21/97 Date HCGS Rev. 00

ECG Section i Pg 1of10 HOPE CREEK EVENT CLASSIFICATION GUIDE INTRODUCTION & USAGE Section i I. PltRPOSE.0.F.. HE EVENT CLASSIFICATION GUIDE ECG A To provide a central reference document which enables the Senior Nuclear Shift Supervisor (SNSS) or the Emergency Coordinator (EC) to classify emergency or non-emergency events and conditions. B. To provide the required procedures for immediate and prompt notifications and direction to other required written reports. C. To direct the Emergency Coordinator to implement procedures which will ensure appropriate response as required by the classified emergency level. II. EMERGENCY CLASSIFICATION DESCRIPTIONS A Emergency Classes: I. The NRC and Federal Emergency Management Agency (FEMA) established four emergency classes for fixed nuclear facilities.

2. An emergency class is used for grouping off-normal nuclear power plant conditions according to their relative radiological seriousness and the time sensitive onsite and offsite actions needed to respond to such conditions.

3. The four emergency classes are (in order):

Unusual Event (UE) Least Severe Alert (A) Site Area Emergency (SAE) General Emergency (GE) Most Severe B. Unusual Event:

l. Plant events which are in progress or have occurred which indicate a potential degradation of the plant safety level.

2. The lowest level of emergency at the plant, which can usually be handled by the normal operating shift.

HCGS REV. 00

ECG Section i Pg 2of10

3. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs. Dose consequences in Unrestricted Areas would not reach 20 mRem TEDE.

C. Alert:

1. Plant events which are in progress or have occurred that are more serious than an Unusual Event which involve an actual or potential substantial degradation of the plant safety level.

2. Emergency Response personnel are required in addition to the nonnal operating shift. The entire emergency response organization is called in.

The TSC is activated, and the EOF and ENC are manned and may activate if needed for support.

3. Any release of radioactive material is expected to be limited to a small fraction of the EPA Protective Action Guideline exposure levels. Dose consequences in Unrestricted Areas would not reach 100 mRem TEDE.

D. Site Area Emergency: I. Serious plant events are in progress or have occurred which involve actual or likely major failure of plant functions required for protection of the public.

2. The entire emergency response organization is activated.

3. Any release of radioactive material is not expected to exceed EPA Protective Action Guideline exposure levels beyond the plant boundary.

Dose consequences in Unrestricted Areas not to exceed 1000 mRem TEDE. E. General Emergency:

1. Serious plant events are in progress or have occurred which involve actual or imminent core degradation or core melting with potential for loss of containment integrity.

2. The entire emergency response organization is activated.

3. Release of radioactive material can be expected to exceed EPA Protective Action Guideline exposure levels of 1000 mRem TEDE in Unrestricted Areas.

HCGS REV. 00

ECG Section i Pg 3of10 III. EVENT CLASSIFICATION GUIDE (ECG) STRUCTURE A. Overall Layout: The ECG is divided into 4 segments which are: I. Front Matter: Information which includes the Table of Contents, Introduction & Usage, and a Glossary of Acronyms.

2. Classification Sections: Flow chart diagrams used to classify events/conditions as emergencies or non-emergencies.

3. Attachments: Implementing documents that provide direction for emergency and non-emergency classification, notification, reporting requirements, references and forms required to facilitate event communications.

4. ECG Chart: Wall chart (Located at Emergency facilities) used to classify events/conditions as emergencies.

B. Classification Sections Format With the exception of ECG Section 3. 0, the ECG section flowcharts are comprised of the following elements:

1. Initiating Condition (IC): A generic nuclear power plant condition or event where either the potential exists for a radiological emergency or non-emergency reportable event OR such an emergency or non-emergency reportable event has occurred.

2. OPCON: Refers to the Operational Condition at Hope Creek during which a particular IC/EAL is applicable. The Operational Condition that the plant was in when the event started, prior to any protection system or operator actions. should be utilized when classifying events.

(from HCGS Technical Specifications, Sect. I, Definitions) OPCON TITLE MODE SWITCH RCS TEMP I. POWER OPERATION Run ANY

2. STARTUP Startup/Hot Standby ANY

                 ....) . HOT SHUTDOWN                Shutdown                       > 200 °F

4. COLD SHUTDO\VN Shutdown .'.S 200 °F

5. REFUELING"' Shutdown Or Refuel .'.S 140 °F HCGS REV. 00

ECG Section i Pg 4of10

Fuel in the RPV with the head closure bolts less than fully tensioned or with the head removed.

3. EAL Number (EAL#): Each Emergency Action Level (EAL) has been assigned a unique alpha numeric identifier. EAL# s are used in communication within PSE&G's Emergency Response Organization as well as when communicating with offsite officials who use an Offsite Reference Manual which is indexed in accordance with the EAL#s.

Each digit of the EAL# has a specific meaning that is not important to the users, but is important to the personnel who develop and maintain the ECGs. The digit and EAL# are defined below. Example EAL#= 9.4.1.a First Digit= Identifies which section of the ECG that a particular EAL is contained in. In the example the Digit 9 identifies that the EAL is from Section 9, Hazards - Internal/External. Second Digit Identifies the subsection that the EAL is contained in. In the above example the Digit 4 identifies that the EAL is found in subsection 4 of Section 9 thus 9.4, Toxic Gases. Third Digit = The third digit identifies the emergency class associated with that particular EAL as follows: If 3rd Digit is a I, then EAL results in VE If 3rd Digit is a 2, then EAL results in A If 3rd Digit is a 3, then EAL results in SAE If 3rd Digit is a 4, then EAL results in GE If looking at a RAL in Section 11 ONLY, the Third Digit identified the type of non-emergency event report to be made as follows. If 3rd Digit is a I, then .RAL is l hr report If 3rd Digit is a 2, then RAL is 4 hr report If 3rd Digit is a 3, then RAL is 24 hr report OR GREATER HCGS REV. 00

ECG Section i Pg 5of10 Fourth Digit= If a fourth digit is used, it is always a lower case letter and delineates one of multiple events which lead to similar emergency or non-emergency class levels. In the above example the "a" delineates I of 3 EALs that result in an Unusual Event and fall under a common Initiating Condition.

4. Emergency Action Level (EAL) or Reportable Action level (RAL): A predetermined, site-specific, observable threshold used to define when the generic initiating condition has been met, placing the plant in a given emergency class or non-emergency report. An EAL/RAL can be an instrument reading, an equipment status indicator, a measurable parameter, a discrete observable event, analysis results, entry into specific EOPs, or another phenomenon which indicates the need for classification of an emergency or non-emergency.

5. Action Required: Identifies the specific emergency class or non-emergency report that is required and refers the user to a specific ECG Attachment for implementation direction for the emergency or non-emergency event declared.

C. ECG Attachments: The ECG Attachments are written in various formats depending on their intended use. The attachments are used for implementing notifications, protective actions, directions to Emergency Plan Implementing Procedures (EPIPs), as well as providing essential phone listings and informational data for immediate reference. D. ECG Chart: (Located at Emergency Facilities)

1. Emergency Action Level (EAL): A predetermined, site-specific, observable threshold used to define when the generic initiating condition has been met, placing the plant in a given emergency class . An EAL can be an instrument reading, an equipment status indicator, a measurable parameter, a discrete observable event, analysis results, entry into specific EOPs. or another phenomenon which indicates the need for classification of an emergency.

HCGS REV. 00

ECG Section i Pg 6of10

2. OPCON: Refers to the Operational Condition at Hope Creek during which a particular EAL is applicable. The Operational Condition that the plant was in when the event started, prior to any protection system or operator actions, should be utilized when classifying events.

(from HCGS Technical Specifications, Sect. I, Definitions) OPCON TITLE MODE SWITCH RCS TEMP

1. POWER OPERATION Run ANY

2. STARTUP Startup/Hot Standby ANY

3. HOT SHUTDOWN Shutdown > 200 °F

4. COLD SHUTDOWN Shutdown :::: 200 °F

5. REFUELING* Shutdown Or Refuel :::: 140 °F

Fuel in the RPV with the head closure bolts less than fully tensioned or with the head removed.

3. The specific emergency classification identifies the ECG Attachment for implementation. Specific EALs identify "Common Site Events -

Attachment 24" for implementation. IV. EVENT CLASSIFICATION GUIDE (ECG) USE CAUTION ECG Sections referenced in other documents may have incorrect numbers, ASSESS the event and/or plant conditions and DETERMINE which ECG section(s) is most appropriate. A. EC Judgment: The EALs described in the ECG are not all inclusive and will not identify each and every condition, parameter or event which could lead to an event classification. The following guidance should be used by the EC; IF an EAL has been exceeded, but satisfaction of the IC is in question, THEN CLASSIFY the event JAW the EAL. IF however, it is clear that the EAL has NOT been exceeded (and will not), THEN DO NOT classify the event based solely on the IC. IF an IC has been satisfied, but exceeding the specific EAL is in question, THEN CLASSIFY the event IA W the IC. HCGS REV. 00

ECG Section i Pg 7of10 In any case, IF the plant conditions are equivalent to one of the four emergency classes as described in Section II above, THEN CLASSIFY the event based on EC discretion IAW ECG Section 4.0. Assessment Time: Assessment of an Emergency Condition should be completed in a timely manner which is considered to be within about 15 minutes of recognition of an event. If an EAL specifies a duration time (e.g. loss of annunciators for> 15 min.), then the assessment time runs concurrently with the EAL duration time and is the same length. If an event is recognized or reported and the required duration time is known to have already been exceeded then the duration portion of the EAL should be considered as being satisfied and the assessinent time for the remaining portions of the EAL should be within about 15 minutes from the time of recognition. B. Implementing Actions: The ECG is not a stand alone document. At times, the ECG will refer the user to other attachments or procedures for accomplishment of specific evolutions such as: Accountability, Recovery, development of P ARs, etc. They should be followed in a step-by-step fashion. The ECG should be considered an "Ir:plementing Procedure" and used in accordance with the requirements of a "Category II" procedure as defined in NC.NA-AP.ZZ-0001 (Q). The EC G's classification sections allow for judgment and decision making as to whether or not an EAL or RAL is exceeded. C. Classification: To use this ECG volume, follow this sequence: NOTE Confirmation of actual plant conditions should be made by comparing redundant instrumentation, indications, and/or alarms.

1. ASSESS the event and/or plant conditions and DETERMINE which ECG section(s) is most appropriate.

2. REFER to Section EAL/RAL Flowchart diagram(s), review and identify the Initiating Conditions that are related to the event/condition that has occurred or is ongoing.

(ECG Section 3.0 has its own unique usage instruction as part of the Fission Product Barrier Table 3 .0) HCGS REV. 00

ECG Section i Pg 8of10 NOTE The Emergency Coordinator should classify and declare an emergency before an Emergency Action Level (EAL) is exceeded if, in the EC's judgment, it is determined that the EAL will be exceeded.,

3. REVIEW the associated EALs or RALs as compared to the event and SELECT the highest appropriate emergency or reportable action level. If identification of an EAL is questionable refer to paragraph IV.A above.

If there is any doubt with regard to assessment of a particular EAL or RAL, the ECG Technical Basis Document should be reviewed. Words contained in an EAL or RAL that are bold face are either threshold values associated with that action level or are words that are defined in the basis for that specific EAL/RAL.

4. IDENTIFY and IMPLEMENT the referenced Attachment under Action Required ..

5. CONTINUE assessment after classification and attachment initiation, by returning to the ECG Sections to review EALs that may result in escalation/deescalation of the emergency level.

D. Emergency/Non-Emergency Conditions Discovered After-The-Fact Guidance NOTE Plant emergency events that are in progress or that have occurred with ongoing adverse consequences/effects should not be considered "After-The-Fact" events and should therefore be classified and declared as an ongoing emergency event.

1. EMERGENCY CONDITIONS - if "After-The-Fact" (not on-going at the time of discovery) it is discovered that an event or condition occurred that exceeded an Emergency Action Level (EAL), but was not declared as an emergency, then an emergency declaration is NOT required. A non-emergency, One-Hour Report should be initiated in accordance with ECG Section 11.6, After-The-Fact.

2. NON-ErvlERGENCY COI\1DITIONS - if After-The-Fact (regardless of whether the event is on-going at the time of discovery) it is discovered that an event or condition had occurred that should have resulted in the HCGS REV. 00

ECG Section i Pg9of10 classification and implementation of a non-emergency report ( 1 hour, 4 hour. 24 hour), the applicable non-emergency report attachment in the ECG should be implemented. E. NRC Communications During An Emergency Guidance

1. Complete and accurate communications with the NRC Operations Center during emergencies is required and expected. The purpose of notifying the NRC within one-hour of an emergency, is to provide event information when immediate NRC action may be required to protect the public health and safety OR when the NRC needs accurate and timely information to respond to heightened public concern. If the information we provide is not accurate or does not contain sufficient detail, then we hamper the NRC from doing their job.

2. The NRC Data Sheet, along with the Initial Contact Message Form, is the primary vehicle to ensure the NRC is kept informed. General Guidance on completing the event description portion of the NRC Data Sheet is provided in Attachment 5 of the ECG.

F. Voluntary/Courtesy Reporting of Non-Emergency Events Guidance In accordance with NUREG I 022, Rev I. voluntary reporting is encouraged. PSE&G may make voluntary or courtesy NRC notification (RAL 11.10.2) concerning events or conditions which may be of interest to the NRC. The NRC responds to any voluntary notification of an event or conditions as its safety significance warrants. regardless of how PSE&G classifies the event. IF it is determined at some later time that the event was reportable under a specific part of I OCFR50. 72 as defined in the ECG, THEN PSE&G should update the NRC with this information. G. Event Retraction Guidance IF an ENS notification to the NRC was made as directed by the applicable ECG Attachment AND it is later determined that the event or condition is not reportable, THEN the notification may be retracted as follows:

1. OBT ATN both the Hope Creek General Manag~r's and Operations Manager's approval of any proposed retractions.

HCGS REV. 00

ECG Section i Pg 10of10

2. COfvfPLETE "page l" of the NRC Data Sheet which was implemented to make the original notification. Event Description Section of NRC Data Sheet should explain the rationale for the retraction.

3. NOTIFY the NRC Operations Center and NRC Resident Inspector

4. RECORD on the "NRC Data Sheet" the name of the NRC Contact that received the retraction information.

5. FORWARD the retraction "NRC Data Sheet" with the rest of the original attachment of the ECG that was implemented when the original notification was made.

H. Non-emergency Information Update Guidance IF additional information needs to be transmitted to the NRC concerning a previously reported non-emergency event, THEN MAKE notifications as follows: I. COJ\.1PLETE Page 3 of the NRC Data Sheet form for event update.

2. OBTAIN the approval of the SNSS to release the information.

3. NOTIFY all organizations and individuals who were initially contacted AND DOCUt-.1ENT the update.

4. FORWARD all update paperwork with the original ECG Attachment package.

I. Common Site E\"ents Guidance Selected EALs (Unusual Event level only) and selected RALs have been designated as "Common Site" events. These events will be annotated with the words, "Common Site" in the Action Required portion of the EAL sections. Common Site Events need not be reported by both Salem and Hope Creek. The referenced ECG Attachment will direct the SNSS's to establish agreement on which SNSS will declare and report the event. Events classified at an Alert or higher level require plant specific information to be provided to the states of New Jersey and Delaware, the NRC, and to PSE&G Emergency Response Facilities and therefore will not be classified as common site events. HCGS REV. 00

ECG Section ii Pg 1 of 5 HOPE CREEK EVENT CLASSIFICATION GUJ.Il.~--H**--* --,1 Glossary of Acronyms & Abbreviation~Dr;*-;-;1c;L COP\! # Sect" ** 1 100 II I t J c;s-- ; Ii AAAG Acciden~ Assessment Advisory Group (Dra~are) AC Altematmg Current --- ADS Automatic Depressurization System ALARA As Low As Reasonably Achievable APRM Average Power Range Monitor ARI Alternate Rod Insertion ARM Area Radiation Monitor ASAP As Soon As Possible ASM Administrative Support Manager AS Administrative Supervisor ATWS Anticipated Transient Without Scram BKGD Background BKR Breaker (electrical circuit) BNE Bureau of Nuclear Engineering (NJDEPE) CACS Containment Atmosphere Control System CAS Central Alarm Station CCPM Corrected Counts per Minute CEDE Committed Effective Dose Equivalent CDE Committed Dose Equivalent CFR Code of Federal Regulations CIS Containment Isolation System CNTMT Containment (Barrier) CP Control Point CPM Counts Per Minute CR Control Room CREF Control Room Emergency Filter System CRIDS Control Room Integrated Display System CRD Control Rod Drive css Core Spray System DC Direct Current DAPA Drywell Atmosphere Post Accident (Radiation monitor) DDE Deep Dose Equivalent DEi Dose Equivalent Iodine DEMA Delaware Emergency Management Agency DEP Department of Environmental Protection (NJ) HCGS Rev. 00

ECG Section ii Pg 2 of5 DID Direct Inward Dial (phone system) DLD Drywell Leak Detection _ DOE Department of Energy DOT Department of Transportation DPCC/DCR - Discharge Prevention, Containment, & Countermeasures/ Discharge Cleanup & Removal Plan DPM Disintegrations per Minute DRCF Dose Rate Conversion Factor EACS ESF Equipment Area Cooling System EAL Emergency Action Level EAS Emergency Alert System (Broadcast) EC Emergency Coordinator ECCS Emergency Core Cooling Systems ECG Emergency Classification Guide EDG Emergency Diesel Generator EDO Emergency Duty Officer EMRAD Emergency Radio (NJ) ENC Emergency News Center ENS Emergency Notification System (NRC) EOC Emergency Operations Center (NJ & DE) EOF Emergency Operations Facility EOP Emergency Operating Procedure EPA Emergency Preparedness Advisor EPA Environmental Protection Agency EPC Emergency Preparedness Coordinator EPIP Emergency Plan Implementing Procedure EPZ Emergency Planning Zone EQPT Equipment ERDS Emergency Response Data System ERM Emergency Response Manager ERO Emergency Response Organization ESF Engineered Safety Feature ESSX Electronic Switch System Exchange (Centrex) FC Fuel Clad (Barrier) FFD Fitness For Duty FRVS Filtration, Recirculation, and Ventilation System FTS Federal Telecommunications System (NRC) GE General Emergency HCLL Heat Capacity Level Limit HCGS Hope Creek Generating Station HCGS Rev. 00

I I ECG Section ii Pg 3of5 HCTL Heat Capacity Temperature Limit HEPA High Efficiency Particulate Absorbers HPCI High Pressure Coolant Injection HTV Hardened Torus Vent HVAC Heating, Ventilation & Air Conditioning HWCI Hydrogen Water Chemical Irijection HX Heat Exchanger IAW In Accordance With IC Initiating Condition ICMF Initial Contact Message Form IDLH Immediately Dangerous to Life and Health IRM Intermediate Range Monitor I/S In Service KI Potassium Iodide KV Kilovolt LAC Lower Alloways Creek LCO Limiting Condition for Operation LDE Lens Dose Equivalent LEL Lower Explosive Limit LLD Lowest Level Detectable LOCA Loss of Coolant Accident LOP Loss of Offsite Power LPCI Low Pressure Coolant Injection LPZ Low Population Zone MCR Main Control Room MDA Minimum Detectable Amount MEA Minimum Exclusion Area MEES Major Equipment & Electrical Status (Form) MET Meteorological M.O.U. Memorandum of Understanding MRO Medical Review Officer MSIV Main Steam Isolation Valve MS IV SS Main Steam Isolation Valve Sealing System MSL Main Steam Line NA WAS National Attack Warning Alert System NCO Nuclear Control Operator NDAB Nuclear Department Administration Building (TB2) NEO Nuclear Equipment Operator NETS Nuclear Emergency Telecommunications System HCGS Rev. 00

ECG Section ii Pg 4of5 NFE Nuclear Fuels Engineer NFPB Normal Full Power Background

 " NJSP   New Jersey State Police NOAA   National Oceanographic and Atmospheric Administration NPV    North Plant Vent NRC    Nuclear Regulatory Commission NSS    Nuclear Shift Supervisor NS SSS Nuclear Steam Supply Shutoff System NSTA   Nuclear Shift Technical Advisor NUMARC Nuclear Management and Resources Council NWS    National Weather Service QBE    Operating Basis Earthquake OCA    Owner Controlled Area ODCM   Offsite Dose Calculation Manual OEM    Office of Emergency Management (NJ)

OHA Overhead Annunciator OPCON Operating Condition OSB Operational Status Board (Form) osc Operations Support Center PAG Protective Action Guideline PAR Protective Action Recommendation PASS Post Accident Sample System PC Primary Containment (Barrier) PCIG Primary Containment Instrument Gas System PCIS Primary Containment Isolation System PSIG Pounds Square Inch Gauge RAD Radiation RAL Reportable Action Level RC Reactor Coolant RCA Radiologically Controlled Area RCAM Repair and Corrective Action Mission RCIC Reactor Core Isolation Cooling RCS Reactor Coolant System (Barrier) RHR Residual Heat Removal (Containment Heat Removal) RM Recovery Manager RMO Recovery Management Organization RMS Radiation Monitoring System RPS Radiation Protection Supervisor RPS Reactor Protection System RPV Reactor Pressure Vessel RRCS Redundant Reactivity Control System HCGS Rev. 00

ECG Section ii Pg 5 of 5 RSM Radiological Support Manager RWCU Reactor Water Cleanup (System) SACS Safety Auxiliaries Cooling System SAE Site Area Emergency SAM Severe Accident Management SAS Secondary Alarm Station (Security) SBO Station Blackout SCBA Self Contained Breathing Apparatus SCP Security Contingency Procedure SDE Shallow Dose Equivalent SDM Shutdown Margin SLC Standby Liquid Control SJAE Steam Jet Air Ejector SNM Special Nuclear Material SNSS Senior Nuclear Shift Supervisor sos Systems Operations Supervisor (Security) SPDS Safety Parameter Display System SPV South Plant Vent SRM Source Range Monitor SRPT Shift Radiation Protection Technician SRV Safety Relief Valve SSCL Station Status Checklist SSE Safe Shutdown Earthquake ssws Station Service Water System SSNM Strategic Special Nuclear Material TAF Top of Active Fuel TDR Technical Document Room TEDE Total Effective Dose Equivalent TIP Traversing Incore Probe TLV Threshold Limit Value TIS Technical Specifications TSC Technical Support Center TSS Technical Support Supervisor TSTL Technical Support Team Leader TSTM Technical Support Team Member UE Unusual Event UFSAR Updated Final Safety Analysis Report UHS Ultimate Heat Sink USCG United States Coast Guard VDC Volts Direct Current WB Whole Body HCGS Rev. 00

1.0 Fuel Clad Challenge 1.1 RCS Activity 2.0 RCS Challenge 2.1 RCS Leakage 3.0 Fission Product Barriers (Table) 4.0 E C Discretion 4.1 Emergency Coordinator Discretion 5.0 Failure to SCRAM 5.1 ATWS 6.0 Radiological Releases/Occurrences

6.1 Gaseous Effluent Release 6.2 Liquid Effluent Release 6.3 In Plant Radiation Occurrences 6.4 Irradiated Fuel Event 7.0 Electrical Power 7.1 Loss of AC Power Capabilities 7.2 Loss of DC Power Capabilities 8.0 System Malfunctions 8.1 Loss of Heat Removal Capability 8.2 Loss of Overhead Annunciators 8.3 Loss of Communication Capability 8.4 Control Room Evacuation 8.5 Technical Specifications 9.0 Hazards - Internal/External 9.1 Security Threats 9.2 Fire 9.3 Explosion 9.4 Toxic/Flamable Gases 9.5 Seismic Event 9.6 High Winds 9.7 Flooding 9.8 Turbine Failure/Vehicle Crash/Missile Impact 9.9 River Level

HCGSECG 1.0 Fuel Clad Challenge Rev. 00 Page I of I 1.1 RCS Activity Initiating Fuel Clad Degradation Condition OPCON ( 1,2,3,4,S ) ( 1,2,3,4 ) ( 1,2,3,4 ) ( _____ 1,2,3 ____..) EAL# ~.1.1.a 1.1.1.b 1.1.1.c IF IF IF E M E Reactor Coolant Valid Offgas Pretreatment Valid Main Steam Line R Sample Activity Radiation Monitor Radiation Monitor j' G > 4 µCi/gm (9RX62 I I 9RX622) High High Alarm Condition E Dose Equivalent High Alarm Condition ( ~ 3 times Normal Full N c y 1-131 ( ~ 2.2E+04 mRem/hr) Power Background) I L_____ _ A c T ANY SRV I AND is determined 0 to be N Stuck Open L NOTE: THEN E v _______ Refer to Section 3.0, E Fission Product Barrier Table L prior to Event Classification s Action Required L___-----

                           !Refer to Attachment I
                                 ~NUSUAL EVENT Refer to       Attach~~nt ALERT 2]

1.0 Fuel Clad Challenge 1.1 RCS Activity 2.0 RCS Challenge 2.1 RCS Leakage 3.0 Fission Product Barriers (Table) 4.0 E C Discretion 4.1 Emergency Coordinator Discretion 5.0 Failure to SCRAM 5.1 ATWS 6.0 Radiological Releases/Occurrences 6.1 Gaseous Effluent Release ~

            . 6.2      Liquid Effluent Release 6.3      In Plant Radiation Occurrences 6.4      Irradiated Fuel Event 7.0   Electrical Power 7.1      Loss of AC Power Capabilities 7.2      Loss of DC Power Capabilities 8.0   System Malfunctions 8.1      Loss of Heat Removal Capability 8.2      Loss of Overhead Annunciators 8.3      Loss of Communication Capability 8.4       Control Room Evacuation 8.5       Technical Specifications 9.0   Hazards - Internal/External 9.1       Security Threats 9.2       Fire 9.3       Explosion 9.4       Toxic/Flamable Gases 9.5       Seismic Event 9.6       High Winds 9.7       Flooding 9.8       Turbine Failure/Vehicle Crash/Missile Impact 9.9       River Level

HCGSECG 2.0 RCS Challenge Rev. 00 Page I of I 2.1 RCS Leakage Initiating RCS Leakage Condition OPCON ( 1, 2, 3 ) ( 1, 2, 3 ) ( 1, 2, 3 J ( 1, 2, 3 ) EAL# 2.1.1.a 2.1.1.b 2.1.1.c 2.1.1.d IF IF IF IF E M Reactor Coolant System Reactor Coolant System Reactor Coolant System Successful Isolation of a E Identified Leakage Pressure Boundary Leakage. Unidentified Leakage Reactor Recirc Pump R > 25 gpm

                   > 10 gpm                  > 10 gpm                                                        Dual Seal Failure within G                                                                     averaged over any (Using 10 minute average) (Using 10 minute average)                                             10 minutes of recognition E                                                                       24 hour period N

c y THEN A c T I 0 NOTE: N Refer to Section 3.0, Fission Product Barrier Table L prior to Event Classification E v E L s Action Refer to Attachment 1 Required UNUSUAL EVENT

1.0 Fuel Clad Challenge 1.1 RCS Activity 2.0 RCS Challenge 2.1 RCS Leakage 3.0 Fission Product Barriers (Table) 4.0 E C Discretion 4.1 Emergency Coordinator Discretion r* 5.0 Failure to SCRAM 5.1 ATWS 6.0 Radiological Releases/Occurrences 6.1 Gaseous Effluent Release 6.2 Liquid Effluent Release 6.3 In Plant Radiation Occurrences 6.4 Irradiated Fuel Event 7.0 Electrical Power 7.1 Loss of AC Power Capabilities 7.2 Loss of DC Power Capabilities 8.0 System Malfunctions 8.1 Loss of Heat Removal Capability 8.2 Loss of Overhead Annunciators 8.3 Loss of Communication Capability 8.4 Control Room Evacuation 8.5 Technical Specifications 9.0 Hazards - Internal/External 9.1 Security Threats 9.2 Fire 9.3 Explosion 9.4 Toxic/Flamable Gases 9.5 Seismic Event 9.6 High Winds 9.7 Flooding 9.8 Turbine Failure/Vehicle Crash/Missile Impact 9.9 River Level 6-.*

n TABJ ~E 3.0

                           -cot~IROL CIW'llf (g_~-

IJ I 3.1 8.1.1 Bl'.A.Cl'OB WA'lml LEVEL

                                                                                                    ~** Clad Barri"                                           :         3.2 RCS Barrier el 1r""'

D 8.8.1 REACTOR WATEBLEVEL 3.3 CNTMT Barrier FISSION P DUCT POTEN11AL l.088

BPr.

MLtlU.La MLll&Ll.b KALIS.2.1.a LOSS* <WT* EAL *8.2.1.b POI'EN'l1AL LOBS

1Pf EALtSJl.J. Not Appllcahle l.OSS :OPI'!

a-etol'WaWrL!rnd. BARRIERS ~-Ulr('l'upor Acdn1 Fuel) El'.CUJDJNG lll18Dtlonal 1owatog o1 a-mor S-ctor W.ter 1""el CAHNarBEREfm>B.ZD AND MAOO'AJNED al>cor8 - 200" OOnlmna z..o 8-tor Waler LeTel

                                                                                                                                        ~-129"
                                                                                                                                         ~

8-&or Water Lenil

                                                                                                                                                                           ~-16l"(Topor Active Fnel) EXCLUDING lnleullonal loworln* d. U-etar ln&emlonal lowerlna' d. Beaetor
                                                                                                                                                                                                              ~rW*t.erLerrel CANNOT BE BFBJ'ORED AND MAJNTAINEQ
                                                                                                                                                                                                               *hove -200" CMlnlmwq Zero Wat.er LeTel dmtng an ATW8        loJectlon BPV 'Wder Lend.)        Wda' 1-el durlq -.n ATW8          W*l<<' Lerel during u ATWB           IDJedloo RPV Wa&er ievel)

APPLICABLE ~ DBYWELL PRF.88UREJ B, OPERATIONAL

I LOSS. -4PT*

POTENTIAL l.088

1 PT CONDITIONS ARE EAL I 8.2.2-a EAL I 3.'-2.b EALtS.8.2..a Supp Cbambel' ~ CANNOf EALtaa.2.b C.:.,tam.ll!Dtf.rJD-"'*lndlcaW 1,2,3 ONLY Uulaolahte ace 1.-.k a.&.e

.GOGPMINSIDE Primary Conialnmeal Valid IDgb lhywelJ ,.,_,,.,

Cc:mdlUoo. ( :z: J.68 pslg ) BE MA.lNTAJNm ~a. 6llpal!{ ........ bt'*.....,Addroplno.y..,D~

                                                                                                                                                                                                                                               ~*""--111~.bo.e EALI .......                          EAl.18.8.2.d Prtmary Coolalnmeot B,coooentntloo > -4'\. and       ~~
                                                                                                                                                                                                                                               -.iW..tw:lilr.LOCA~
                                                                                                                                                                                                                                                                   .........°'

Nm'K O,eoncentntloo > lRo u the t.o.. o:r Potential ~OCJiDlddmoed IMMINENT EALlll..lle (may occur within ii houn), 1181!1 Jud&iauml and cl.aadty &1111 Coatamr:. .... t b v..,,Ud by the U the thrMhold. la mr-1ed. 8.1.J DKYWEI.L ATM06PllERE P08T ACX::IDENT a>APA.) F.--.-,.OJ>-.tm*~ RADIAnON LEVEL lruitructions: P01EN'nAL 1.()88 * <Pl'll 8.&8 DRYWELL A.TMOfWJIEB.E POln' ACCIDENT <DAPA>

1. In the table review the Emergency Action Not App)leahle EAL I 8.1.! RAD..ATION LEVEL Level8 of all cohmuuo and Identify which DAPA B.adlatJoo Momtcw PO'l'ENTIAL 1.088 a 1 PT

                                                                                                         ,_dlq: " 6000 Mr n - further review.

EALtS.S.S NotAppllcable DAPABadlaUon Moollor

2. For each of the three blll'l'leno, determine the remtlng: ;p, 28000 R/hr EAL with the hlgh...t point value, and circle 8.2.BBCS LINE BREAIL'CONTA.INMENT BYPASS the correopondlng EAL I and point value.

No more the one EAL llhould be oelected for PC1J'ENI1AL l.088

3PI'9 wss.4Pr* a.&4. RCS LINE BB.EAKICONTAJNMENT BYPASS each barrier. PO'n!N'11AL L088
I Pl' EAL I 8.2-3-a EALtB.2-ll.b WSS:i2f'l'*

Ma.In Bteam. line Break RCS Une Br-.11: Q!ZDllDB EA.I..13.S.4.9

3. Add the point valuea circled for the three Prtmary Coot.abunoo.L, EAL*S.S.4..b Q!!ll!!!!!I RCS lJne Break Q1l1lilllt lllolatlon BlftD.al for ANY barrlen and enter the BU.JD. below: Primary Cootahun6DL, J"IEmlllloglo

Valldholallon ooeoftbe roDowtng~

Primary Contalmaent, remhinglo-.a~ 81gnal for~ ooe d the MSIV hola.Uoo Slp.al ncdtiq In* Valld laolaUon following .ya&t'll!lm

NBS88 Bi,o.aI for Am' one or the

4. ClaMlfy follows:

ha- on the point value IRllD ao

8. L3 RCS IODJNE CONCENTILNl10N

                                                                                                                                        .11!1!

N8888

                                                                                                                                                                            - HPCI

BCIC folJowtDS myatems:
NBSSS (e:eludlng Main 8team.Lln8)

                                                                                                                                                                                                              *BPCI
                                                                                                                                                                                                              *RCIC PClB llPCI RCIC Uthe             Clnll!dfy am              Refer tot                   POTENTIAL L088

o Pl'* lnd.leaUood~

IRllD lo: n.ow 11 E"HACiE 1,2 S,4 UNUSUAL EVENT ALERT Attachment 1 Attachment 2

                                                                                                        -Coolmrt-         MLlll.l.S AdlYUy 11 ~ uCL'p:i De..

Equtnleot 1-181 AFmR vain clc.tre from lhe iiaiili5ootrol lloom hM ~ Slllll!lllll THE PRIMARY CoolalDIDaJ.l lhrougb the effected. eyste:m M:mB. valve closure from the liO lndfeation or CONIDIUJNG fl.pWIJ...EAKAGEOlJTSIDE GE 5,6,7,11 SITE AREA Attachment 3 Mato Control Booa::i hM the Prlnary ContaJm:aenl throuab the effected ayt1tem A17.1B valve clowre from effected system~ Primary t through lh8 tbe Main Ccmtrol Hoom baa valve closure from Main 9,10 GENERAL Attachment 4 bil!!leD*Ueuipted Cb:itrol Room W beoo

5. Implement the appropriate ECG 8.14 l:MEBGENCY COORDINATOR .JUDGMENT BAB EMERGENCY COORDINATOR .nJOOMENT Attachment per above chart.

ron:NTIAL LOSS* 8 PI'a

a. Contlnne to review the EAUo on this Table EAL*S.L4.a EALl&.JA.b

                                                                    .4HI CIODdldon. lo the oplnkm d 4HI tJODdltkm,, In lh8 apnlon                            EAL

8.2.4.a RAI.*3.14.b EALllUl.11.a EALtll.lLO.b for chang"" that conld retJOlt In emergency Alfi' cooditlon,, lo the oplo!on AMI ocudttloo. lo the oplolon ANY eondltlon,. ln th8 opinion ANY cxiodftloa., lo &bu oplnlon the ZC. lhat lndh.W. a PatiiimW ol. &be EC. thal hldlcai. a Lo. ol the r.c. lhal. lndleate.
ol the EC. lhal. lndlealea a Lo.

eocalatlon or ~tlon. ol the EC, that lodleatea a ol the EC. thal lndlcatm

U-U- ~the Fuel Clad BuTler ol tba FtMi Clad Barrlcw P~u.J Lo. oltM RCS eun.- o1 the nee Barri- Poleutlal I.--~ the ol Iba Cootalom.8Dt Burler Coatalnm8Dt Burler HCalr.m RmOO

1.0 Fuel Clad Challenge 1.1 RCS Activity 2.0 RCS Challenge 2.1 RCS Leakage 3.0 Fission Product Barriers (Table) 4.0 EC Discretion 4.1 Emergency Coordinator Discretion ~ 5.0 Failure to SCRAM 5.1 ATWS 6.0 Radiological Releases/Occurrences 6.1 Gaseous Effluent Release .. 6.2 Liquid Effluent Release 6.3 In Plant Radiation Occurrences 6.4 Irradiated Fuel Event 7.0 Electrical Power 7.1 Loss of AC Power Capabilities 7.2 Loss of DC Power Capabilities 8.0 System Malfunctions 8.1 Loss of Heat Removal Capability 8.2 Loss of Overhead Annunciators 8.3 Loss of Communication Capability 8.4 Control Room Evacuation 8.5 Technical Specifications 9.0 Hazards - Internal/External 9.1 Security Threats 9.2 Fire 9.3 Explosion 9.4 Toxic/Flamable Gases 9.5 Seismic Event 9.6 High Winds 9.7 Flooding 9.8 Turbine Failure/Vehicle Crash/Missile Impact 9.9 River Level

HCGSECG 4.0 E C Discretion Rev.00 Page I of I 4.1 Emergency Coordinator Discretion Other Conditions Exist * *, . Other Conditions Exist

                                                                                                  )    /other Conditions Exist Whic~           Other Conditions Exist Whi~

Which In the Judgment of the Which In the Judgment of the In the Judgment of the Emergency In the Judgment of the Emergency ) Initiating ( ( Emergency Coordinator Warrant . Emergency Coordinator Warrant ( Coordinator Warrant Declaration of Coordinator Warrant Declaration of Condition pL'Claration of an Unusual Event ' ______________ Declaration of an Alert .... ,.,. ',, a Site Area Emergency \ a General Emergency .-

                                                                                                          ---*--------~                     '------                       ~/

OPCON All ) All All )

                                                                                                                                                         -----~~---)

EAL# 4.1.1 4.1.2 4.1.3 4.1.4

-------*, IF IF IF IF E r*- -- *- - - . . ------*-- 1- ------------- --- *-- -**--.. -

M Events are in progress or E Events are in progress or 1 Events are in progress or Events are in progress or have occurred which, in the have occurred which, in the have occurred which, in the have occurred which, in the R judgment of the Emergency G judgment of the Emergency judgment of the judgment of the Emergency Coordinator, indicate a Coordinator, indicate Emergency Coordinator, Coordinator, indicate an E Potential Degradation N EITHER one of the indicate an Actual or imminent of Plant Safety following: Actual or likely major substantial core c y

Plant safety systems failure of plant functions degradation with the THEN needed for protection of potential for loss of (more than one) are or A the public containment may be degraded c
ANY Plant Vital T

I Structure is degraded or potentially degraded THEN 0 N AND L E Increased monitoring of v Safety Functions E is warranted L THEN s Action Refer to Attachment I Refer to Attachment ~ Refer to Attachment 3 Refer to Attachment 4 Required UNUSUAL EVENT ALERT SITE AREA EMERGENCY GENERAL EMERGENCY L-----------------

1.0 Fuel Clad Challenge 1.1 RCS Activity 2.0 RCS Challenge 2.1 RCS Leakage 3.0 Fission Product Barriers (Table) 4.0 E C Discretion 4.1 Emergency Coordinator Discretion II' r 5.0 Failure to SCRAM 5.1 ATWS 6.0 Radiological Releases/Occurrences 6.1 Gaseous Effluent Release ,. 6.2 Liquid Effluent Release 6.3 In Plant Radiation Occurrences 6.4 Irradiated Fuel Event 7.0 Electrical Power 7.1 Loss of AC Power Capabilities 7.2 Loss of DC Power Capabilities 8.0 System Malfunctions 8.1 Loss of Heat Removal Capability 8.2 Loss of Overhead Annunciators 8.3 Loss of Communication Capability 8.4 Control Room Evacuation 8.5 Technical Specifications 9.0 Hazards - Internal/External 9.1 Security Threats 9.2 Fire 9.3 Explosion 9.4 Toxic/Flamable Gases 9.5 Seismic Event 9.6 High Winds 9.7 Flooding 9.8 Turbine Failure/Vehicle Crash/Missile Impact 9.9 River Level A.

HCGSECG 5.0 Failure to Scram Rev.00 Page I of I 5.1 ATWS Failure of the Reactor Protection System (RPS)

                                                               ~)
                                                                       ,,,-----Failure of the Reactor Protection Failure of the Reactor Protection System (RPS)              System (RPS) to Successfully Complete              to Successfully Complete a Reactor Scram Initiating      to Successfully Complete a Reactor Scram

( a Reactor Scram (Automatic and Manual) (Automatic and Manual) and there is indication of an Condition (Automatic or Manual) , and Reactor Power is above 4% Extreme Challenge to the Ability lo Cool the Core OPCON (.__ __..,,, 1, 2 ( 1, 2 ) ( I, 2 ) ( 1, 2 ) EAL# 5.1.2.a 5.1.2.b 5.1.3 5.1.4 IF IF E -- M An Automatic ANY Manually E Reactor Scram Initiated R Condition exists Reactor Scram G (RPS) E I AND from the N An Automatic Control Room c Reactor Scram IS NOT y (RPS) IS NOT successful successtrul A c I I ALL Reactor Scram attempts from the Control Room A ND T -----*-------***------- (RPS and ARI) DID NOT - - - - - - - - - - - - - - - ------------- R I REDUCE and MAINTAIN EITHER one of the following: 0 Reactor Power to ,::: 4 % N

Reactor Water Level CANNOT BE AND MAINTAINED> -190" L
The combination of Suppression Pool E

Temperature and RPV Pressure v CANNOT BE MAINTAINED E below the HCTL Curve L s ,. ~-------~--------*-- i THEN I 0 Action o Attach~mnt 2 Refer to Attachment 3 Refer to Attachment 4*------- Required ALERT SITE AREA EMERGENCY

                                                                       ~-----------*------

L____ GENERALEMERGENCY

                                                                                                                                 -~-----*---- . -     -*----------------  ---

1.0 Fuel Clad Challenge 1.1 RCS Activity 2.0 RCS Challenge 2.1 RCS Leakage 3.0 Fission Product Barriers (Table) 4.0 E C Discretion 4 .1 Emergency Coordinator Discretion 5.0 Failure to SCRAM 5.1 ATWS 6.0 Radiological Releases/Occurrences 6.1 Gaseous Effluent Release ' 6.2 Liquid Effluent Release 6.3 In Plant Radiation Occurrences 6.4 Irradiated Fuel Event 7 .0 Electrical Power 7 .1 Loss of AC Power Capabilities 7 .2 Loss of DC Power Capabilities 8.0 System MaJfunctions 8.1 Loss of Heat Removal Capability 8.2 Loss of Overhead Annunciators 8 .3 Loss of Communication Capability 8.4 Control Room Evacuation .._. 8.5 Technical Specifications 9 .0 Hazards - Internal/External 9 .1 Security Threats 9.2 Fire 9 .3 Explosion 9.4 Toxic/Flamable Gases 9 .5 Seismic Event 9 .6 High Winds

9. 7 Flooding 9.8 Turbine Failure/Vehicle Crash/Missile Impact

9. 9 River Level

HCGSECG 6.0 Radiological Releases/Occurrences Rev. 00 Page I of4 6.1 Gaseous Effluent Release Initiating 'l\ny Unplanned Release of Gaseous Radioactivity to*' Any Unplanned Release of Gaseous Radioactivity to Condition ( Environment that Exceeds 2 Times the Radiological , the Environment that Exceeds 2 times*the IOCFR20, Appendix 8 limits for 60 minutes or longer

            ..._ Technical Specifications for 60 minutes or longer __ )

OPCON ( All J ( All J ( All

                                                                                                            )                                      (     All   J EAL#                    6.1.1.a                      6.1.1.b                                    6.1.1.c                                               6.1.1.d Field Dose        IF              Measured     IF                                         IF                        Alarm                       IF Assessment                       Dose Rate                                  is                                  Indications E                                                                                                                      ~------------------------

M Dose Assessment Valid High Alarm received from ANY one of the Dose Rate measured Gaseous effiuent release E indicates EITHER following Plant Effiuent RMS Channels: at the Protected sample analysis for ANY one R one of the following Area Boundary of the following indicates a

FRVS Noble Gas (Grid 1/3; 9RX680) at the MEA or beyond or beyond EXCEEDS concentration of:

G as calculated on

NPV Noble Gas (Grid 1/3; 9RX590)

E .05 mRem/hr the SSCL:

FRVS:
NPV Iodine (Grid 3; 9RX601) above normal N background ;::: l.IJE-03 pCi/cc Total Noble Gas
SPV Noble Gas (Grid 1/3; 9RX580)
TEDE 4-Day Dose c ;::: 2.0E-01 mRem

                                                                              ;::: 2. 71 E-07 pCi/cc 1-131

SPV Iodine (Grid 3; 9RX605) y
NPV:
Thyroid-COE Dose
HTV Noble Gas (Grid 3; 9RX516)

                                                                              ~     2.43E-04 pCi/cc Total Noble Gas A             ;::: 6.8E-Ol mRem                                           '.'.: 5.SIE-08 pCi/cc 1-131                                                  I AND c       based on Plant Vent effiuent sample analysis

SPV: Total Plant Vent release rate EXCEEDS T and NOT on a default . :=: 2.27E-05 pCi/cc Total Noble Gas EITHER one of the fol!owiffg*1~~rrs:****--1 I Noble Gas to Iodine '.'.: 5.44E-09 pCi/cc 1-131 0 Ratio
4.80E+o3 µCi/sec Total NoblJ Gas *

                                                                                                                                                               *.            e***.

N

1.15E+oo µCi/sec 1-13l(NPV ~ SPV QfilY) ',:*_:!

I I ....... I AND

L r:~:;~

r* - [ Dose Assessment results ~ available E C) v I J

                                                                                                                                                                              <.:)
                                                                                                                                                                              '"'CJ E                                                        -* *-*------------, AND

tt:

L -~------~ s Release is ongoing for::=: 60 minutes l THEN J IL. ___ Action - Refer to Attachment l Required [ UNUSUAL EVENT

HCGSECG 6.0 Radiological Releases/Occurrences Rev.00 Page 2 of 4 6.1 Gaseous Effluent Release Any Unplanned Release of Gaseous Radioactivity to Any Unplanned Release of Gaseous Radioactivity to the Environment that Exceeds Initiating Environment that Exceeds 200 Times the Radiological 200 times the IOCFR20, Appendix B limits for 30 minutes or longer Condition Technical S cifications for 15 minutes or longer ( All ) ( All ) ( All ) ( All ) OPCON EAL# 6.1.2.a 6.1.2.b 6.1.2.c 6.1.2.d Field Dose IF Measured IF Sample IF Alarm IF Assessment Dose Rate Analvsis Id. r ns n 1ca10 E Valid High Alarm received from ANY one of the M Dose Assessment Dose Rate measured Gaseous effiuent release following Plant Effiuent RMS Channels: E indicates EITHER at the Protected sample analysis for ANY one one of the following Area Boundary of the following indicates a

FRVS Noble Gas (Grid 1/3; 9RX680)

R at the MEA or beyond or beyond EXCEEDS concentration of:

NPV Noble Gas (Grid 1/3; 9RX590)

G as calculated on 5 mRem/hr E the SSCL:

FRVS:
NPV Iodine (Grid 3; 9RX601)

N ~ l.IJE-01 pCi/cc Total Noble Gas

SPV Noble Gas (Grid 1/3; 9RX580)
TEDE 4-Day Dose c ~ 2.0E+ol mRem

                                                                            ~ 2.71 E-05 pCi/cc 1-131

SPV Iodine (Grid 3; 9RX605) y
NPV:
HTV Noble Gas (Grid 3; 9RX516)
Thyroid-COE Dose ~ 2.43E-02 pCi/cc Total Noble Gas A ~ 6.8E+ol mRem ~ 5.81 E-06 pCi/cc 1-131 I AND c based on Plant Vent effiuent sample analysis
SPV: Total Plant Vent release rate EXCEEDS T and NOTon a default ~ 2.27E-03 pCi/cc Total Noble Gas EITHERone of the following limits:

I Noble Gas to Iodine ~ 5.44E-07 pCi/cc 1-131

4.80E+o5 pCi/sec Total Noble Gas 0 Ratio N
l.15E+o2 pCi/sec 1-B l (NPV & SPV ONLY)

I I L I AND E Dose Assessment results NOT available v I AND I I

                                                                                                   '                 I AND E      r     Release is ongoing   for~    15 minutes      I L                                ITHEN                                                   I   Release is ongoing for~ 30 minutes            l s                                                                                                               1 THEN l

Action Refer to Attachment 2 Required ALERT

6.0 Radiological Releases/Occurrences HCGSECG 6.1 Gaseous Effluent Release Rev.00 Page 3 of4

Initiating Condition Boundary Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity Exceeds 100 mrem Total Effective Dose Equivalent (TEDE) or 500 mRem Thyroid CDE Dose for the actual or projected duration of the release ) .*

OPCON ( All ) ( All ) ( All ) ( All ) EAL## 6.1.3.a 6.1.3.b 6.1.3.c 6.1.3.d Field . Dose IF Measured IF Field Survey IF Alarm IF Assessment n - a g,,ta Analvsis Indications E M Dose Assessment Dose Rate measured Valid High Alarm received from ANY one of the Analysis of field E indicates EITHER at the Protected following Plant Effiuent RMS Channels: survey samples at the R one of the following Area Boundary Protected Area Boundary

FRVS Noble Gas (Grid 1/3; 9RX680)

G at the MEA or beyond or beyond EXCEEDS indicates EITHER as calculated on 100 mRem/hr E the SSCL: one of the following:

NPV Noble Gas (Grid 1/3; 9RX590)

N * ~ 4.36E+o2 CCPM AND

SPV Noble Gas (Grid 1/3; 9RX580) c
TEDE 4-Day Dose y ~ 1.0E-+-02 mRem * ~ 3.SSE-07 pCi/cc 1-ljl
HTV Noble Gas (Grid 3; 9RX516)
Thyroid-COE Dose Release is expected to continue for ~ 15 minutes A ~ 5.0E-+-02 mRem c based on Plant Vent effiuent sample analysis and NOT on I AND T

a default Noble Gas to Total Plant Vent release rate EXCEEDS I Iodine Ratio 7.6E+o7 pCi/sec Total Noble Gas 0

                                                                                                               ~---------,----------*--**-

N

                                                                                                                                           !AND L                                                                                                                   Dose Assessment results NOT available E

v I AND E I Release is ongoing for::::, 15 minutes I L I s l THEN Action Refer to Attachment 3 Required SITE AREA EMERGENCY

HCGSECG 6.0 Radiological Releases/Occurrences Rev. 00 Page 4 of4 6.1 Gaseous Effluent Release Initiating Boundary Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity Exceeds 1000 mrem Total Effective Dose Equivalent (TEDE) Condition or 5000 mRem Thyroid COE Dose for the actual or projected duration of the release OPCON (~ (_~ ( All ) ( All ) EAL# 6.1.4.a 6.1.4.b 6.1.4.c 6.1.4.d Field Dose IF Measured IF Field Survey IF Alarm IF Assessmen Dose Rate Analvs1s Indications E M Dose Assessment Dose Rate measured Valid High Alarm received from ANY one of the Analysis of field E indicates EITHER at the Protected following Plant Effiuent RMS Channels: survey samples at the R one of the following Arca Boundary Protected Area Boundary

FRVS Noble Gas (Grid 113; 9RX680) at the MEA or beyond or beyond EXCEEDS G indicates EITHER as calculated on 1000 mRcm/hr E the SSCL:

one of the following:

NPV Noble Gas (Grid 113; 9RX590)

N * ~4.36E+o3 CCPM c 9 TEDE 4-Day Dose AND

SPV Noble Gas (Grid 1/3; 9RX580) y ~ 1.0E+oJ mRem
_::: 3.85E-06 JLCi/cc 1-131
HTV Noble Gas (Grid 3; 9RX516) 9 Thyroid-COE Dose Release is expected to continue for _::: 15 minutes A ~ 5.0E+oJ mRem c based on Plant Vent effiuent AND T sample analysis and NOT on a default Noble Gas to Total Plant Vent release rate EXCEEDS I I Iodine Ratio 0 ~~~~7-.6_E_+o~8-"~c_u_se_c~T-o-ta_i_N_o_b_le_G~as~~~-*

N AND L ~~~D~os_e_A_s_se_s_s_m_e_nt_re,s_u_lts_N~O_T_a_v_ai_la_b_le_.~~J E v AND E I Release is ongoing for~ 15 minutes L s Action l Refer to Attachment 4 THEN Required GENERAL EMERGENCY*

HCGSECG 6.0 Radiological Releases/Occurrences Rev.00 Page I ofl 6.2 Liquid Effluent Release Any Unplanned Release of Liquid Radioactivity Any Unplanned Release of Liquid Radioactivity Initiating to the Environment that exceeds 200 Times Radiological to the Environment that Exceeds 2 Times the Radiological Condition Technical Specifications for 60 minutes or longer Technical Specifications for 15 minutes or longer OPCON QD ( All ) EAL# 6.2.1 6.2.2 IF E M Valid Cooling Tower Blowdown Effluent E Radiation Monitor High Alarm Condition R G E N c y l THE.: A AND c AND T Sample analysis of liquid effluent indicates Sample analysis of liquid effluent indicates I concentration in excess of 2 times concentration in excess of 200 times 0 Technical Specification limits Technical Specification limits N L AND AND E v Release continues for ~ 60 minutes Release continues for~ 15 minutes E after the alarm occurs after the alarm occurs L s THEN THEN

                                                                                                                                                ~-J Action                     Refer to Attachment 1                                                       Refer to Attachment 2 Required                      UNUSUAL EVENT                                                                       ALERT

HCGSECG 6.0 Radiological Releases/Occurrences Rev. 00 Page I of! 6.3 In-Plant Radiation Occurrences Initiating Release of Radioactive Material or increases in Radiation Levels within the Unplanned Increase in Plant Radiation facility that impedes operation of systems required to maintain Condition safe o rations or to establish or maintain Cold Shutdown OPCON c All ) ( All ) c All ) EAL# 6.3.1 6.3.2.a 6.3.2.b IF IF IF E M Unplanned rise in Unplanned Dose Rates Unplanned Dose Rates E radiation levels inside the ~ 2000 mRem/hr ~ 15 mRem/hr R Protected Area in ANY area of the plant in EITHER one of the following: G ~ 1000 times normal which requires ACCESS to E

Main Control Room as indicated by EITHER maintain plant safety N one of the following: functions (EXCLUDING the
Security Central Alarm Station (CAS) c Main Control Room and CAS) y
Permanent or portable Area Radiation Monitors - as indicated by EITHER one of the following:

A

General Area Radiological Survey
Permanent or portable Area c Radiation Monitors T THEN I
General Area Radiological Survey 0

N L E v E L THEN s Action Refer to Attachment l Refer to Attachm~ Required UNUSUAL EVENT ALERT _ _ J

HCGSECG 6.0 Radiological Releases/Occurrences Rev. 00 Page I of2 6.4 Irradiated Fuel Event Initiating Condition ~~~~~~-U-np_l_an_n_ed~In_c_rea_~~in_P_l_an_t_Ra~di-au_*o_n~~~~-~ OPCON ( 5 ) ( All ) EAL# 6.4.1.a 6.4.1.b IF IF E M Uncontrolled water level Uncontrolled water level E drop in the drop in the R Reactor Cavity as Spent Fuel Pool G indicated by EITHER as indicated by E one of the following: Valid Fuel Pool Low N Level Alarm Condition c

Visual Observation y
Reactor Water Level Shutdown Range Level AND A Indicator 1BBLI-R605 c Visual Observation T

I 0 N L THEN E v E L s Action Refer to Attachment 1 Required UNUSUAL EVENT

HCGSECG 6.0 Radiological Releases/Occurrences Rev. 00 Page2 of2 6.4 Irradiated Fuel Event Initiating Condition Major Damage to Irradiated Fuel ) Events that have or may result in uncovering Irradiated Fuel outside the Reactor Vessel OPCON c All ) c All ) ( All ) EAL# 6.4.2.a 6.4.2.b 6.4.2.c IF IF IF E M Unplanned rise Major Damage to Visual observation of E on ANY one of the Irradiated Fuel Irradiated Fuel uncovered R following has occurred G Area Rad Monitors or E I AND by general area rad survey N indicates c Valid High Alarm ::::, 2000 mRem/hr: y received from

Spent Fuel Storage Pool ANY one of the A following Area (9RX707) c RMS channels:
New Fuel Criticality T Storage Channel A
Refuel Floor Exhaust I {9RX612) 0 Channel A {9RX627)

N

Refuel Floor Exhaust
New Fuel Criticality Channel B {9RX628) Storage Channel B L {9RX613)
Refuel Floor Exhaust E

v Channel C {9RX629) E L I s THEN

                                                                 *Ir Action Refer to Attachment 2 Required ALERT

    . 1.0  Fuel Clad Challenge 1.1      RCS Activity 2.0  RCS Challenge 2.1      RCS Leakage 3.0  Fission Product Barriers (Table) 4.0  E C Discretion 4 .1     Emergency Coordinator Discretion 5.0  Failure to SCRAM 5.1      ATWS 6.0  Radiological Releases/Occurrences 6.1       Gaseous Effiuent Release 6.2      Liquid Effiuent Release 6.3      In Plant Radiation Occurrences 6.4       Irradiated Fuel Event 7 .0 Electrical Power 7 .1     Loss of AC Power Capabilities 7 .2     Loss of DC Power Capabilities 8.0  System  Malfunctions 8.1       Loss of Heat Removal Capability 8.2       Loss of Overhead Annunciators 8.3       Loss of Communication Capability 8.4       Control Room Evacuation

.i.. 8.5 Technical Specifications 9.0 Hazards - Internal/External

9. I Security Threats 9.2 Fire 9 .3 Explosion 9.4 Toxic/Flamable Gases 9 .5 Seismic Event 9.6 High Winds

9. 7 Flooding 9.8 Turbine Failure/Vehicle Crash/Missile Impact 9.9 River Level

HCGSECG 7.0 Electrical Power Rev.00 Page I of2 7.1 Loss of AC Power Capabilities Loss of All OITsite Power AC power capability to Vital Buses reduced to a Single Power ') /" Loss of All Offsite Power and All Onsite AC --...,) Initiating to Vital Buses for greater Source for greater than 15 minutes such that any additional single ( Power to 4.16 KV Vital Buses during either in Cold Condition than 15 minutes ~ilure would result in a complete loss of all 4.16 KV Vital Busey, ~hutdown or Refueling for greater than 15 minute§.

          ~--------./

OPCON ( I, 2, 3 ) ( 4, 5, Defueled) EAL# 7.1.1 7.1.2.a 7.1.2.b IF IF IF E .-----*-------------------- M Unpl~nned Loss of Loss of 4.16 KV Vital Bus Power Sources ALL 4.16 KV Vital Buses E Power from Station (Offsite and Onsite) which results in the are deenergized R Service Transformers availability of ONLY G IAXSOI AND IBXSOI one 4.16 KV Vital Bus Power Source E to ALL (Offsite or Onsite) N 4.16 KV Vital Buses c y AND A > 15 monutes c have elapsed AND T 0 I THEN C > 15 minutes have elapsed I N THEN L E v E L s Action Required Refer to Attachment I UNUSUAL EVENT [ Refer to Attachment 2 ALERT J

HCGSECG 7.0 Electrical Power Rev. 00 - Page 2 of2 7.1 Loss of AC Power Capabilities Initiating ( Loss of All Offsile Power and Alfofisite~(C Power to*) All Vital AC Buses during either Power OpCration, r-----

Prolonged Loss of All Offsite and Onsite AC Power to All Vital AC Buses

                                                                                                                                               --------~--
                                                                                                                                                                 )

Condition Startup or Hot Shutdown for greater than 15 minute~/ '--- ----------------~ OPCON ill~_) (1~~ ( 1, 2, 3 ) EAL# 7.1.3 7.1.4.a 7.1.4.b IF E M ALL 4.16 KV Vital Buses E are deenergized R G E N c y A AND AND Restoration of Power to __ JAND_ c Loss of any 2 T at least one

                         > 15 minutes have elapsed                                                                             Fission Product Barriers I                                                                         4.16 KV Vital Bus has occurred or is Imminent 0                                     THEN                            within 4 hours is NOT likely
                                                                                                                     ~----------------*-~----*-*-.

N THEN THEN L

                                                                                        '------------~------------

E v E L s ________ _J

                                                                                               ----------------* - . - *--~*--- --- ---

Action Refer to Attachment 4 Required GENERAL EMERGENCY

HCGS ECG 7.0 Electrical Power Rev.00 Page I of I 7.2 Loss of DC Power Capabilities

 ~-------------------~

Unplanned Loss of All Vital 125 VDC Power during either

Unplanned Loss of All Vital 125 VDC Power during either Power\

( Cold Shutdown or Refueling Mode for greater than 15 minutes ( . Operation, Startup or Hot Shutdown for greater than 15 Minutes) ( 4, 5, Defueled ) ( 1, 2, 3 ) 7.2.1 7.2.3 IF IF

                                                                ~----------------------*                       *-*-*-------

Unplanned degraded voltage condition for Unplanned degraded voltage condition for ALL Vital 125 VOC Buses, ALL Vital 125 VOC Buses, such that voltage is < 108 VOC such that voltage is< 108 voe AND AND

                   > 15 minutes have elapsed L_______--,--------
                                                                                       > 15 minutes have elapsed                      J THEN                                                               THEN Ir                                                                 ,.

j ---------------* L ---- -------~-- Refer to Attachment I UNUSUAL EVENT Refer to Attachment 3 SITE AREA EMERGENCY

1.0 Fuel Clad Challenge 1.1 RCS Activity 2.0 RCS Challenge 2.1 RCS Leakage 3.0 Fission Product Barriers (Table) 4.0 E C Discretion 4.1 Emergency Coordinator Discretion ' 5.0 Failure to SCRAM 5.1 ATWS 6.0 Radiological Releases/Occurrences 6.1 Gaseous Effiuent Release 6.2 Liquid Effiuent Release 6.3 In Plant Radiation Occurrences 6.4 Irradiated Fuel Event 7.0 Electrical Power 7.1 Loss of AC Power Capabilities 7.2 Loss of DC Power Capabilities 8.0 System Malfunctions 8.1 Loss of Heat Removal Capability 8.2 Loss of Overhead Annunciators 8.3 Loss of Communication Capability 8.4 Control Room Evacuation ...... 8.5 Technical Specifications 9.0 Hazards - Internal/External 9.1 Security Threats 9.2 Fire 9.3 Explosion 9.4 Toxic/Flamable Gases 9.5 Seismic Event 9.6 High Winds 9.7 Flooding 9.8 Turbine Failure/Vehicle Crash/Missile Impact 9.9 River Level .oil..

HCGS ECG 8.0 System Malfunctions Rev..00 Page I of I 8.1 Loss of Heat Removal Capabililty Loss of Reactor Water Level Initiating Inability to Maintain the Plant that has or will Uncover Fuel Complete Loss of Fwictions Needed Condition in Cold Shutdown to Achieve Cold Shutdown Conditions in the Reactor Vessel OPCON c~ c 4, 5 ) Qil) EAL# 8.1.2 8.1.3.a 8.1.3.b IF IF IF E M Unplanned, Complete Loss of Reactor Water Level Loss of Main Condenser capabilities, as evidenced by E ALL Technical Specification required REACHES -161" an inability to remove Decay Heat from the Reactor R systems available to provide (Top of Active Fuel) G Decay Heat Removal functions AND E N I c y I AND I AND Loss of Torus capabilities as evidenced by EITHER one of the following: An RCS Temperature *Entry into an Unsafe region of ANY one of the A UNCONTROLLED has risen to c > 200°F temperature following curves: T nse *Heat Capacity Temperature Limit (HCTL) Curve (Excluding a <15 *Heat Capacity Level Limit (HCLL) Curve I is RAPIDLY min111tes rise

Pressure Suppression Pressure (PSP) Curve 0 approaching

                >200°F

SRV Tailpipe Level Limit Curve N 200°F with a I.

(with NO heat j L heat removal removal function

Insufficient SRV capacity to reduc*e RPV press~r:~

function restored) E restored) .... *'i v

                                                                                                                                                          ~ ~'\

E I I L THEN s THEN c:.:.~

                                    ,,                                                                                 ,,                                             **y*1 Action Required Refer to Attachment 2 ALERT                         ]                                     .__

Refer to Attachment 3 SITE AREA EMERGENCY *--*** I-~***-*-~**- *----

                                                                                                                                                                       =t::\

HCGSECG Rev. 00 8.0 System Malfunctions 0 Page I of I 8.2 Loss of Overhead Annunciators Unplanned Loss of Most or All Unplanned Loss of Most or All Control Room Inability to Monitor a Significant Transient in Initiating Annunciators and a Significant Transient is in Progress or Compensatory Annunciation or Indication in the Control ( Progress Condition Room for Greater Than IS minutes Indicators are Unavailable OPCON ( 1, 2, 3 ) c 1, 2, 3 ) 8.2.2.a ( 1, 2, 3 ) 8.2.2.b ( 1, 2, 3 ) 8.2.3 EAL# 8.2.1 IF IF E M Unplanned loss of> 75% Loss of> 75% of Main Control Room] E of Main Control Room Overhead Annunciators **

                                                                                                                                        - - - - - - - - i - - - - - - * - --

R Overhead Annunciators 1 AND G E AND I AND A significant transient** is in progres~l N c y FAND

                                                         ~-----*--L----

A significant transient** is in progress

CRIDS BOTH of the following:

I AND BOTH of the following:

CRIDS A
SPDS
SPDS c 5 minutes are NOT are NOT AVAILABLE T ave elapsed AVAILABLE I AND I ce the loss of 0 OHAs lTHEN
I Main Control Room Indications are NOT available to monitor ANY one of the N following:

I AND

RCS Status L

15 minutes have

Reactivity Control E

THEN elapsed since the loss v ofOHAs

ECCS '

E

Containment Parameters L lTHEN *-

s ~THEN Action Refer to Attachment 1 Refer to Attachment 2 Refer to Attachment 3 Required UNUSUAL EVENT ALERT SITE AREA EMERGENCY

           ** NOTE: A Significant Transient is based on EC judgment, but includes as a minimum ANY one of the following:

RX SCRAM, LOAD REJECTION >25% POWER, ECCS INJECTION, THERMAL POWER OSCILLATION >10%.

HCGSECG Rev. 00 8.0 System Malfunctions Page I of I 8.3 Loss of Communications Capabililty Initiating Condition ( Unplanned Loss of All Onsite or Offsite Communications Capabilities OPCON ( All ) ( All ) 8.3.1.a 8.3.1.b EAL# IF IF E M Unplanned loss of Unplanned loss of ALLONSITE ALLOFFSITE E communications communications R as evidenced by the loss of as evidenced by the loss of G ALL of the following ALL of the following E systems: systems: N

Direct Inward Dial System
Direct Inward Dial System c (DID) (DID) y
Nuclear Emergency
Station Page System Telephone System (NETS)

(Gaitronics) A

Essx System (Centrex)
Station Radio System c phone T

I 0 N L E v E THEN L s *Ir Action Refer to Attachment I Required UNUSUAL EVENT

HCGSECG 8.0 System Malfunctions Rev.00 Page I of! 8.4 Loss of Control Room Habitabililty Initiating Main Control Room Evacuation ) Main Control Room Evacuation has been Initiated Condition has been Initiated and Plant Control cannot be established OPCON ( All ) (...___All_) EAL# 8.4.2 8.4.3 IF E M Main Control Room Evacuation E has been initiated R G E N AND c y Control of the plant CANNOT be established A from outside the Main Control Room c within 15 minutes T I THEN 0 N L E v E L s Action Required

                    --~-----

r to Attachment 2 ALERT J [_. Refer to Attachment 3 SITE AREA EMERGENCY

                                                                                              ***--- l

HCGS ECG 8.0 System Malfunctions Rev.00 Page I of I 8.5 Technical Specifications Initiating Inability to Reach Required Operational Condition Condition within Technical Specification Limits OPCON ( 1, 2, 3 ) EAL# 8.5.1 IF E M Plant is NOT brought to the REQUIRED E Operational Condition within the R Technical Specification G required time limit E N c THEN y A c T I 0 N L E v E L s Action Refer to Attachment 1 Required UNUSUAL EVENT

1.0 Fuel Clad Challenge 1.1 RCS Activity 2.0 RCS Challenge 2.1 RCS Leakage 3.0 Fission Product Barriers (Table) 4.0 E C Discretion .,, 4.1 Emergency Coordinator Discretion 5.0 Failure to SCRAM 5.1 ATWS 6.0 Radiological Releases/Occurrences 6.1 Gaseous Efiluent Release ' 6.2 Liquid Effluent Release 6.3 In Plant Radiation Occurrences 6.4 Irradiated Fuel Event 7.0 Electrical Power 7.1 Loss of AC Power Capabilities 7.2 Loss of DC Power Capabilities l 8.0 System Malfunctions 8.1 Loss of Heat Removal Capability 8.2 Loss of Overhead Annunciators 8.3 Loss of Communication Capability 8.4 Control Room Evacuation 8.5 Technical Specifications 9.0 Hazards - Internal/External 9.1 Security Threats 9.2 Fire 9.3 Explosion 9.4 Toxic/Flamable Gases 9.5 Seismic Event 9.6 High Winds 9.7 Flooding 9.8 Turbine Failure/Vehicle Crash/Missile Impact 9.9 River Level

HCGSECG 9.0 Hazards - Internal/External Rev.00 Page I of I 9.1 Security Threats Initiating ,.-- Confirmed

                      .          Security
                                     . Event Which
                                                  . .      --... )   ( - Sccun*ty Even t m  .~-)

a ( Security Event in a ) ,,--- Security Event Resulting *) ( in Loss of Ability to Reach and ( Indicates a Potential Degradation m the Plant Protected Arca Plant Vital Area Condition Level of Safety of the Plant / ._ _ _ _ _ _ _______,,* *- - - - - - - - - - - - - ,__ Maintain Cold Shutdown _../ OPCON (_All ) All ( All ) ( , ____ All __.) EAL# 9.1.1 9.1.2 9.1.3 9.1.4 IF IF IF *IF E ~--------*--**-- *-- M Confirmed security threat directed Confirmed hostile Confirmed hostile intrusion or E toward the station as evidenced by ANY malicious acts in Plant Vital Areas Security event resulting in the intrusion or malicious acts actual loss of physical control R one of the following: as evidenced by ANY as evidenced by : of EITHER one of the following: G

Credible threat of malicious acts or one of the following:
Discovery of an intruder(s),

armed and violent, within the

Main Control Room E destructive device within the Protected
Discovery of an Arca, resulting in SCP-5 implementation intruder(s), armed and Plant Vital Area, resulting in
Remote Shutdown Panel N

SCP-6 implementation c

Credible intrusion or assault threat violent, within the Protected Area, *
Malicious acts or destructive y to the Protected Arca, resulting in *------~*----

SCP-5 implementation resulting in device discovered in a Plant Vital THEN

Attempted intrusion or assault to the SCP-6 implementation Area, resulting in SCP-10 A
Hostage held implementation Protected Area, resulting in SCP-7 or c SCP- I I implementation on-site in a non-vital r-*****--* . .

T I

Malicious acts attempted or discovered area, resulting in SCP-8 implementation THEN t ("

within the Protected Area, resulting in 0 SCP-JO implementation \ N L

Hostage/Extortion situation that threatens normal plant operations, resulting in SCP-8 implementation THEN &.\

E

Destructive Device discovered within v the Protected Area, resulting in E SCP-10 implementation L

__s__i THEN

                                                                                       -=1 Action                    Refer to Attachment 24                                  Attachment 2                      Refer to Attachment 3                        fer to Attachment 4 Required          UNUSUAL EVENT (Common Site)                                       LERT                        SITE AREA EMERGENCY                               ERAL EMERGENCY
                                                                       ----------~--**------ --*-*-***- ...  ---------------------                         . '          -- --* -** . ****- -  .. - - ... --

HCGSECG 9.0 Hazards - Internal/External Rev. 00 Page I of I 9.2 Fire

               ~--------------------~

Initiating , Fire within the Protected Area Boundary ) Fire Affecting the OJ_>erability ~f P.lant Safety Syste~s Condition ( ( . Required to Establish or Mamtam Safe Shutdown

            -, ______N_o_t_E_x_:ti_n_gu_i_sh_e_d_w_1_.th_i_n_1_5_m_i_n_u_te_s_o_f_D_et_e_ct_io_n_ ___,,/

OPCON ( All ) c__ All ( All ') EAL# 9.2.1 9.2.1 9.2.2 IF IF IF E ----- -- M Valid Fire Alarm is received Report of a fire from Fire within ANY one of the following Plant Vital Structures: E in the Main Control Room personnel at the scene R I I

Reactor Building G I AND
Control/Aux Building E Fire is within ANY one of the following Plant Structures (EXCLUDING small fires that have NO potential to affect
Service Water Intake Structure N

Safety Systems or Protected Area Permanent Plant Structures) c

Service/Rad Waste Building y
Reactor Building IAND
Turbine Building A

The Fire is of a magnitude that it SPECIFICALLY c

Control/Aux Building results in Damage to ANY one of the following:

T

Service Water Intake Structure
TWO OR MORE subsystems of a Safety System I

0

Service/Rad Waste Building
MORE THAN ONE Safety System N
Low Level Radwaste Interim Storage Facility
Any Plant Vital Structure which renders the structure incapable of L I AND perfonning its Design Function E Fire is NOT extinguished within 15 minutes of v

E L EITHER one of the following:

Receipt of a Valid Fire Alann JAND Damaged Safety System(s) or Plant Vital Structure----]

is required for the present Operational Condition ___

Report of a fire from the scene s

                                                          ~     THEN                                                                     .         lTHEN
                                ~to Attachm~l l

Action Refer to Attachment 2 ] Required l_____~SUAL EV~ ------ ALERT

HCGSECG 9.0 Hazards - Internal/External Rev.00 Page I of I 9.3 Explosion

               -----------*------..                                 ----=-----=-----:---:--=:--:---;--;:::-----;-:-;::--;:-;:;;--::-:~~~.

(~* Explosion Affecting the Operability of Plant .) Initiating Condition Natural and Destructive Phenomena Affecting the Protected Arca ) ( '"---- Safety Systems Required to Establish or Maintain Safe Shutdown OPCON ( All ) ( ~-----/ AJI J EAL# 9.3.1 9.3.2 IF IF E -----------*- ---- M R G E [---*-- Confirmed within Ex11losion the Protected Area J Confirmed Explosion within ANY one of the following Plant Vital Structures:

Reactor Building E I AND
Control/Aux Building N
Service Water Intake Structure c Report of visible damage to Plant equipment or Protected Arca y
Service/Rad Waste Building Permanent Plant Structures A

IAND THEN c The Explosion is of a magnitude that it SPECIFICALLY T results in Damage to ANY one of the following: I

TWO OR MORE subsystems of a Safety System 0

N

MORE THAN ONE Safety System
Any Plant Vital Structure which renders the structure L

incapable of performing its Design Function E v IAND E Damaged Safety System(s) or Plant Vital Structure is L required for the present Operational Condition s iTHEN Action Refer to AttachmenUJ Refer to Attachment 2 Required UNUSUAL EVENT ALERT

               -- ~------*--*~*--*--*---

L____. - - - - - - - - - - - - - - - - - -

HCGSECG 9.0 Hazards - Internal/External Rev. 00 Page I of 2 9.4 Toxic/Flammable Gases Initiating Condition l~ ______ R_e_le_a_se_o_f_Toxic or Flammable Gases Deemed Detrimental to Safe Operation of the Plant OPCON ( All ) ( All ) ( AJI ) EAL# 9.4.1.a 9.4.1.b 9.4.1.c IF IF IF E M Notification by Local, County, or Uncontrolled Toxic Gas Uncontrolled Flammable Gas E State Officials for the potential need release within the Protected Area release within the R to EVACUATE in ANY area which Protected Area G non-essential personnel that RES UL TS in does not normally require an E due to an Flammable Gas concentrations atmospheric survey N Offsite Toxic Gas release EXCEEDING or Respiratory Protection for c entry 25% of the LEL y AND A SNSS deems evacuation I I c of non-essential personnel AND T is required I L_-------,..------- 0 Routine Plant Operations are IMPEDED based N on EITHER one of the following:

Access restrictions caused by the uncontrolled release L
Personnel injuries have occurred as a result of the release E

v E L THEN THEN s Refer to Attachment 24 l Refer to Attachment 1*-* .. *--- -1 Action UN_USUAL EVENT (Common ~~j UNUSUAL EVENT Required

HCGSECG 9.0 Hazards - Internal/External Rev. 00 Page 2 of 2 9.4 Toxic/Flammable Gases Initiating Release of Toxic or Flammable Gases Within a Facility Structure Which Jeopardizes Operation of Systems ~) Condition Required to Maintain Safe Operations or to Establish or Maintain Cold Shutdown Conditions

../

OPCON ( All ) ( All ) EAL# 9.4.2.a 9.4.2.b IF IF E M Uncontrolled Toxic Gas release within ANY one Uncontrolled Flammable Gas release within ANY one E of the following Plant Structures of the following Plant Structures R

Reactor Building
Reactor Building G
Turbine Building
Turbine Building E
Control/Aux Building
Control/Aux Building N
Service Water Intake Structure
Service Water Intake Structure c
Service/Rad Waste Building y
Service/Rad Waste Building I AND AND A Toxic Gas concentrations result in ANY one c of the following: Flammable Gas concentrations EXCEED T
An IDLH atmosphere 50% of the LEL I
Plant personnel report severe adverse health reactions, 0

including burning eyes, nose, throat, dizziness N

The Threshold Limit Value (TLV) being EXCEEDED L I E

v I AND E Plant personnel are unable to perform actions necessary to complete a Safe L Shutdown of the plant without appropriate personnel protection equipment s l THEN Action 1 - - R ; f e r to Attachment 2 Required ALERT

HCGSECG 9.0 Hazards - Internal/External Rev.00 Page I of I 9.5 Seismic Event Initiating ( Natural and Destructive Phenomena ) Natural and Destructive Phenomena Condition ___________ Affi_ec_t_in_g_t_he_P_ro_t_e_c_t_ed_Ar_e_a_ _ _ _ __ Affecting the Plant Vital Area OPCON ( All ) (_~_ _) ( All ) EAL# 9.5.1.a 9.5.1.b IF IF E M Seismic Event felt Valid Actuation of the E by personnel Seismic Trigger(> O.Olg) R within the has occurred as verified by the G Protected Area SMA-3 Event Indicator (flag) E being WHITE

           ---~---

N on Panel 10-C-673 in the c Upper Relay Room y A c Valid Actuation of the Seismic Switch(> O.lg) T has occurred as verified by EITHER one I of the following: 0 AND N r-------~-------

Valid actuation of Main Control Room

                                                            ~                                     Overhead Annunciator C6-C4 L

E l

AMBER Alarm light on the Seismic Switch Power Supply Drawer is lit on Panel 10-C-673 v in the Upper Relay Room E

L s Action Refer to Attachment 24 ~ Refer to Attachment 2 Required '--* UNUSUAL EVENT (Common Site)

                                                          ---------~

L______- - - - - ALERT

HCGSECG 9.0 Hazards - Internal/External Rev. 00 Page I of I 9.6 High Winds Natural and Destructive Phenomena Natural and Destructive Phenomena ----...\ ' Affecting the Protected Area ________________________________/ J Affecting the Plant Vital Area ( All ) <~~-) ( All

9.6.1.a 9.6.1.b IF IF Report of a Sustained wind speeds Tornado > 75 MPH TOUCHING for 15 minutes, DOWN measured at within the ANYelevati on of Protected Area the Met To wer The Wind Speed is of a magnitude that it SPECIFICALLY results in Damage to ANY one of the following: I I

TWO OR MORE subsystems of a Safety System
MORE THAN ONE Safety System

                         > - - - - - - - - -*

Rendering ANY of the following structures incapable of performing its Design Function:
Reactor Building
Control/Aux Building
Service Water Intake Structure
Service/Radwaste Building I AND Damaged Safety System(s) or Plant Vital Structure is required for the present Operational Condition

                        ...                                                                     ~   THEN Refer to Attachment 24                                                  Refer to Attachment 2 UNUSUAL EVENT (Common Site)
    ------------------                                                    .                  ALERT
                                                                                      -------*-            -***---~*-*-*-**

HCGSECG 9.0 Hazards - Internal/External Rev. 00 Page I ofl.

9. 7 Flooding Initiating Condition ( Internal Flooding in Excess of Sump Handling Capability *. )

Affecting Safety Related Areas of the Plant . ( Internal Flooding Affecting the Operabilityof Plant Safety Systems Required to Establish or Maintain Safe Shutdown /

*)

OPCON ( All ) ( All

EAL# 9.7.1 9.7.2 IF IF E *----------- M Visual Observation of Uncontrolled Flooding that confirms Visual Observation of Flooding within ANY one E ANY one of the following: of the following Plant Vital Structures: R

Reactor Building Floor Levels above the Maximum
Reactor Building G

Normal Floor Level (> l ") referenced in EOP 103,

Control/Aux Building E

Secondary Containment Control

Service Water Intake Structure N

c

Receipt of a SSWS Pump Room Flooded Alarm
Service/Rad Waste Building y
Greater than 2" of water in ANY area that contains a Safety System(s), not included above I AND A

c The Flooding is of a magnitude that it SPECIFICALLY T results in Damage to ANY one of the following: I

TWO OR MORE subsystems of a Safety System 0
MORE THAN ONE Safety System N
Any of the above listed Plant Vital Structures which renders L the structure incapable of performing its Design Function E

v I AND E Damaged Safety System(s) or Plant Vital Structure is required forj L the present Operational Condition s Action Required [. Refer to Attachment 1 UNUSUAL EVENT l l THEN Refer to Attachment 2 ALERT

HCGS ECG 9.0 Hazards - Internal/External Rev. 00 Page I of I 9.8 Turbine Failure I Vehicle Crash I Missile Impact Initiating Natural and Destructive Phenomena Natural and Destructive Phenomena Condition Affecting Certain Structures Within the Protected Area Affecting Certain Structures Within the Plant Vital Area OPCON (_All ) ( All ) ( All ) EAL# 9.8.1.a 9.8.1.b 9.8.2 IF IF IF E M Catastrophic damage to the Vehicle Crash I Missile Impact with or within Vehicle Crash I E Main Turbine as evidenced by Missile Impact with or ANY one of the following Plant Vital Structures: R EITHER one of the following: within ANY one of the

Reactor Building G
Main Turbine casing following Plant Structures:
Control/Aux Building E

penetration

Reactor Building N
Service Water Intake Structure
Main Turbine/Generator c
Turbine Building
Service/Rad Waste Building y Damage potentially releasing
Control/Aux Building Lube Oil or Hydrogen Gas
Service Water Intake I AND to the Turbine Building A Structure c The Vehicle Crash I Missile Impact is of a magnitude
Service/Rad Waste Building T that it SPECIFICALLY results in Damage to ANY I one of the following:

0

TWO OR MORE subsystems of a Safety System N
MORE THAN ONE Safety System
Any of the above Plant Vital Structures which renders L the structure incapable of performing its Design Function E THEN v I AND E

L s Damaged Safety System(s) or Plant Vital Structure is required for the present Operational Condition

                                                                                                               ~THEN l

Action Required Refer to Attachment I UNUSUAL EVENT

                                                                                                 ~to Attachment 2 ALERT

HCGSECG 9.0 Hazards - Internal/External Rev.00 Page I of I 9.9 River Level Initiating Natural and Destructive Phenomena Condition Affecting the Protected Area OPCON EAL# c All ) ( All ) 9.9.1.a 9.9.1.b IF IF E M River Level > 99.5' River Level< 80.0' E R G E N c THEN y A c T I 0 N L E v E L s Refer to Attachment 24 Action UNUSUAL EVENT (Common Site) Required

1 leO Reportable Action Levels (RALs) 11.1 Technical Specifications 11.2 Design Basis/ Unanalyzed Condition 11.3 Engineered Safety Features (ESF) 11.4 Personnel Safety/Overexposure 111 11.5 Environmental 11.6 After-the-Fact

11. 7 Security/Emergency Response Capabilities 11.8

Public Interest 11.9
Accidental Criticality/ Special Nuclear Material (SNM)/

Rad Material Shipments - Releases 11.10 Voluntary Notifications

HCGS ECG 11.0 Reportable Action Levels Rev. 00 Page I of 2 I I. I Technical Specifications

                                                                                                                 ,.----ANY oEvi~loN FR~;:;-~~c;R-- *
              ~--------------

Initiating INITIATION OF ANY UNIT SHlJrDOWN ) ,-EXCEEDING ANY TECHNICAL SPECIFICATION-,'.) REQUIRED BY THE TECHNICAL SPEC IFICATIO NS SAFETY LIMIT ( LICENSE CONDITION PURSUANT TO ) ( Condition ( , IOCFR50.54(x) [ IOCFR50.72(b)(l)(i)(B)) [IOCFR50.72(b)(I)(i)(A) ___ __/ '* [IOCFR50.36(c)(l)) ____/ '*- ___________:_:.../ OPCON ( 1,2 ) ( I, 2, 3, 4, 5 (as applicable in T/S)) ( All ) RAL# 11.1.1.a 11.1.1.b 11.1.1.c IF IF IF R E Unit shutdown is Exceeding ANY one p Action required because INITIATED of the following no action consistent with 0 to comply with Technical Specification Safety Limits: Technical Specifications R Technical Specifications T

TIS 2.1.1, THERMAL POWER, or license can provide adequate Low Pressure or Low Flow or equivalent protection in an I

emergency N

TIS 2.1.2, THERMAL POWER, (See NC.NA-AP.ZZ-0005 (Q)

G High Pressure and High Flow for guidance on deviation

T/S 2.1.3, REACTOR COOLANT from written procedures)

A c SYSTEM PRESSURE T

T/S 2.1.4, REACTOR VESSEL I'",.., ._.,,1<1ro'_...---... -- ... ' '
1 I WATER LEVEL 0 c:,;

( . 'j i

                                                              ~-------~--------'

N ::

:L.1 L

E v THEN E L s __J Action Required

HCGSECG 11.0 Reportable Action Levels Rev.00 Page 2 of 2 11.1 Technical Specifications VIOLATION OF THE REQUIRE1vffiNTS___ -"' ___ ) Initiating CONTAINED IN THE OPERATING LICENSE ) Condition [HCGS Operating License, Section~_l!l_____

./

OPCON (~_AI_I_) (_Ali-) (~_Alt~) ( All ) RAL# 11.1.3.a 11.1.3.b 11.1.3.c 11.1.3.d IF IF IF IF R E p Any of the The conductivity, One or more Violation of ANY one of the TIS LCOs chloride concentration snubbers 0 requirements contained in Section 2.C for RCS or pH in the RCS are found to be R (Items 3 through 13) Pressure/ is in excess of its INOPERABLE T of the Operating License EXCEPT Temperature specified limits per and I as otherwise provided in the (TIS 3.4.6.1) TIS 3.4.4 have been replaced N Technical Specifications are exceeded Action Statement C. l or G or Environmental Protection Plan thereby requiring thereby requiring an restored to an A an Engineering Engineering Evaluation OPERABLE status, c THEN Evaluation. an Engineering T Evaluation I shall be performed 0 per T/S 4.7.5.g N L E ' - - - - - - - - - - - - - - - - - - * --------- *-*-* --- - v THEN E L s Action Refer to Attachment 22 Required OTHER Reports

HCGSECG 11.0 Reportable Action Levels Rev.00 Page.I of 2 11.2 Design Basis I Unanalyzed Condition ANY EVENT OR CONDITION DURING OPERATION ANY EVENT FOUND WHILE SHUTDOWN THAT, HAD IT BEEN FOUND WHILE THE Initiating THAT RESULTS IN THE CONDITION OF THE PLANT BEING REACTOR WAS IN OPERATION, WOULD HAVE SERIOUSLY DEGRADED THE PLANT OR Condition SERIOUSLY DEGRADED II OCFR50. 72(b XI )(ii)! RESULTED IN BEING IN AN UNANALYZED CONDITION [I OCFRS0.72(b)(2Xi)I ( 1,2 ) ( 3,4,5, Defueled) OPCON RAL# 11.2.1 11.2.2.a IF IF R E p As judged by the SNSS/EDO, Any event, found while the Reactor is shutdown, 0 an event or condition found that, had it been found during operation, would have resulted in R during plant operations that results the plant, including it principal safety barriers T in ANY one of the following: being in EITHER one of the following conditions: I N

The condition of the plant, including its
Seriously degraded principal safety barriers, being seriously
In an unanalyzed condition that G

degraded. significantly compromises plant safety. A

The plant being in an unanalyzed condition c that significantly compromises plant safety. THEN T

I

The plant being in a condition outside the 0 design basis of the plant.

N

The plant being in a condition not covered by normal/abnormal or emergency L operating procedures.

E v THEN E L s

                                                                                                       *Ir Action                  Refer to Attachment 12 Refer to Attachment 14 Required                       I Hour Report                                                     4 Hour Report

HCGSECG 11.0 Reportable Action Levels Rev.00 Page 2 of 2 11.2 Design Basis I Unanalyzed Condition VENT/CONDmON 1llATALONE COULD HA VE.) PRESENCE OF A LOOSE PART IN Initiating PREVENTED CERTAIN SAFETY FUNCTIONS THE REACTOR COOLANT SYSTEM Condition (IOCFRS0.72 (b)(2) (iii)) [Reg. Guide 1.133 I OPCON ( __ AI_l _) ( __AI_l~) RAL# 11.2.2.b 11.2.2.c IF IF R .----------------*- E p Any event or condition that Presence of a Loose Part in the 0 alone could have prevented the RCS is Confirmed R fulfillment of the safety function T of structures or systems that are I needed to perform N ANY one of the following: G

Control the release of radioactive material A

c

Shutdown the reactor and maintain it T in a safe shutdown condition I
Remove residual heat 0

N

Mitigate the consequences of an accident L

E v E L THEN s Action Refer to Attachment 14 Required [ 4 Hour Report

HCGSECG 11.0 Reportable Action Levels Rev.00 Page I of I 11.3 Engineered Safety Features (ESF) ANY EVENT THAT RES ULTS OR SHOULD HA VE ACTUATION OF ENGINEERED SAFETY FEATURE Initiating Condition RESULTED IN ECCS DISCHARGE INTO THE RCS AS THE RESULT OF A VALID SIGNAL [IOCFR50.72(b)(l)(iv)] (INCLUDING THE REACTOR PROTECTION SYSTEM) EXCEPT PREPLANNED [IOCFR50.72(b)(2)(ii)]

OPCON ( All =) ( All ) RAL# 11.3.1 11.3.2 IF IF R E Valid ECCS Actuation, Manual or Automatic, has or p Any event or condition that results in manual or automatic should have occurred 0 actuation of any Engineered Safety Feature (ESF), except as R AND part of a preplanned sequence during reactor operation or T testing, including the Reactor Protection System (RPS) I ECCS Actuation results or should have resulted in, N discharge to the vessel AND G THEN ESF I RPS Actuation is determined A to be reportable in accordance with c NC. NA-AP .ZZ-OOOO(Q) T I THEN 0 N L E v E L s Action Refer to Attachment 12 Refer to Attachment 14 Required 1 Hour Report 4 Hour Report

                       ~-----------             --

HCGSECG 11.0 Reportable Action Levels Rev.00 Page I ofl 11.4 Personnel Safety I Overexposure Initiating ANY INCIDENT OR EVENT INVOLVING BYPRODUCT, ANY INCIDENT OR EVENT INVOL VINO LOSS OF ONSITE FATALITY Condition (SOURCE, OR SPECIAL NUCLEAR MATERIAL CAUSING CONTROL OF LICENSED MATERIAL CAUSING (10CFR50.72(b)(2)(vi))

           '-- ANY OF THE LISTED RESULTS(IOCFR20.2202(a)] /           ANY OF THE LISTED RESULTS [I OCFR20.2202(b ))

OPCON (~_AI_I_) (....._____ All ~

                                                                                                     )                    (     All RAL#                               11.4.1                                                  11.4.2.a                          11.4.2.b IF                                                       IF                                IF R

E Any fatality has p PERSONNEL OVEREXPOSURE or potential PERSONNEL OVEREXPOSURE or potential for occurred within the for overexposure as indicated by overexposure within a 24 hour period, as indicated Owner Controlled 0 ANY one of the following: by ANY one of the following: Area (OCA) R T

TEDE exposure _:: 25 Rem
TEDE exposure > 5 Rem THEN I

N

LOE exposure > 75 Rem
LOE exposure > 15 Rem G
SDE exposure _:: 250 Rem
SOE exposure > 50 Rem A
Release of radioactive material inside
Release of radioactive material inside c or outside of a Restricted Area so that or outside of a Restricted Area so that T had an individual been present for had an individual been present for I 24 hours the individual could have 24 hours the individual could have 0

received _:: 5 times the occupational received > 1 times the occupational N ALI (Annual Limit oflntake) which ALI (Annual Limit of Intake) which would usually equate to 25 Rem would usually equate to 5 Rem CEDE. L CEDE. This Does NOT apply to areas This Does NOT apply to areas where E where personnel are NOT normally personnel are NOT normally stationed v stationed during routine operations. during routine operations. E L THEN THEN s

                                                                                                                     .-------'-----~

Action Refer to Attachment 12 Refer to Attachment 14 Refer to Attachment 17 Required 1 Hour Report 4 Hour Report 4 Hour Report

JICGS ECG 11.0 Reportable Action Levels Rev. 00 Page 2 of 2 11.4 Personnel Safety I Overexposure

                                                                       --------..__   ,,~-----                              "    ,,,----*-----------*---------------- -.
                                                                                                                                . FITNESS FOR DUTY PROGRAM: FALSE POSITIVE DUE )

Initiating RADIOACTIVELY CONTAMINATED PERSON TRANSPORTED) SIGNIFICANT FITNESS FOR DUTY) FROM THE SITE TO AN OFFSITE MEDICAL FACILITY FOR TO ADMINISTRATIVE ERROR (BLIND TEST BY LAB) EVENTS [IOCFR26.73) ( Condition TREATMENT [IOCFRS0.72(b)(2)(v))

                                                                                   , (.                                                         (IOCFR26, APP. A, 2.8(e)(S)]

OPCON (~_Al_l~) CAI!) c==A11=) RAL# 11.4.2.c 11.4.3.a 11.4.3.b IF IF IF R E Transportation of a radioactively contaminated Any event that is determined to be The occurrence of a false positive error on p or potentially contaminated individual a blind lab performance test specimen reportable by the Medical Review 0 from the site to an offsite medical facility for Officer (MRO) or designee IAW under I OCFR26 as determined by the R treatment PSE&G's Fitness for Duty Program Medical Review Officer (MRO) IAW T (NC.NA-AP.ZZ-0042(Q)) PSE&G's Fitness for duty Program I THEN (NC.NA-AP .ZZ-0042(Q) N AND G T~ND A ________L ___________. c The reportable details of the event The reportable details of the event are made T are made available to the SNSS by available to the SNSS by the MRO I the MRO or designee or designee 0 N THEN THEN L E v E L s

                     -*-*----       -**------*,  -***--***- ---* -* *-- .                    --- --*--**-----~'------,

Action Required [ Refer to Attachment 17 4 Hour Report Refer to Attachment 19 24 Hour Report Refer to Attachment 19 24 Hour Report

11.0 Reportable Action Levels HCGSECG Rev.00 11.5 Environmental Pagel of l SPILUDISCHARGE OF ANY NON- SPILUDISCHARGE OF ANY NON-RADIOACTIVE UNUSUAL OR IMPORTANT Initiating RADIOACTIVE HAZARDOUS SUBSTANCE HAZARDOUS SUBSTANCE INTO OR UPON TIIE ENVIRONMENTAL EVENTS Condition [IOCFRS0.72(bX2Xvi); N.J.A.C. 7:1E) RIVER [IOCFRS0.72(bX2) (vi); N.J.A.C.7:1E] [E.P.P. SECTION 4.1] OPCON ( All ) ( All ) ( All ) RAL# 11.5.2.a 11.5.2.b 11.5.2.c IF IF IF R E Spill/discharge of an industrial chemical or EITHER one of the following events occur: As judged by the SNSS/EDO, ANY one of p petroleum product outside of a plant the following events has occurred:

Observation of a spill/discharge of an industrial 0 structure within the Owner Controlled
Unusually large fish kill Area (OCA) that results in EITIIER one of chemical or petroleum product from on-site into R the Delaware River or into a storm drain the following:
Protected aquatic species impinge on T
Spill / discharge that has passed through
Observation of an oil slick on the Delaware River Circulating or Service Water intake I

the engineered fill and into the ground from any source screens (eg.; sea turtle, sturgeon) as N reported by Site personnel G water as confirmed by licensing

Spill / discharge that CANNOT be THEN
Any occurrence of an unusual or A cleaned up within 1 hour and no contact important event that indicates or could result in significant environmental c with groundwater is suspected impact casually related to plant T operation; such as the following:

I THEN

Onsite plant or animal disease 0 outbreaks N
Mortality or unusual occurrence of Note: any species protected by the This event MAY require IMMEDIATE Endangered Species Act of 1973 L ..... ~ (15 minute) notifications. DO NOT <II ...
Increase in nuisance organisms or E delay implementation of Attachment conditions v 16.
Excessive bird impactation E I
NJPDES Permit violations L
Excessive Opacity (smoke) s lTHEN Action Refer to Attachment 16 Refer to Attachment 16 Refer to Attachment 15 Required Spill/Discharge Reporting Spill/Discharge Reporting Environmental Protection Plan

HCGSECG 11.0 Reportable Action Levels Rev.00 Page I.oft 11.6 After-the-Fact Initiating (~ EMERGENCY CONDITIONS DISCOVERED Condition AFfER-THE-FACT OPCON ( All ) RAL# 11.6.1 IF R E p Discovery of events or conditions that had previously occurred 0 (event was NOT ongoing at the time of discovery) R which EXCEEDED an Emergency Action Level (EAL) T and was NOT declared as an emergency I N G AND A There are currently NO adverse consequences c in progress as a result of the event T I THEN 0 N L E v E L s Action Refer to Attachment 12 Required l Hour Report

                   ~---------------'

HCGS ECG 11.0 Reportable Action Levels Rev. 00 Page I of I

11. 7 Security I Emergency Response Capabilities

           *,..--SAFEGUARDS EVENTS THAT ARE DETERMINED~"

Initiating MAJOR LOSS OF EMERGENCY ASSESSMENT CAPABILITY. OFFSITE RESPONSE * ') Condition (TO BE NON-EMERGENCIES, BUT ARE REPORT ABLE)

           <:ro

( CAPABILITY, OR COMMUNICATIONS CAPABILITY (lUCFR50.72(b)(l)(v)) /

               - THE NRC WITHIN ONE HOUR I IOCFR73.7l(b)(l)j.

OPCON All ) ( All All RAL# 11.7.1.a 11.7.1.b 11).1.c IF IF IF R r--*-* ------------..*-----------*------- --- . . i r* . - ....... *- * -----* -----*--*---.. --.... __________...... ___ .........-... - _.......--..--------* ------1 E l Any Non-Emergency safeguards event that is : SNSS/EC determines that an event(s) (excluding a scheduled test or preplanned p reportable in accordance with IOCFR 73. 71 as maintenance activity) has occurred that would impair the ability to deal with an 0 determined by Security (SCP-15) accident or emergency as indicated by the Loss of ANY one of the following: R T I THEN

                                                                                             *---=::::=*-~=:--=r= _____ :::::-~--~-- ----=-==~~~--.=-_:T_-

Nuclear Emergency

Use of the TSC for> 8 hrs

                                                                                                                                                                                               -----=-_11 N

Telecommunications System

SPY, NPV, or FR VS vent G

(NETS) for> 1 hr radiation efiluent monitors for A

ENS for> l hr in the Control

                                                                                                                                                       > 8 hrs                                                      I c                                                                                           Room, TSC, or EOF

SPDS OR CRIDS for > 8 hrs T (NIA ifr.:port..:d by th..: NRC).
EROS, NRC phone line, Modem for
More than seven Offsite Sirens for > 1 hr (NIA ifr..:ported by the NRC).

0

                                                                                                 > l hr

More than > 75% OHA's for N

                                                                                                                                                      < 15 minutes

Use of the EOF for> 8 hrs L
Concurrent multiple accident or E
All Meteorological data (Hope emergency condition indicators which in v Creek AND Salem) for> 8 hrs the judgement of the SNSS significantly E impairs assessment capabilities
Site access due to Acts of Nature L (snow, flood, etc.)

s THEN

                                                                                                                        ~ THEN Action                       Refer to Attachment 11                                                     Refer to Attachment 25                                   Refer to Attachment 12 Required                 1 Hour Report (Common Site)                                              I Hour Report (Common Site)                                             1 Hour Report
                                                                                                                                                      ~-------------------

HCGSECG 11.0 Reportable Action Levels Rev.00 Pagelofl. 11.8 Public Interest Initiating UNUSUAL CONDITIONS WARRANTING A NEWS UNUSUAL CONDITIONS DIRECTLY AFFECTING LOWER RELEASE OR NOTIFICATION OF GOVERNMENT Condition ALLOWAYS CREEK TOWNSHIP (LACT) (LAC -MOU] AGENCIES ( IOCFR50.72(b)(2)(vi)) OPCON (~_Al_I_) ( All

                                                                                                  ~---~
                                                                                                             )

RAL# 11.8.2.a 11.8.2.b IF IF R E SNSS/EDO judges that an event or situation has As judged by the SNSS/EDO, events which are the p occurred that is related to ANY one of the following: responsibility of PSE&G which have or may result in 0 EITHER one of the following: R

The health and safety of the public T
Anticipated unusual movement of equipment or
The health and safety of onsite personnel I personnel which may significantly affect N
Protection of the environment local traffic patterns G
Onsite events which involve alarms, sirens or other noise which may be heard off-site A

c AND AND T THEN A news release Notifications to a I is planned Local, State or Federal 0 agency has been or will be N made L I E THEN v E L s IR~ to Attachment Action

14 Refer to Attachment 21 Required ~Hour Report LACT I MOU Report __

HCGSECG 11.0 Reportable Action Levels Rev. 00 Page I of2 11.9 Accidental Criticality I Special Nuclear Material I Rad Material Shipments - Releases UNPLANNED I ACCIDENTAL LOSS AND INVESTIGATION OF TIIE LOSS OF SPECIAL TIIEFT OR LOSS OF Initiating CRITICALITY NUCLEAR MATERIALS/ SPENT FUEL LICENSED MATERIAL Condition [IOCFR70.52(a)) [IOCFR73.27(c), IOCFR73.7l(a)) [IOCFR20.220l(aXIXi))

                                          ---------------------~

OPCON (_ All ___ ) ( All ~D RAL# 11.9.1.a 11.9.1.b 11.9.1.c IF IF if R E Any unplanned or ANY one of the following events occur involving Lost, stolen or missing p accidental criticality Special Nuclear Material (SNM) or Spent Fuel: licensed material > I 000 0 times the quantity specified THEN R in 10CFR20 Appendix C in

Shipment of formula quantities of strategic T such circumstances that an I SNM (SSNM) or Spent Fuel that is lost or exposure could result to N unaccounted for after the estimated tirne of arrival persons in Unrestricted G *A lost or unaccounted for shipment of SSNM or Areas.

Spent Fuel has been recovered or accounted for A THEN

Results of a trace investigation of lost or c

T unaccounted for SSNM shipment are received I THEN 0 N L E v E L s

                       ,,                                            1 Action     Refer to Attachment 12                        Refer to Attachment 11                      Refer to Attachment 11 Required         1 Hour Report                      I Hour Report (Common Site)                  I Hour Report (Common Site)

HCGSECG 11.0 Reportable Action Levels Rev.00 Page 2 ofl 11.9 Accidental Criticality I Special Nuclear Material I Rad Material Shipments - Releases EXCESSIVE CONTAMINATION ACCIDENT DURING TRANSPORT ) CONTAMINATION OUTSIDE OF TIIB Initiating RECEIPT OF SSNM MATERIAL AND/OR RADIATION LEVELS ON A OF LICENSED MATERIAL RADIOLOGICALLY CONTROLLED Condition [IOCFR73.27(b)) PACKAGE [IOCFR20.1906(d)) '- [IOCFR71.5(aXI Xv)] / AREA [IOCFR50.72(bX2Xvi)) OPCON ( __ M_I_) ( All ( __ AI_I_) (_AI_l~J RAL# 11.9.1.d 11.9.1.e 11.9.2.a 11.9.2.b IF IF IF IF R E Receipt of shipment of Receipt suivey indicates that Accidents during the Discovery of a Contaminated p Strategic Special Nuclear Area OUTSIDE of the RCA package contamination/radiation transportation of 0 Material (SSNM) levels equal or exceeds ANY radioactive material with removable activity R one of the following: which are reported to PSE&G I T THEN as the shipper that involve

2200 dpm/100 cm 2 N

I

200 mR/hr on contact (or potentially involve) damage to the cargo I

AND I AND G Location of Location of

10 mR/hr at 3 feet Contaminated Contaminated A THEN Area is Area is such c THEN OUTSIDE of that a T Plant contaminated I Structures person or 0 material may N AND have left the I Protected L Size of Area E Contaminated --

v Area is E LARGE L (>JOO FT2) s I Action Required

          ~er to Attachment L     I Hour Report I0 L__
                                        \Refer to Attachment I 0 I Hour Report Refer to Attachment 18 4 Hour Report I  THEN Refer to Attachment 13 4 Hour Report l

HCGSECG 11.0 Reportable Action Levels Rev.00 Page I of I 11.10 Voluntary Notifications Initiating ~ts/conditions warrant voluntary/courtesy Condition \ NRC notification [IOCFRS0.72 - Voluntary Report] OPCON <~ All ) RAL# 11.10.2 IF R E p In the judgement of the SNSS, notification to the NRC is warranted 0 R T AND I N G NO other EALs or RALs appear to be applicable A THEN c T I 0 N L E v E L s Action I Refer to Attachment 14 Required l_~ 4 Hour Report

                  --~--~~-~~--~--~-~

ECG ATI 1 Pg. 1 of 9 ATTACHMENT 1 UNUSUAL EVENT I. EMERGENCY COORDINATOR (EC) LOG SHEET . ;,*

                                                             ~                              I'
                                                           -------~*-*........
                                                                          -~-

I A. DECLARE AN UNUSUAL EVENT AT HOPE CREEK EC EAL# _ _ _ _ __ Declared at - - - - - hrs on - - - - - - time date B. NOTIFICATIONS ( ) 1. CALL communicators to the Control Room. ( ) 2. COJ\1PLETE the INITIAL CONT ACT MESSAGE FORM (IC:rvtf) (last page of this attachment). ( ) 3. PROVIDE the IC:rvtf to the Communicator (CMI) and DIRECT the CMI to implement Attachment 6. ( ) 4. DIRECT the Secondary Communicator (CM2) to implement Attachment 8 for an Unusual Event. ( ) 5. SOUND the Radiation Alert Alarm and make the following page announcement:

              "Attention all personnel. Attention all personnel."
              "Hope Creek is in an UNUSUAL EVENT condition due to (Repeat)

C. SECURITY RELATED EVENT

1. IF Security Related, EC THEN DIRECT the PSE&G Security Supervisor (x2222) to implement the Security Contingency Plan.

2. a bomb search is required, EC THEN*

              -a.--' DIRECT the OSC Coordinator to;

( ) ACTIVATE the OSC IAW EPIP 202H, OSC Activation and Operations AND ( ) IMPLEMENT Bomb Search Operations IAW Appendix 1. ( ) b. DIRECT the NCOs to check control boards for correct equipment lineups. HCGS Rev. 00

ECG ATTl Pg. 2 of 9 D. EMERGENCY COORDINATOR DUTIES ( ) 1. NOTIFY the Salem SNSS, (NETS 5121; DID 5200) with Event Description. ( ) 2. IF required, IMPLE.MENT Accountability by referring to the Accountability Instructions in Section II.

3. COMPLETE and APPROVE the NRC Data Sheet (Attachment 5) for transmittal EC by the CMl within 60 minutes.

4. WHEN provided by the CM2, EC THEN REVIEW and APPROVE the Station Status Checklist (SSCL) for transmittal.

( ) a. REPEAT this step approximately every half hour. ( ) b. PERFORM immediately for any significant change in emergency status. (operational or radiological) E. TURNOVER IF relieved prior to termination of the Unusual Event, EC THEN DOCUMENT the name of your relief below: Name time F. ESCALATION IF event classification escalates above Unusual Event, EC THEN EXIT this attachment and implement a new attachment as directed by the EALs. G. TERMINATION

1. TERMINATE the UE IAW Section III., Emergency Termination/Reduction EC /Recovery (Pg. 5).

2. ENSURE appropriate reports are made IAW Sectic:>n IV., Reporting, of this SNSS attachment.

HCGS Rev. 00

ECG ATI 1 Pg. 3 of 9 II. ACCOUNTABILITY INSTRUCTIONS FOR THE PROTECTED AREA A. IMPLEMENTATION OF ASSEMBLY AND ACCOUNTABILITY Initials/Time I 1. IF NOT already done, EC THEN DIRECT the OSC Coordinator to activate the OSC IAW EPIP 202H, OSC Activation and Operations. I 2. DIRECT Security (x2222) to IMPLEMENT, EC o EPIP 901, Onsite Security Response and, o EPIP 902, Accountability/Evacuation, Sections 3.1 and 3.2 ONLY, for Assembly and Accountability. (NO Evacuation) I 3. DIRECT the Salem SNSS to implement EPIP IOlS, Appendix 6, Accountability EC Instructions For An Unusual Event at Hope Creek. NOTE Steps A.4 thru A.8 may be delegated by the EC to any available CR Staff member. I 4. SOUND the Radiation Alert Alarm and make the following page announcement:

                    "Attention all personnel. Attention all personnel."
                    "Hope Creek is in an UNUSUAL EVENT condition due to "All PSE&G personnel assemble at your Accountability Stations. All contractors leave the Owner Controlled Area immediately". (Repeat)

I 5. WAIT for 5 minutes for key personnel to reach their Accountability Stations, THEN CONTINUE with Step 6. I 6. SOUND the Radiation Alert Alarm and ANNOUNCE the following; (T= 0 Min)

                    "Attention, Attention. All accountability stations, IMPLEMENT Accountability." (Repeat)

I 7. WHEN 10 minutes have elapsed from Step 6, ANNOUNCE the following; (T+lO Min)

                    "Attention, Attention. All accountability stations, COMPLETE YOUR INITIAL Accountability." (Repeat)

HCGS Rev.00

ECG ATf l Pg. 4 of 9 II. ACCOUNTABILITY INSTRUCTION FOR THE PROTECTED AREA (CONT) Initialsffime I 8. WHEN 20 minutes have elapsed from Step 6, ANNOUNCE the following; (T+20 Min)

                   "Attention, Attention. All accountability stations COMPLETE YOUR 30 MINUTE Accountability." (Repeat)

I 9. WHEN 30 minutes have elapsed from Step 6, EC (T+30 Min) COORDINATE with the TSC Security Liaison and OBTAIN a list of unaccounted-for personnel. B. LOCATION OF UNACCOUNTED-FOR PERSONNEL

1. LOCATE unaccounted-for personnel as follows:

EC ( ) a. PAGE individuals over the plant page. ( ) b. OBTAIN feedback from co-workers/supervisors on the last known location/job assignment. ( ) c. DIRECT Security to assist in locating unaccounted for personnel. ( ) d. CALL individual's home to verify work schedule. ( ) e. IF REQUIRED, THEN DIRECT the OSCC to INITIATE Search and Rescue Operations IAW EPIP 202H. ( ) 2. UPDATE Security as missing personnel are accounted for. HCGS Rev. 00

ECG ATTl Pg. 5 of 9 ID. TERMINATION

1. WHEN EITHER of the following conditions are met, EC THEN TERMINATE the emergency by proceeding to Step 2.

( ) a. NO EALs are exceeded AND the Plant is stable. ( ) b. IF any EAL CONTINUES to be exceeded AND the Plant is stable THEN REFER to the "RECOVERY CHECKLIST" (Pg. 6) AND DETERMINE if the UE can be tenninated by entering Recovery.

2. WHEN the above Step is completed, EC THEN C01\1PLETE the "UNUSUAL EVENT TERMINATION/RECOVERY NOTIFICATION FORM," (Pg. 7), as follows:

( ) a. IF tenninating WITHOUT Recovery, C01\1PLETE Part A. ( ) b. IF tenninating WITH Recovery, C01\1PLETE Part B.

3. IF tennination with Recovery is chosen, EC THEN DIRECT the EDO to assume the duties of the Recovery Manager including:

EVALUATE the emergency and its consequences.
DETERMINE measures required to return the Plant to Normal Operations (tennination of Recovery Status).
COORDINATE contractor support, as required.

4. Make Reduction in Event Notifications (Termination) by; EC

( ) a. PROVIDE the completed "EMERGENCY TERMINATION/ RECOVERY NOTIFICATION FORM," to the CMl. ( ) b. DIRECT the CMI to make the termination notifications IAW ECG Attachment 6.

5. MAKE a PA announcement to update Plant personnel.

EC

6. NOTIFY the Salem SNSS.

EC

7. GO TO Section IV., Reporting.

SNSS HCGS Rev. 00

ECG ATI l Pg. 6 of 9 III. TERMINATION (cont'd) RECOVERY CHECKLIST FOR AN UNUSUAL EVENT THE EMERGENCY COORDINATOR SHOULD: A. ANSWER each of the following questions which are PREREQUISITES for Terminating

        . WITH Recovery.

CHECK IF YES ( ) 1. . Is the Radiological Release terminated(< Technical Specifications)? ( ) 2. Are Radiation levels in ALL areas of the Plant EITHER stable or decreasing? ( ) 3. Is the Plant in a safe, stable condition with NO reason to expect further degradation? ( ) 4. Is the integrity of the Station power supplies and ECCS equipment required for safe. shutdown intact? ( ) 5. Can full time operations of the OSC be terminated? B. IF ANY of the above are negative (unchecked), THEN termination should NOT be performed, at this time. RETURN to Section I. C. IF ALL of the above are checked as YES, THEN PROCEED with Step D. D. EDO been briefed AND (CHECK IF YES); ( ) 1. EDO concurs that terminating the UE with an EAL still exceeded is correct under the current circumstances? ( ) 2. EDO is prepared to assume the duties of Recovery Manager. Time E. IF EITHER of the above are negative (unchecked), THEN termination should NOT be performed, at this time. RETURN to Section I. F. IF BOTH D.1 & D.2 are checked as YES, THEN SIGN below and GO TO Sect. ill., Step 2 for Terminating WITH Recovery. Emergency Coordinator Date Time HCGS Rev. 00

ECG ATil Pg. 7 of 9 ID. TERMINATION (cont'd) UNUSUAL EVENT TERMINATION/RECOVERY NOTIFICATION FORM PART "A" - EMERGENCY TERMINATION WITHOUT RECOVERY: nns IS _ _ _ _ _ _ __, COMMUNICATOR IN THE CONTROL ROOM AT THE HOPE CREEK NUCLEAR GENERATING STATION. THIS MESSAGE IS TO NOTIFY YOU THAT AS OF ON - - - - - time date THE UNUSUAL .EVENT HAS BEEN TERMINATED. (EC Approval to transmit) PART "B" - EMERGENCY TERMINATION WITH RECOVERY: THIS IS COMMUNICATOR IN THE CONTROL ROOMAT

              -------~

THE HOPE CREEK NUCLEAR GENERATING STATION. THIS MESSAGE IS TO NOTIFY YOU THAT AS OF ~----' ON ____ __, time date THE UNUSUAL EVENT HAS BEEN TERMINATED AND HOPE CREEK IS NOW IN A RECOVERY STATUS. IS THE RECOVERY MANAGER.

                         ~--------

(DUTYEDO) (EC Approval to transmit) HCGS Rev. 00

ECG ATTl Pg. 8 of 9 IV. REPORTING INSTRUCTIONS

1. This is a permanent document.

2. ATTACH appropriate documents to this form and EXPEDITE the package through all steps.

1. PREPARE an Action Request (AR).

SNSS

2. FORWARD this attachment and supporting documentation to the Operations SNSS Manager (OM).

3. REVIEW this attachment, the (AR) and any other relevant information for correct OM classification of event and corrective action taken.

4. CONT ACT the LER Coordinator (LERC) and request that the required reports be OM prepared. Provide this attachment and any other supporting documentation to the LERC.

5. PREPARE required reports.

LERC Report or LER Number

6. FORWARD this attachment to the Central Technical Document Room for LERC microfilming.

HCGS Rev. 00

ECG A.TT 1 Pg. 9 of 9 INITIAL CONTACT MESSAGE FORM I. THIS I S - - - - - - - - - - , C01\1MUNICATOR IN THE CONTROL ROOM (NAME) AT THE HOPE CREEK NUCLEAR GENERATING STATION. II. 0 THIS IS NOTIFICATION OF AN UNUSUAL EVENT WHICH WAS DECLARED AT ON _ _ _ _ _ _ _ __ (Time - 24 HR CLOCK) (DATE) EAL# - - - - - - DESCRIPTION OF EVENT:

                                                                   ----------~

III. 0 NO RADIOLOGICAL RELEASE IS IN PROGRESS. see NOTE

                                                                                 } for release 0    THERE IS A RADIOLOGICAL RELEASE IN PROGRESS.                           definition 33 FT. LEVEL WIND DIRECTION (From):                            WIND SPEED: _ _ __

(From JMET Computer) (DEGREES) (:MPH) IV. NO PROTECTIVE ACTIONS ARE RECOMMENDED AT THIS TIME EC Initials (Approval to Transmit ICMF) NOTE: Radiological Release is defined as: Plant Effluent > Tech Spec Limit of 1.20E+04 µCi/sec Noble Gas or 1.70E+Ol µCi/sec I-131. HCGS Rev. 00

ECG ATT2 Pg. 1 of 4 ATTACHMENT 2 ALERT c>:..... \ I. EMERGENCY COORDINATOR CECl LOG SHEET l

                                                             !                                    I ii I                                   \

A. DECLARE AN ALERT AT HOPE CREEK La---*------~* EC EAL# - - - - - - Declared at _ _ _ _ _ hrs on _ _ _ __ time date B. NOTIFICATIONS ( ) 1. CALL communicators to the Control Room. ( ) 2. COMPLETE the INITIAL CONTACT MESSAGE FORM (ICMF) (last page of this attachment) .

    .( ) 3. PROVIDE the ICMF to the Communicator (CMl) and DIRECT the CMl to implement Attachment 6.

( ) 4. DIRECT the Secondary Communicator (CM2) to implement Attachment 8 for an ALERT.

5. NOTIFY the I.T.O.C. Operator on NETS x5027 (201-430-7191 SNSS or 201-430-8153) with the following message:

               "This is (your name) , Senior Nuclear Shift Supervisor at Hope Creek. Please IMPLEMENT EPIP 204H, Hope Creek Emergency Response Callout, immediately.

This procedure is being implemented for an Actual Emergency." notified at

               ~----------------~                                                       -----

I.T.O.C. Operator name time (EP96-003) ( ) 6. NOTIFY the Salem SNSS. (NETS 5121; DID 5200)

a. PROVIDE a briefing on the ALERT conditions.

b. DIRECT implementation ofEPIP IOlS, Section 3.1.

7. IF Security Related, EC THEN DIRECT the PSE&G Security Supervisor (x2222) to implement the Security Contingency Plan.

HCGS Rev. 00

ECG ATT2 Pg. 2 of 4 C. EMERGENCY COO RD INA TOR DUTIES

1. IF NOT done previously, EC THEN DIRECT the OSC Coordinator to ACTIVATE the OSC IAW EPIP 202H, OSC Activation and Operations.

2. IMPLEMENT EPIP 102H, Alert, while continuing in this attachment.

EC

3. C01\.1PLETE and APPROVE the NRC Data Sheet (Attachment 5) for transmittal EC by the CM I within 60 minutes.

4. WHEN provided by the CM2, EC THEN REVIEW and APPROVE the Station Status Checklist (SSCL) for transmittal.

( ) a. REPEAT this step approximately every half hour. ( ) b. PERFORM immediately for any significant change in emergency status. (operational or radiological) D. TURNOVER ( ) I. WHEN turning over EC duties, THEN DIRECT your Communicators to turnover notifications responsibilities to the oncoming facility communicators. ( ) 2. IF relieved as EC prior to termination of the ALERT, THEN DOCUMENT the name of your relief below: _ _ _ _ _ _ _ _ _ _ _ _ _ _ assumed EC duties at _ _ __ Name time E. ESCALATION IF the event classification escalates above an Alert, EC THEN EXIT this attachment and implement a new attachment as directed by the EALs. F. TERMINATION

1. TERMINATE the ALERT IAW EPIP 106H, Emergency Termination/Reduction EC /Recovery.

2. ENSURE appropriate reports are made IAW Section III, Reporting, of this SNSS attachment.

HCGS Rev. 00

ECG ATT2 Pg. 3 of 4 Il. REPORTING INSTRUCTIONS

1. This is a permanent document.

2. ATIACH appropriate documents to this form and EXPEDITE the package through all steps. .

1. PREPARE an Action Request (AR).

SNSS

2. FORWARD this attachment and supporting documentation, to the Operations SNSS Manager (OM).

3. REVIEW this attachment, the (AR) and any other relevant information for correct OM

classification of event and corrective action taken.

4. CONTACT the LER Coordinator (LERC) and request that the required reports OM be prepared. Provide this attachment and any other supporting documentation to the LERC.

5. PREP ARE required reports.

LERC Report or LER Number

6. FORWARD this attachment to the Central Technical Document Room for LERC microfilming.

HCGS Rev. 00

ECG ATT2 Pg. 4 of 4 INITIAL CONTACT MESSAGE FORM I. THIS IS

             - - - - - - - - - , COMMUNICATOR IN THE 0                          CONTROL ROOM (NAME)                                                0   TSC AT THE HOPE CREEK NUCLEAR GENERATING STATION.

II. D nns IS NOTIFICATION OF AN ALERT WHICH WAS DECLARED AT ON (Time - 24 HR CLOCK)

                                                                - - -(DATE)

EAL# _ _ _ _ _ DESCRIPTION OF E V E N T : - - - - - - - - - - - - - - III. D NO RADIOLOGICAL RELEASE IS IN PROGRESS. see NOTE

                                                                               } for release D     THERE IS A RADIOLOGICAL RELEASE IN PROGRESS.                          definition 33 FT. LEVEL WIND DIRECTION (From):                              WIND SPEED: _ _ __

(From MET Computer) (DEGREES) (MPH) IV. NO PROTECTIVE ACTIONS ARE RECOMMENDED AT THIS TIME EC Initials (Approval to Transmit IC:MF) NOTE: Radiological Release is defined as: Plant Effluent> Tech Spec Limit of l.20E+04 µCi/sec Noble Gas or l.70E+OI µCi/sec I-131.

ECG ATT3. Pg. 1 of 5 ATTACHMENT 3 SITE AREA EMERGENCY t I. EMERGENCY COORDINATOR (EC) LOG SHEET l

                                                             ~
                                                               ~*

l. j l~.---*-a--*'11..,..,:J~~~~ A. DECLARE A SITE AREA EMERGENCY AT HOPE CREEK EC EAL #(s) _ _ _ _ _ _ - - - - - - - " * ' _ _ _ _ _ __ Declared at hrs on *

                      -----~                      -----~

time date B. NOTIFICATIONS ( ) I. CALL communicators to the Control Room. ( ) 2. COMPLETE the INITIAL CONT ACT MESSAGE FORM (ICMF) (last page of this attachment). ( ) 3. PROVIDE the ICMF to the Communicator (CM!) and DIRECT the CMI to implement Attachment 6. ( ) 4. DIRECT the Secondary Communicator (CM2) to implement Attachment 8 for a SITE AREA EMERGENCY.

5. IF NOT done previously, SNSS NOTIFY the I.T.O.C. Operator on NETS x5027 (201-430-7191 or 201-430-8153) with the following message:

                "This is (your name) , Senior Nuclear Shift Supervisor at Hope Creek. Please IMPLEMENT EPIP 204H, Hope Creek Emergency Response Callout, immediately.

This procedure is being implemented for an Actual Emergency." notified at I.T.O.C. Operator name time (EP96-003) ( ) 6. NOTIFY the Salem SNSS. (NETS 5121; DID 5200)

a. PROVIDE a briefing on the SAE conditions.

b. DIRECT implementation ofEPIP IOIS, Section 3.2.

7. IF Security Related, EC THEN DIRECT the PSE&G Security Supervisor (x2222) to implement the Security Contingency Plan.

HCGS Rev. 00

ECG ATTJ Pg. 2 of 5 C. EMERGENCY COO RD INA TOR DUTIES

l. IF NOT done previously, EC THEN DIRECT the OSC Coordinator to ACTIVATE the OSC IAW EPIP 202H, OSC Activation and Operations.

2. IF the Emergency Coordinator is the EDO or SNSS, SNSS/EDO THEN REFER TO EPIP 103H, Site Area Emergency, AND IMPLEMENT emergency actions assigned to the EDO until relieved while continuing at Step C.4.

3. IF the Emergency Coordinator is the ERM, ERM THEN continue to REFER to EPIP 401 AND

( ) a. NOTIFY the EDO of SAE details;

Time of declaration
EAL exceeded (Basis)
DIRECT the EDO to IMPLEMENT EPIP 103H, Site Area Emergency

( ) b. NOTIFY EOF Staff of the change in classification.

4. COMPLETE and APPROVE the NRC Data Sheet (Attachment 5) for transmittal by EC the CMl within 60 minutes.

5. WHEN provided by the CM2, EC THEN REVIEW and APPROVE the Station Status Checklist (SSCL) for transmittal.

( ) a. REPEAT this step approximately every half hour. ( ) b. PERFORM immediately for any significant change in emergency status. (operational or radiological)

6. IF a Protective Action Recommendation (PAR) is developed with no escalation EC of the SAE level, THEN;

( ) a. COMPLETE a new IC:MF (ECG Attachment 3) for PAR UPGRADE. ( .) b. PROVIDE the ICMF to the CMl and DIRECT the CMl to IMPLEMENT a new ECG Attachment 6 for PAR UPGRADE notifications. HCGS Rev. 00

ECG ATT3 Pg. 3 of 5 D. TURNOVER ( ) 1. WHEN turning over EC duties, THEN DIRECT your Communicators to turnover notifications responsibilities to the oncoming facility communicators. ( ) 2. IF relieved as EC prior to termination of the SAE, THEN DOCUMENT the name of your relief below: Name time E. ESCALATION IF event classification escalates above an SAE, EC THEN EXIT this attachment and IMPLEMENT a new attachment as directed by the EALs. F. TERMINATION

1. TERMINATE the SAE IAW EPIP 106H, Emergency Termination/Reduction EC /Recovery.

2. ENSURE appropriate reports are made IAW Section II, Reporting, of this SNSS attachment.

HCGS Rev. 00

ECG ATT3 Pg. 4 of 5 II. REPORTING

                                        . INSTRUCTIONS

1. This is a permanent document.

2. ATTACH appropriate documents to this form and EXPEDITE the package through all steps.

1. PREP ARE an Action Request (AR).

SNSS AR# _ _ _ _ _ _ __

2. FORWARD this attachment and supporting documentation, to the Operations SNSS Manager (OM).

3. REVIEW this attachment, the (AR) and any other relevant information for correct OM classification of event and corrective action taken.

4. CONT ACT the LER Coordinator (LERC) and request that the required reports be OM prepared. Provide this attachment and any other supporting documentation to the LERC.

5. PREP ARE required reports.

LERC Report or LER Number _ _ _ _ _ __

6. FORWARD this attachment to the Central Technical Document Room for LERC microfilming.

HCGS Rev. 00

,----------~- ---- ~---- ECG ATT3 Pg. 5 of 5 INITIAL CONTACT MESSAGE FORM THIS IS I.

                          - - - - - - - - - , CO:l'.WUNICATOR IN THE                    0 CONTROL ROOM (NAME)                                              0 TS C 0EOF AT THE HOPE CREEK NUCLEAR GENERATING STATION.
        *Ila. D       THIS IS NOTIFICATION OF A SITE AREA EMERGENCY WHICH WAS DECLARED AT _ _ _ _ _ _ _ ON _ _ _ _ _ __

(TIME - 24 HOUR CLOCK) (DATE) EAL #(s) _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ DESCRIPTION OF EVENT: lib. 0 THIS IS NOTIFICATION OF A PROTECTIVE ACTION RECOMMENDATION UPGRADE WHICH WAS MADE AT HRS ON _ _ _ __ (24 HOUR CLOCK) (DATE) Reason for PAR U p g r a d e : - - - - - - - - - - - - - - - - - - - III. D NO RADIOLOGICAL RELEASE IS IN PROGRESS. see NOTE

                                                                                           } for release D       THERE IS A RADIOLOGICAL RELEASE IN PROGRESS.                        definition 33 FT. LEVEL WIND DIRECTION (From): _ _ _ _ _ _ WIND SPEED: _ _ __

(From MET Computer) (DEGREES) (MPH) IV. D NO PROTECTIVE ACTIONS ARE RECOMMENDED AT THIS TI.ME Sectors Dist-Miles 0 WE RECOMMEND EVACUATION AS FOLLOWS 0 WE RECOMMEND SHELTERING AS FOLLOWS EC Initials (Approval to Transmit ICMF) NOTE: Radiological Release is defined as: Plant Effiuent >Tech Spec Limit of 1.20E+04 µCi/sec :'.'Jobie Gas or 1.70E+Ol µCi/sec 1-131. HCGS Rev. 00

ECG ATT4 Pg. 1 of 7 ATTACHMENT 4 GENERAL EMERGENCY I. EMERGENCY COORDINATOR (EC) LOG SHEET ~..: .. A. DECLARE A GENERAL EMERGENCY AT HOPE CREEK t . - - r....- -

EC EAL #(s) _ _ _ _ _ _~ --------~ - - - - - - - - -

Declared at - - - - - - - hrs on - - - - - - - time date B. NOTIFICATIONS ( ) 1. CALL communicators to the Control Room. CAUTION A Protective Action Recommendation (PAR) SHALL be made on the Initial Contact Message Form (ICMF).

2. MAKE A PAR by the following steps; EC

( ) a. REFER to Predetermined PAR Flowchart on Pg. 5 and CHOOSE the appropriate PAR. ( ) b. REFER to Recommended Protective Actions Worksheet on Pg. 6 to DETERMINE the compass designations for the downwind sectors affected. ( ) c. IF a Radiologically Based PAR is IMMEDIATELY available, THEN COMP ARE the two P ARs and choose the most appropriate for inclusion on the ICMF. ( ) 3. COMPLETE the INITIAL CONT ACT MESSAGE FORM (IC:MF) (last page of this attachment). ( ) 4. PROVIDE the ICMF to the Communicator (CMI) and DIRECT the CMI to implement Attachment 7. ( ) 5. DIRECT the Secondary Communicator (CM2) to implement Attachment 8 for a GENERAL EMERGENCY. HCGS Rev. 00

ECG ATT4 Pg. 2 of 7

6. IF NOT done previously, SNSS NOTIFY the I.T.O.C. Operator on NETS x5027 (201-430- 7191 or 201-430-8153) with the following message:

              "This is (your name) , Senior Nuclear Shift Supervisor at Hope Creek. Please IMPLEMENT EPIP 204H, Hope Creek Emergency Response Callout, immediately.

This procedure is being implemented for an Actual Emergency." I.T.0.C. Operator name time (EP96-003) ( ) 7. NOTIFY the Salem SNSS. (NETS 5121; DID 5200)

a. PROVIDE a briefing on the GE conditions.

b. DIRECT implementation ofEPIP lOlS, Section 3.2.

8. IF Security Related, EC THEN DIRECT the PSE&G Security Supervisor (x2222) to implement the Security Contingency Plan.

C. EMERGENCY COORDINATOR DUTIES

1. IF NOT done previously, EC THEN DIRECT the OSC Coordinator to ACTIVATE the OSC IAW EPIP 202H, OSC Activation and Operations.

2. IF the Emergency Coordinator is the EDO or SNSS, SNSS/EDO THEN REFER TO EPIP 104H, General Emergency, AND IMPLEMENT emergency actions assigned to the EDO until relieved while continuing at Step C.4.

3. IF the Emergency Coordinator is the ERM, ERM THEN continue to REFER to EPIP 401, ERM Response, AND

( ) a. NOTIFY the EDO of General Emergency details;

Time of declaration
EAL exceeded (Basis)
DIRECT the EDO to IMPLEMENT EPIP 104H, General Emergency

( ) b. NOTIFY EOF Staff of the change in classification.

4. COMPLETE and APPROVE the NRC Data Sheet (Attachment 5) for transmittal EC by the CMl within 60 minutes.

HCGS P.ev. 00

ECG ATI4 Pg. 3 of 7

5. WHEN provided by the CM2, EC THEN REVIEW and APPROVE the Station Status Checklist (SSCL) for transmittal.

( ) a. REPEAT this step approximately every half hour. ( ) b. PERFORM immediately for any significant change in emergency status. (operational or radiological) D. TURNOVER ( ) I. WHEN turning over EC duties, THEN DIRECT your Communicators to turnover notifications responsibilities to the oncoming facility communicators. ( ) 2. IF relieved as EC prior to termination of the GE, THEN DOCUMENT the name of your relief below:

             ~~~~~~~~~~~~~~-

assumed EC duties at -~~~~ Name time E. TERMINATION I. TERMINATE the GE IAW EPIP 106H, Emergency Termination/Reduction EC /Recovery.

2. ENSURE appropriate reports are made IAW Section II, Reporting, of this SNSS attachment.

HCGS Rev. 00

ECG ATT4 Pg. 4 of 7 II. REPORTING INSTRUCTIONS

1. This is a permanent document.

2. ATTACH appropriate documents to this form and EXPEDITE the package through all steps. .

1. PREPARE an Action Request (AR).

SNSS

2. FORWARD this attachment and supporting documentation, to the Operations SNSS Manager (OM).

3. REVIEW this attachment, the (AR) and any other relevant information for correct OM classification of event and corrective action taken.

4. CONT ACT the LER Coordinator (LERC) and request that the required reports be OM prepared. Provide this attachment and any other supporting documentation to the LERC.

5. PREP ARE required reports.

LERC Report or LER Number _ _ _ _ _ _ __

6. FORWARD this attachment to the Central Technical Document Room for LERC microfilming.

HCGS Rev. 00

ECG AIT4 Pg. 5 of 7 PREDETERMINED PROTECTIVE ACTION RECOMMENDATION CHART PAR REQUIRED FOR GENERAL EMERGENCY yes EVACUATE ALL SECTORS 0- 5 MILES

                         >--~EVACUATE        DOWNWIND+/- I SECTOR 5-10 MILES SHELTER ALL REMAINING SECTORS 5-10 MILES No
             --~~~~~--1~          EVACUATEALLSECTORS              0-5 MILES DEFAULT PAR (any other GE)

CAUTION: IF TRAVEL CONDITIONS PRESENT AN EXTREME HAZARD (SEVERE ICE, SNOW, WIND, FLOODS, QUAKE DAMAGE, ETC.), CONSIDER SHELTER INSTEAD OF EV A CU ATE IN THE ABOVE SELECTED PAR. HCGS Rev.00

EC~

                                                                                    ,;TT .;

RECO:, '.lE\DED PROT[C'Tl\"E AC'TI0\3 wom~S'HEE:T Pg G of 7 WI\D DIRECTION FROM DEGREES COMPASS __. PAR AFFECTED SECTORS DOW~WIND !l SECTOR 349 - 011 N SSE - s - SSW 011 - 034 NNE s - SSW - SW 034 - 056 NE SSW - SW - WSW 056 - 079 ENE SW - WSW - w 079 - 101 E WSW - w - WNW 101 - 124 ESE w - WNW - NW 124 - 146 SE WNW - NW - NNW 146 - 169 SSE NW - NNW - N 169 - 191 s NNW - N - NNE 191 - 214 SSW N - NNE - NE 214 - 236 SW NNE - NE - ENE 236 - 259 WSW NE - ENE - E 259 - 281 w ENE - E - ESE 281 - 304 WNW E - ESE - SE 304 - 326 NW ESE - SE - SSE 326 - 349 NNW SE - SSE - s NOTE: CONSIDER ADDING A SECTOR TO THE PAR IF THE WIND DIRECTION (FROM) JS WITHIN

       +/- 3° OF A SECTOR DIVIDING LINE.

281' 079' WEST EAST 259' 101' 191' SOUTH 169' HCGS Rev. 00

ECG ATT4 Pg. 7 of 7 INITIAL CONTACT MESSAGE FORi'\1 I. THIS IS

              ~--------
                                             , COMMUNICATOR IN THE            0 CONTROL ROOM (NAME)                                               . 0 TSC DEOF AT THE HOPE CREEK NUCLEAR GENERATING STATION.

Ila. 0 THIS IS NOTIFICATION OF A GENERAL EMERGENCY WHICH WAS DECLARED AT ON (TIME - 24 HOUR CLOCK) - - - - (DA -- TE)- - - EAL #(s) _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ DESCRIPTION OF E V E N T : - - - - - - - - - - - - - - - - - - lib. D nns IS NOTIFICATION OF A PROTECTIVE ACTION RECOMMENDATION UPGRADE WHICH WAS MADE AT HRS ON - - - - - - (24 HOUR CLOCK) (DATE) Reason for PAR Upgrade:------------~------ III. D NO RADIOLOGICAL RELEASE IS IN PROGRESS. see NOTE

                                                                                 ) for release D     THERE IS A RADIOLOGICAL RELEASE IN PROGRESS.                           definition 33 FT. LEVEL WIND DµIBCTION (From):                              WIND SPEED: _ _ __

(From MET Computer) (DEGREES) (MPH) IV. Sectors Dist.- Miles 0 WE REC01\1MEND EVACUATION AS FOLLOWS 0 WE REC01\1MEND SHELTERING AS FOLLOWS EC Initials (Approval to Transmit ICMF) NOTE: Radiological Release is defined as: Plant Effiuent >Tech Spec Limit of l.20E+04 µCi/sec Noble Gas or 1.70E+Ol µCi/sec 1-131. HCGS Rev. 00

ECG ATT 5 Pg. 1 of 7 ATTACHMENT 5 f~*- -{;~~~**.*:.~*~**~~~~JL COPY# NRC DATA SHEET COMPLETION I . REFE~NCE I I. INSTRUCTIONS ,_ 1* I NOTE This attachment is implemented when the NRC Operations Center or Regional Office is notified of an Emergency OR Non-Emergency as specified by the appropriate ECG Attachment. Information is offered as a GUIDELINE to personnel completing the Event Description and the NRC Event Update Sections of the NRC DATA SHEET. A. OBTAIN a working copy of the NRC Data Sheet (last three pages of this attachment) each time you are directed to complete it. (i.e., each change in classification or new event, begin again) ~ B. C01\1PLETE the NRC Data Sheet with reference to the following information and guidance, as needed.

1. The following paragraphs briefly describe the type of information expected by the NRC when making notifications.

2. Event Description Instructions from the NRC Data Sheet state:

                     " Include systems affected, actuations & their initiating signals, causes, effect of event on plant, actions taken or planned, etc. note anything unusual or not understood Indicate systems and safety-related equipment that are not operational. "

a) Include systems affected, ...

Description:

The NRC is primarily concerned about the safety significance of the event and the current conditions of the plant. However, some events may be caused by non-safety related equipment failures and this information should also be provided to the NRC. Common information should be the response of available systems, (ESF or ECCS systems required to respond) or any other system utilized to mitigate the consequences of the event. HCGS Rev. 00

ECG ATT 5 Pg. 2 of 7 b) ... actuations and their initiating signals, causes, ...

Description:

The NRC routinely needs to know what specific signal caused the Reactor trip or ECCS/ESF actuation. If the cause of the event or actuation is known, it should be provided. If the cause is not yet known, that information should be provided to the NRC. When the information becomes available, the NRC should be provided updated information (utilize the bottom of page two of the NRC DATA SHEET to provide the updated information). Common information should be the specific signal that caused the Reactor Scram or ECCS/ESF actuation and, if known, whether the parameter has been restored to the previously established band for the current plant conditions. c) ...effect of event on plant, ...

Description:

This information should be complete to allow a clear evaluation of current plant conditions. Incorporated in the explanation should be a description of how the event has affecte*d overall plant safety. Common information should be which safety parameters are affected. This explanation should also include how the parameters are being maintained. (Examples: Rx Press. control is being maintained by cycling SRVs or RPV Water level is being maintained by HPCI) d) ... actions taken or planned, ...

Description:

This should be a description of the current plans to mitigate the event or restore the plant to a normal configuration. The focus should be on the short term considerations and not on what you expect to have to accomplish tomorrow or next week. Common information should be corrective actions taken to mitigate the consequences of the event and the OSC priorities to reestablish specific control of plant safety parameters. e) Note anything unusual or not understood.

Description:

The NRC is interested in what systems did NOT respond as you expected and there is no apparent reason why they did not function. HCGS Rev. 00

ECG ATT 5 Pg. 3 of 7 Common information should be systems that failed to respond, systems that had responded correctly, but are currently failing to properly restore monitored parameters to their nominal values, or any unexpected plant response. f) Indicate systems and safety related equipment that are not operational.

Description:

All non-operational safety related equipment should be provided. Also provide non-operational plant equipment that may be important to event response or assessment. Common information should be equipment that was inoperable prior to the event that is safety related, non safety related equipment that caused the transient, or plant systems that would ease the operational response to the transient. Example: SPDS.

3. NRC Event Update Instructions from the NRC Data Sheet state:

       " (Document additional information provided to the NRC due to their request or as a result ofplant/ event status changes.). "

a) This section of the NRC Data Sheet is intended to be utilized to document additional information requested by the NRC. The individual communicating with the NRC should document the requested information and the response given. This section should also be utilized to update the NRC as plant conditions or equipment availability changes occur or any actions taken in accordance with 10CFR50.54(x). Also to report the results of investigations or event analysis that yields information previously reported as unknown OR that is now known to have been incorrect as reported earlier. b) If changing plant conditions result in a change in Emergency Classification, the Communicator should implement another ECG Attachment 8. This will result in a new NRC Data Sheet being completed and provided to the NRC within the 1 hour time limit. HCGS Rev. 00

ECG ATT 5 Pg. 4 of 7 II. NRC DATA SHEET FORM A. The following two page form with continuation sheet(s) is used for both emergencies and non-emergencies. B. NRC Data Sheet (Page 1 of_) should always be completed as thoroughly as possible prior to notifying the NRC, but in no case should notifications be delayed because of missing information.

c. (Page 2 of_) may or may not be applicable as determined by the Emergency Coordinator (EC).

D. (Page_ of_) is a continuation form to be used by the Communicator (or EC) to document any additional information reported to the NRC, as needed. Information recorded here as NRC updates should log the time that the NRC was updated. HCGS Rev. 00

ECG ATT. 5 NRC DAT A SHEET (Page l of _ ) Pg. 5 of 7

\: .  :::.:.~                   ':*"'**         ..
                                  ............ 1,
~ ,*~r~i   -.

J USTE~tJ 7iME z**j'lE EVENT CLASSIFICATION (Check One) GC:NE. AL:K lHR 10CFR50.72(b)(1) *( I HR SECLRITY j SAFEGUARDS Si TE UNUSUAL E'o/ENT 4HR 10CFRSC.~2(b)(2) *( TP.ANS=G~TA TION t:VEtH OTHE.R (DESCi!BE): roP. NO N-E~ERGE~CIES FP.OVIDE Tri~ STAEMo~L

SP;::nc SUEPAP.T NUMBEP. or THE IOCfRSC.72 REPO*TING REOJiREMENT rROM THE ECG INITIATING :cNO TCN EVENT DESCRIPTION Inc ude systems affected. ~ctuations & their initiating signals, :auses, effect of event on plant, a:tions taken or planned, etc.

Note on y*hing un~sual or not understood. Indicate sys:ems and sc'ety-related equipment thct are not operational. {Use a co~tinuation page if more room is neede.:) Complete only 1r event mclc:des an ) RCS LEAX: DAT A ( RCS leek > Tech Specs LOCATIO N OF LEAK (e.g. PUMP, VALVE, PIPE, etc.) : TIME & DATE L:AK STARTED: ON TIME DAE LEA~ RA E: gpm. T/S LEAK L!t.llTS: - - - - - - - - LAST KN OWt\ REACTOR COOLANT ACTIVITY: - - - - - - - - - WAS TH! S LEA:< A SUDDEN OR LONG-TERI.I DEVELOPt.IENT? - - - - - - NOTIFICATIONS

lRGANIZ ATICN NOTIFIED YES HO w~~ ORGANIZATION NOTIFIED YES NO ~~ ORGANIZATION NOTIFIED

 ~RC     RES :DENT                                              STATE OF NEW JERSEY                             STATE OF DELAWARE LOCAL {-AC TOWNSHIP)                                         OTHER GOVERNt.IENT AGENCIES                     t.IEDIA / PRESS RELEASE MODE CF OPrnATION UNTIL CORRECTED:_ ESTIMATED RESTART DATE: _ _ _ _ , ADOITIONAL INFO ON PAGE 2?

Addition al Information for Non-Emergency Notifications: Reportab le Action Level (RAL #) _1""'1-'-._ _ SNSS/EC APPROVAL TO TRANSMIT HCGS Rev. 00

ATT. 5 NRC DAT A SHEET (Page 2 of Pg. 6 of 7 MESSAGE DATE/TIME: - - - - - - - - - RADIOLOGICAL RELEASE DATA: (This section is only required ;o be completed if a release exceeding Te:~ S;:;ecs is :n ~r.J.~ress or has already occurred). Check Jcorrect statements and provide to the NRC. _ There is/ was a gaseous release cbove Tech Spec limits in progress(Tech Spec Limit: Noble Gas = 1.2JE +04 µC:/ sec). _There is/was an Iodine release above Tech Spec limits in progress(7ech Spec ~:mil: iodine-131 = 1.70E+OI µCi/sec). _There is/was a liquid release above Tech Spec limits in progress. _The release is ongoing (still above Tech Specs) at this time. The release was terminated (no longer above Tech Specs) at ___ hrs. _ The release was planned and can be isolated. _ The release pathway is monitored by the Radiation Monitoring System. _ Areas evacuated on site due to release concerns a r e : - - - - - - - - - - - - - - - - - - - - - Station personnel have received exposure above 10CFR20 limits. _Station personnel have been contaminated to on extent requiring offsite assistance to decon. SPECIFIC RADIOLOGICAL PARAMETERS: (Provide current values) Current Time: _ _ __ hrs. The Noble Gas release rate (from SSCL) is: _ _ _ _ _ _ _ _ _ µCi/sec. The lodine-131 release rote (from SSCL) is: µCi/sec. RELEASE PATHWAY MONITORS: (Provide monitor reading with units and alarm setpoinl only for those below listed monitors in Alarm or the HTV whenever it is venting). Monitor # and Name Current Reading High Alarm Setpoint 9RX580 South Plant Vent (SPV) Effluent µCi/sec 3.08E +03 µCi/ sec 9RX590 North Plant Vent (NPV) Effluent µCi/sec 3.08E+03 µCi/sec 9RX680 FRVS Vent Effluent µCi/sec 1.45E +03 µCi/ sec 9RX518 Hard Torus Vent (HTV) Effluent µCi/sec N/A 9RX509-512 Highest Main Steam Line mR/hr mR/hr 9RX621 Offgas "A" PreTreatment mR/hr 2.20E +04 mR/hr 9RX522 Offgcs "8" PreTreatment mR/hr 2.20E +04 mR/hr 9RX525 Offaas "A" Treated cpm 5.00E +04 cpm 9RX626 Offgcs "8" Treated cpm 5.00E +04 cpm OTHER PERTINENT INFORMATION: (Document additional information related to any radiological release). {Use a continuation page if more room is needed) SNSS/EC APPROVAL TO TRANSMIT HCGS Rev. 00

ECG ATT. 5 NRC DAT A SHEET !Page of Pg. 7 of 7 NOTIFICATION DATE/TIME: - - - - - - - - - 0 EVENT DESCRIPTION (Continued): 0 OTHER PERTINENT INFORMATION (Continued): D NRC EVENT UPDATE (Document additional information to NRC due to their request or as a result of plant/ event status changes): (Use a continuation page if more room is needed) SNSS/EC APPROVAL TO TRANSMIT HCGS Rev. 00

ECG ATT6 Pg. 1 of 8 ATTACHMENT 6 PRIMARY COMMUNICATOR LOG

                                                                             ,~ **-*********"."'.'***--~-~--

Table of Contents i ~:, * *:*nu;_ COPY# 1 Pages i 1 - 3 Notifications & Incoming Calls i 1 4

Termination 5-8 Communications Log

                                                                           --------~-----qa--~--~-\

I Emergency Classification: (circle) UE ALERT SAE Name: ~--~-~~~~~~~~~- Position: CM1 fTSC1/ EOF1 (Print) (circIe) A. NOTIFICATIONS NOTE A new Attachment 6 is required to be implemented if the classification changes. CAUTION Fifteen minute clock for notification starts at time event was declared. Initials

1. CALL each Organization or Individual identified on the Communications Log CMlffSCl (Pgs: 5 - 8) and READ the ICMF.

/EOFl

2. IF required to activate an individual's pager, CMlffSCl THEN PERFORM the following:

/EOFl

a. DETERMINE a non-NETS phone number for the pager holder to call back on and note it here.

Call Back#: 609-339-- - - - - - -

b. DIAL the pager number of the individual you are trying to contact.

HCGS Rev.00

ECG ATT6 Pg. 2 of 8

c. WHEN you hear "Beep, Beep, Beep,."

THEN ENTER.the Call Back.#. .

d. HANG UP the phone and CONTINUE making other notifications per Step 2.

_ _ _ 3. FAX the ICMF to Group A. CMlrfSCl IEOFl B. TURNOVER

1. WHEN CONTACTED by the TSC (or EOF) in preparing for notifications CMlrfSCl responsibilities, THEN PROVIDE the following information;

Organizations/Individuals notified.
Phone numbers or locations of Individuals for updates or changes in status.

_ _ _ 2. WHEN the EC function transfers to the oncoming facility, CM1rrsc1 THEN contact the oncoming communicator and turnover notifications. C. INCOMING CALLS NOTE Initial Notifications take priority over incoming calls. STATE OFFICIALS

1. IF Notifications authority has transferred, CMlrfSCl THEN DIRECT the caller to contact the TSC (or EOF if activated).

_ _ _ 2. WHEN contacted by any State Agency Officials (listed here), CMlrfSCl IEOFl DEMA- Delaware Emergency Management Agency AAAG- Delaware Accident Assessment Advisory Group BNE - NJ Bureau of Nuclear Engineering DEP- NJ Dept. of Environmental Protection OEM - NJ Office of Emergency Management HCGS Rev.00

ECG ATT6 Pg. 3 of 8 C. INCOMING CALLS (cont'd) Initials PERFORM the following; ( ) a. OBTAIN and RECORD; Agency Caller's Name Phone# ( ) b. READ the latest EC approved SSCL. ( ) C. IF caller is NJ-BNE, DEMA, or AAAG, THEN also READ the approved NRC Data Sheet Event Description information. NEWS MEDIA CAUTION Communicators are NOT authorized to release any information to the News Media.

3. WHEN contacted by any News Media representative, CM1rrsc1 READ the appropriate message below;

/EOFl ( ) a. IF the ENC is not activated (Unusual Event), say; "You are requested to contact the Nuclear Communications Office at any of the following numben; 609-339-1001, -1006, or -1002." ( ) b. IF the ENC is activated (ALERT or higher), say; "You are requested to contact the Media Information Operator at any of the following numben; 609-273-0188, -0282, -0386, -0479, or -0586." D. CONTINUOUSDUTIES _ _ _ 1. ASSIST the CM2 gathering and faxing operational data. CMl _ _ _ 2. ASSIST the TSC2 (or EOF2) in maintaining facility status boards TSCl/EOFl. HCGS Rev. 00

ECG ATT6 Pg. 4 of 8

3. IF the telecopier is NOT working correctly, CMl THEN CALL the TSC - Emergency Preparedness Advisor (EPA) for assistance.

E. TERMINATION/REDUCTION

1. WHEN the Emergency has been terminated or reduced in classification, CMl!TSCl THEN; EOFl

( ) a. OBTAIN the EC approved EMERGENCY TERMINATION/ REDUCTION FORM. NOTE Time limits for notifications of Emergency Termination only apply to the NRC (as soon as possible, but< 60 minutes) ( ) b. CALL each Organization or Individual identified on the Communications Log and READ the message. --- 2. WHEN the emergency is terminated, CMl!TSCl THEN FORWARD this document and all completed Forms to the SNSS (TSS/SSM). /EOFl HCGS Rev. 00

ECG ATT6 Pg. 5 of 8 EVENT COMMUNICATIONS LOG INITIAL NOTIFICATIONS REDUCTION NAME CLASSIFICATION: NAME OF TIME (DE/A/SAE) OF DATE CONTACT LIMIT CONTACT /TIME CALLER /TIME ORGANIZATION/INDIVIDUALS DELAWARE STATE POLICE/DEMA 15 MIN. Primary: NETS 5406/5407 Secondary: (SP)302-739-5851 Call Back: or (DEMA)302-834-7250 BACKUP: NAWAS NOTES: IF DELAWARE IS CONTACTED, PROCEED WITH NEW JERSEY. IF NOT, THEN CONTACT BOTH COUNTIES IN DELAWARE. NEW CASTLE COUNTY \: Primary: NETS 5408 Secondary: 302-738-3131 KENT COUNTY Primary: NETS 5409 Secondary: 302-678-9111 NEW JERSEY STATE POLICE/OEM 15 MIN. Primary: NETS 5400 Secondary: 882-4201 Call Back: BACKUP: EM RAD NOTES: IF NEW JERSEY IS CONTACTED, PROCEED WITH NEXT PAGE. IF NOT, THEN CONTACT ALL OF THE FOLLOWING. SALEM COUNTY Primary: NETS 5402 [/ .. Secondary: 769-2959 y CUMBERLAND COUNTY Primary: Secondary: NETS 5403 455-8770 ;~ U.S. COAST GUARD (Speak Only to Duty Desk) Primary: 215-271-4940 ? Secondary: 215-271-4800 :**:** c HCGS Rev. 00

ECG ATT6 Pg. 6 of 8 EVENT COMMUNICATIONS LOG INITIAL NOTIFICATIONS REDUCTION NAME CLASSIFICATION: NAME OF TIME (UE/A/SAE) OF DATE CONTACT LIMIT CONTACT /TIME CALLER /TIME ORGANIZATION/INDIVIDUALS LAC TOWNSHIP 30 MIN. Primary: NETS 5404 Secondary: 935-7300 NRC OPS CENTER COMMUNICATIONS INSTRUCTIONS

1. OBTAIN the approved NRC Data Sheet

2. READ both the ICMF and NRC Data Sheet.

3. DOCUMENT the notification below.

4. IF the NRC requests additional information concerning the event, THEN OBTAIN assistance from CR (TSC/EOF) Staff to ENSURE it is accurate and EC approved.

5. IF the NRC requests an open line be maintained, THEN OBTAIN assistance in completing any remaining calls. (see Note below)

NRC OPERATIONS CENTER 60 MIN. (ICMF & NRC Data Sheet) Primary: (ENS) 301-816-5100 Secondary: 301-951-0550 NOTE An additional communicator (preferably an RO or SRO) may be assigned to provide continuous updates to the NRC under the following circumstances; o NRC requests an open line be maintained o Additional qualified communicator is available AND is not required for actions to mitigate the emergency (higher priority activities) in the judgment of the EC HCGS Rev. 00

ECG ATT6 Pg. 7 of 8 EVENT COMMUNICATIONS LOG INITIAL NOTIFICATIONS REDUCTION NAME CLASSIFICATION: NAME OF TIME (UE/A/SAE) OF DATE CONTACT LIMIT CONTACT /TIME CALLER /TIME ORGANIZATION/INDIVIDUALS EMERGENCY DUTY OFFICER SEE 70 (EDO) NOTE 1 MIN. Primary: Ref er to Roster Secondary: (Contact One) Jim Clancy .**., Office: 3144 Home: 455-9110 Pager: 478-5073 Car: Don Crouch 230-5665 I: Office: Home: Pager: 3060 302-4 78-1136 573-5269 l'i Car: 230-5458 Mark Reddemann Office: 3463 Home: 464-1778 Pager: 478-5284 Car: . 230-5680 Joe Pollock Office: 2193 Home: 728-2478 Pager: 573-5702 .** or 223-3834 Car: 230-5679 PUBLIC INFORMATION SEE 70 MANAGER NUCLEAR NOTE 2 MIN. (Contact One) Trish DuBrois */ Office: 1186 *,. Home: 769-2454 iO Pager: 223-3012 Nancy Sooy i Office: 1007 Home: 795-68*31 .. } Pager: 223-3393 *A>

')

NOTE 1 NOTIFY EDO for Unusual Events ONLY. NOTE 2 After ENC activation, NOTIFY the ENC Manager (NETS -5300 or 273-1961) HCGS Rev. 00

ECG ATT6 Pg. 8 of 8 EVENT COMMUNICATIONS LOG INITIAL NOTIFICATIONS REDUCTION NAME CLASSIFICATION: NAME OF TIME (UE/A/SAE) OF DATE CONTACT LIMIT CONTACT /TIME CALLER /TIME ORGANIZATION/INDIVIDUALS NRC RESIDENTS ..*. 75 (Contact One) .*** MIN. Bob Summers :.. *. Office: 1074 ::: .. or 935-3850 or 935-5373 Home: 848-9171 Pager: 772-7037 *......? Scott Morris Office: 1019 I\)\ or 935-3850 or 935-5373 Home: 302-239-0310 Pager: --- -------- EXTERNAL AFFAIRS SEE 90 (Contact One) NOTE 3 MIN. Ross Bell Office: 1239 Home: 455-7435 Pager: 478-5213 Max LeFevre Office: 1243 Home: 263-7677 Pager: 478-5094

}.:

Ed Johnson :;,/', Office: 1486 *;,::: Home: 678-2257 Pager: 478-5040 Ji AMERICAN NUCLEAR INSURERS SEE

;~

90 NOTE 4 MIN. (ANI) 203-561-3433 NOTE 3 Not required to notify External Affairs After the EOF is activated. NOE 4 Not required to notify ANI for Unusual Events HCGS Rev. 00

ECG ATT7 Pg. l of 7 ATTACHMENT 7 PRIMARY COMMUNICATOR LOG (GE) Pages Table of Contents f-* *L-*::*:"*~t" # ..1 I l 1 - 3 Notifications & Incoming Calls

4 Termination 5-7 Communications Log

                                                                  \;GS
                                                                    . n
                                                                    '              ~_,po~~
                                                                                               \\

Emergency Classification: GENERAL EMERGENCY or PAR UPGRADE Name: ~~~~~~~~~~~~~~ Position: CM1 ffSC1/ EOF1 (Print) (circle) A. NOTIFICATIONS NOTE A new Attachment 7 is required to be implemented if the PAR is changed.

l. OBTAIN an approved Initial Contact Message Form (ICMF) from the Emergency CMlrfSCI

Coordinator (EC).

/EOFl NOTE For 15 minute notifications use NETS x5555 conference call (separate contact required for Coast Guard). Notification clock starts at time event was declared. - - - 2. CALL each Organization or Individual identified on the Communications Log CMlrfSCl (Pgs. 5 - 7) and READ the ICMF. If needed obt~in assistance from Secondary /EOFl Communicator. _ _ _ 3. FAX the IC.MF to Group A. CMlrfSCl /EOFl HCGS Rev. 00

ECG ATT 7 Pg. 2 of 7 _ _ _ 4. IF required to activate an individual's pager, CMlfrSCl THEN PERFORM the following: !EOFl

a. DETERMINE a non-NETS phone number for the pager holder to call back on and note it here.

Call Back#: 609-339-- - - - -

b. DIAL the pager number of the individual you are trying to contact.

c. WHEN you hear "Beep, Beep, Beep,"

THEN ENTER the Call Back #.

d. HANG UP the phone and CONTINUE making other notifications per Step 2.

B. TURNOVER _ _ _ _ l. WHEN CONTACTED by the TSC (or EOF) in preparing for notifications CMlfrSCl responsibilities, THEN PROVIDE the following information;

Organizations/Individuals notified.
Phone numbers or locations of Individuals for updates or changes in status.

____ 2. WHEN the EC function transfers to the oncoming facility, CMlfrSCl THEN contact the oncoming communicator and turnover notifications. C. INCOMING CALLS NOTE Initial Notifications take priority over incoming calls. STATE OFFICIALS

1. IF Notifications authority has transferred, CMlfrSCI THEN DIRECT the caller to contact the TSC (or EOF if activated).

HCGS Rev. 00

ECG ATT7 Pg. 3 of 7 Initials C. INCOMING CALLS (cont'd)

2. WHEN contacted by any State Agency Officials (listed here),

CMlfTSCl /EOFI DEMA- Delaware Emergency Management Agency AAAG- Delaware Accident Assessment Advisory Group BNE - NJ Bureau of Nuclear Engineering DEP- NJ Dept. of Environmental Protection OEM - NJ Office of Emergency Management PERFORM the following; ( ) a. OBTAIN and RECORD; Agency Caller's Name Phone# ( ) b. READ the latest EC approved SSCL. ( ) c. IF caller is NJ-BNE, DEMA, or AAAG, THEN also READ the approved NRC Data Sheet Event Description. NEWS MEDIA CAUTION Communicators are NOT authorized to release any information to the News Media.

3. WHEN contacted by any News Media representative, CMlfTSCl READ the appropriate message below;

/EOFl ( ) a. IF the ENC is not activated (Unusual Event), say; "You are requested to contact the Nuclear Communications Office at any of the following numbers; 609-339-1001, -1006, or -1002." ( ) b. IF the ENC is activated (ALERT or higher), say; "You are requested to contact the Media Information Operator at any of the following numbers; 609-273-0188, -0282, -0386, -0479, or -0586." HCGS Rev. 00

ECG ATT7 Pg. 4 of 7 D. CONTINUOUS DUTIES _ _ _ 1. ASSIST the CM2 gathering and faxing operational data. CMl - - - 2. ASSIST the TSC2 (or EOF2) in maintaining facility status boards TSCl/EOFl . - -- 3. IF the telecopier is NOT working correctly, CMl THEN CALL the TSC - Emergency Preparedness Advisor (EPA) for assistance. E. TERMINATION/REDUCTION - - - l. CMlffSCl __ WHEN the Emergency has been terminated or reduced in classification, THEN*, EOFl ( ) a. OBTAIN the EC approved EMERGENCY TERMINATION/ REDUCTION FORM. NOTE Time limits for notifications of Emergency Termination only apply to the NRC (as soon as possible, but< 60 minutes) ( ) b. CALL each Organization or Individual identified on the Communications Log and READ the message.

2. WHEN the emergency is terminated, CMlffSCl THEN FORWARD this document and all completed Forms to the SNSS (TSS/SSM).

/EOFl HCGS Rev. 00

ECG ATT7 Pg. 5 of 7 EVENT COMMUNICATIONS LOG INITIAL NOTIFICATIONS REDUC'I'!ON N.Ai."1E CLASSIFICATION: NAME OF TIME GENERAL EMERGENCY OF DATE CONTAC:' LIMIT CONTACT /TIME CALLER /TIME ORGANIZATION/INDIVIDUALS NEW JERSEY STATE POLICE/OEM 15 Primary: NETS 5400 MIN. Secondary: 882-4201 Call Back: BACKUP: EM RAD DELAWARE STATE POLICE/DEMA Primc;i.ry: NETS 5406/5407 Secondary: (SP)302-739-5851 or BACKUP: (DEMA)302-834-7250 NAWAS Call Back: LAC TOWNSHIP !) Primary: NETS 5404 < Secondary: 935-7300 Call Back: SALEM COUNTY Primary: NETS 5402 Secondary: 769-2959 Call Back: Backup: EM RAD CUMBERLAND COUNTY Primary: NETS 5403 Secondary: 455-8770 Call Back: Backup: EM RAD NEW CASTLE COUNTY Primary: NETS 5408 Secondary: 302-738-3131 Call Back: *:.:

*:)

KENT COUNTY I.*** L*: Primary: NETS 5409 < Secondary: 302-678-9111 Call Back: .. U.S. COAST GUARD 15 (Speak Only to Duty Desk) MIN. Primary: 215-271-4940 Call Back: Secondary: 215-271-4800 Reminder: Use NETS -5555 (conference call) for 15 min. notifications EXCEPT U.S. Coast Guard. HCGS Rev. 00

ECG ATT 7 Pg. 6 of 7 EVENT COMMUNICATIONS LOG INITIAL NOTIFICATIONS REDUCTION NA.i"1E CLASSIFICATION: NAME OF TIME GENERAL EMERGENCY OF DATE CONTACT LIMIT CONTACT /TIME CALLER /TIME ORGANIZATION/INDIVIDUALS NRC OPS CENTER COMMUNICATIONS INSTRUCTIONS

1. OBTAIN the approved NRC Data Sheet

2. READ both the ICMF and NRC Data Sheet.

3. DOCUMENT the notification below.

4. IF the NRC requests additional information concerning the event, THEN OBTAIN assistance from CR (TSC/EOF) Staff to ENSURE it is accurate and EC approved.

5. IF the NRC requests an open line be maintained, THEN OBTAIN assistance in completing any remaining calls. (see Note below)

NRC OPERATIONS CENTER 60 (ICMF & NRC Data Sheet) MIN. Primary: (ENS) 301-816-5100 Secondary: 301-951-0550 PUBLIC INFORMATION SEE 70 MANAGER NUCLEAR NOTE 1 MIN. (Contact One) Trish DuBrois Office: 1186 Home: 769-2454 Pager: 223-3012 Nancy Sooy Office: 1007 Home: 795-6831 Pager: 223-3393 NOTE An additional communicator (preferably an RO or SRO) may be assigned to provide continuous updates to the NRC under the following circumstances; o NRC requests an open line be maintained o Additional qualified communicator is available AND is not required for actions to mitigate the emergency (higher priority activities) in the judgment of the EC NOTE 1 After.ENC activation, NOTIFY the ENC Manager (NETS -5300 or 273-1961) HCGS Rev. 00

ECG ATT7 Pg. 7 of 7 EVENT COMMUNICATIONS LOG INITIAL NOTIFICATIONS F.EDUCTION NAME CLASSIFICATION: NAME OF TIME GENERAL EMERGENCY OF DATE CONTACT LIMIT CONTACT /TIME CALLER /TIME ORGANIZATION/INDIVIDUALS NRC RESIDENTS 75 (Contact One) MIN. ' Bob Summers Office: 1074 or 935-3850 or 935-5373 Home: 848-9171 > Pager: 772-7037 y<

Scott Morris ,,<** Office: 1019 ? or 935-3850 or 935-5373 ~ Home: 302-239-0310 Pager: --- -------- EXTERNAL AFFAIRS SEE 90 (Contact One) NOTE 2 MIN. Ross Bell Office: 1239 Home: 455-7435 Pager: 478-5213 Max LeFevre Office: 1243 Home: 263-7677 Pager: 478-5094 i>L Ed Johnson 1: Office: Home: Pager: 1486 678-2257 478-5040 f 'i [:, AMERICAN NUCLEAR INSURERS 90 MIN. (ANI) 203-561-3433 ;tJi NOTE 2 Not required to notify External Affairs After the EOF is activated. HCGS Rev. 00

ECG ATT8 Pg. 1 of 8 ATTACHMENT 8 SECONDARY COMMUNICATOR LOG Table of Contents Pages 1- 2 Notifications & Data Collection/Transmission 3-4 Incoming Calls (BNE, DEMA, OEM, AAAG, etc.) 5 Major Equipment & Electrical Status (MEES) form 6 Operational Status Board (OSB) form 7-8 Station Status Checklist (SSCL) form Emergency Classification: (circle) UE ALERT SAE GE N a m e : - - - - - - - - - - - - - Position: CM2 ffSC2/ EOF2 (Print) (circle) A. NOTIFICATIONS NOTE A new Attachment 8 is required to be implemented if the classification changes. Initials

1. If GE classification, assist Primary Communicator with 15 minute notification.

CM2rrsc2 IEOF2

2. DIRECT the Shift Rad Pro Tech (SRPT) (x3741) to implement EPIP 301H, RPT CM2 Onshift Response.

Name: Time: - - -

3. For an ALERT or higher emergency; CM2 ( ) a. DIRECT Security (x2223) to implement both EPIP 901, Onsite Security Response, and EPIP 903, Opening Emergency Operations Facility and Emergency News Center.

Name: Time: - - - ( ) b. CALLOUT an additional NST A. Name: Time: - - - ( ) c. ACTIVATE ERDS within 60 minutes from EITHER the NSS Office or the CR SPDS terminal;

1) PRESS <ERDS> key.

HCGS Rev.00

ECG ATT 8 Pg. 2 of 8 Initials A. NOTIFICATIONS (cont'd)

2) PRESS <Pg Up> key to select "ACTIVATE ERDS COMMUNICATION."

3) FOLLOW screen prompts.

_ _ _ 4. CO~LETE a Station Status Checklist (SSCL) Form; cM2rrsc2 ( ) a. OBTAIN SNSS (TSS/SSM) assistance, as needed for Pg.1. /EOF2 ( ) b. OBTAIN SRPT (RAC/RSM) assistance, as needed for Pg.2. ( ) c. FAX to Group B. ( ) d. IF fax transmission of the SSCL is incomplete, THEN CONT ACT the State Agencies listed below, READ the data, AND DOCUMENT on SSCL, Pg. 2. DEMA Delaware Emergency Management Agency 302-834-4531 BNE NJ Bureau of Nuclear Engineering 984-7700

5. OBTAIN a completed NRC Data Sheet and FAX form to Group B.

cM2rrsc2

/EOF2

6. REPEAT Step 4 approximately every half hour OR Il'vlMEDIATELY for significant cM2rrsc2 changes in Station status, until either Turnover or relief

/EOF2

7. TURNOVER responsibility for offsite notifications and offsite data updates (SSCLs)

CM2ffSC2 to the oncoming facility (TSC or EOF); ( ) a. GIVE names and phone numbers of contacts already made with any Offsite Agencies. ( ) b. GIVE time for next SSCL. B. DATA COLLECTIONffRANSMISSION

1. WHEN in an ALERT or higher emergency OR AFTER significant changes in CMl plant status; THEN COMPLETE the Major Equipment and Electrical Status (MEES) Form.

( ) a~ OBTAIN Licensed Operator review. ( ) b. GIVE a copy to the OSC Coordinator. ( ) c. FAX to Group C. _ _ _ 2. IF requested by the TSC, CMl THEN COMPLETE the Operational Status Board (OSB) Form every 15 minutes; (TSS may modify the frequency or data list as appropriate) ( ) a. OBTAIN Licensed Operator review. ( ) b. FAX to Group C HCGS Rev. 00

ECG ATT 8 Pg. 3 of 8 B. DATA COLLECTIONffRANSMISSION (cont'd) _ _ _ 3. VERIFY availability of"OPERATIONAL STATUS BOARD (OSB) FOR...\1" data on TSC2 the VAX printer.

               ) a. IF      OSB data is available, THEN REQUEST Rad Pro to select Menu Option #2 (Current Ops Status) every 15 minutes on the VAX LA 120.

( ) b. IF VAX data is NOT available, THEN OBI AIN data from CRIDS Page Display # 232. ( ) c. IF CRIDS data is NOT available, THEN REQUEST the CM2 in CR to begin transmitting the OSB form. _ _ _ 4. ENSURE the Facility OSB and MEES Status Boards are updated; TSC2/EOF2 ( ) a. IF OSB data is NOT available, THEN REQUEST CM2 to perform step B.2, above. (data set and frequency of updates may be revised by the TSS based on event circumstances} ( ) b. WHEN significant changes in plant systems status occur, THEN REQUEST CM2 to perform step B. l, above.

5. WHEN the emergency is terminated, CM2!TSC2 THEN FORWARD this document and all completed Forms to the SNSS (TSS/SSM).

/EOF2 C. INCOMING CALLS ST ATE OFFICIALS

1. IF Notifications authority has transferred, CM2!TSC2 THEN DIRECT the caller to contact the TSC (or EOF if activated).

_ _ _ 2. WHEN contacted by any State Agency Officials (listed here), CM2!TSC2

/EOF2                   DEMA - Delaware Emergency Management Agency AAAG -      Delaware Accident Assessment Advisory Group BNE -       NJ Bureau of Nuclear Engineering DEP - NJ Department of Environmental Protection OEM - NJ Office of Emergency Management PERFORM the following, on Pg. 4; HCGS                                                                                 Rev. 00

ECG ATT 8 Pg. 4 of 8 Initials C. INCOMING CALLS (cont'd) STATE OFFICIALS ( ) a. OBTAIN and RECORD; Agency Caller's Name Phone# ( ) b. READ the latest EC approved SSCL. ( ) c. IF caller is NJ-BNE, DEMA, or AAAG, THEN also READ the approved NRC Data Sheet Event Description. NEWS :MEDIA CAUTION Communicators are NOT authorized to release any information to the News Media. _ _ _ 3. WHEN contacted by any News Media representative, CM2/TSC2 READ the appropriate message below;

!EOF2

( ) a. IF the ENC is not activated (Unusual Event), say; "You are requested to contact the Nuclear Communications Office at any of the following numbers; 609-339-1001, -1006, or -1002." * ( ) b. IF the ENC is activated (ALERT or higher), say; "You are requested to contact the Media Information Operator at any of the following numbers; 609-273-0188, -0282, -0386, -0479, or -0586." NRC OPERATIONS CENTER

4. WHEN directed by the NRC to TERMINATE ERDS transmission, CM2 THEN GO TO any CR SPDS terminal AND PROCEED as follows;

a. PRESS <ERDS> key.

b. PRESS <Pg Dn> key to select "TERMINATE £RDS.COMMUNICATION."

C. FOLLOW screen prompts.

d. WHEN completed, NOTIFY the SNSS.

HCGS Rev. 00

ECG MEES ATT. 8 Pg. 5 OF 8

                                                                                                              ~~~~~~~~~

HOPE CREEK DATE: MAJOR EQUIPMENT AND ELECTRICAL STATUS UPDATE TIME:

                                             , REACTIVITY              .ELE~CAL,            y /N ,     CONTAINMENT

coNTROL ,.....,., CONTROL NOTE: Y = IN SERVICE '--~~~-------- ~~~~~~~~~~~~-

N = OUT OF SERVICE ; SLC PUMPS A: B212 __ FRVS RECIRC A B4IO (CIRCLE ANY UNAVAILABLE B, B2Z2 FANS E. B450 EQUIPMENT) B B420

                                             ' RWCU PUMPS Ai                 B254 B:       B264 F      B460 COOLING WATER      ELECFEEDTRICAL ,  y /N : ,

SYSTEMS c B430

                                             . REACTOR              A  I     AllO SW PUMPS       A       A401                                                                                            D       B440
                                             ; RECIRC PUMPS B          !     Al20 c       A403 FRVS VENT       A'     B212 B       A402                   : CRD PUMPS            Ai       B430 I

B440 i--' Bl B222 Bi i ! iFANS Di A404 I i I ELECTRICAL I I I Ai B410 SACS PUMPS A: A401

                                             !STATUS                                        Y/Ni     IH2                 '

c: A403 I I

RECOMBINERS B* B480

                                              . OF'FSITE AC POWER AVAILABLE B'. A402 EMERGENCY                LOADED            RUN. :    PCIG           A       B232 D       A404                    .DIESELS COMPRESSORS B          B242 EDG               A ._*- - - - - - - ' - -

RACS PUMPS A B415 B r-1_ ____,_ _

SERVICE AIR B B428 c f-----'--- COMPRESSORS c B250 oi 'OOK107 Al20

                                               !HVAC                    iELE=CAL: y /N i ' 10Kl07                              AllO CIRC          A        ASOl                                         AI        AllO                   _E_M_ _  IN_ _ _ _ _ _ __

TURBINE BLDG WATER B. A502 j i CHILLED WATER B Al20 ER. ST. AIR ELECTRICAL y /N c, ASOl I COMPRESSOR FEED PUMPS I I

cHILLERS c AIOI I

I

                                                                                                     , lOKlOO                 B450 D:       A502        '       i I

II D AllO I i I I CONDENSA TEI FEEDWATER PRIMARY A I ELECTRICAL FEED AllO I y /N I ITURBINE BLDG CHILLED WATEll CIRC PUMPS A c B Bl30 Bl20 BllO H I [I iECCS I

                                                                                                     ! RHR PUMPS I

A! c:

                                                                                                                         , ELE~CAL I

A401 A403 y /N Bl A402 I CONDENSATE B Al20 CONTROL AREA A B431 I PUMPS c Al02 CHILLED WATER B441 o! A404 : 1 CIRC PUMPS B CONTROL AREA A A403 RCIC PUMPS - STEAM i SECONDARY Al AllO CHILLED WATEll B I A404 CONDENSATE B1 Al20 II CHILLERS HPCI PUMPS - i STEJl..M I I A B4Bl PUMPS Ci Al04 TSC

CHILLED W Jl'!'ER i CORE A A40l I CIRC PUMPS B B481 FEED A STEAM SPRAY c A403 !

TSC A A401 A402 i WATER B STEAM PUMPS B CHILLED WATER I B A402 PUMPS c STEAM CHILLERS Di A404 LICENSED OPERATOR REVIEW: Rev. 00 HCGS INITIALS

JSB OPERATIONAL STATUS BOARD - HOPE CREEK

                                                                                            ~- 5 .)f .3 NOTE: 1) IF REQUESTED, TRANSMIT THIS FORM TO THE TSC AND EOF EVERY 15 MINUTES.

2) PROVIDE A COPY TO THE OSC COORDINATOR. DATE: - - - - - -

3) SEE CRIDS PAGE 232 FOR DATA.

TIMES (34-HOUR CLOCK'.) INST UNITS BALANCE OF PLANT E PLAN-A. CST LEVEL (1) x 10 4 GAL B. CONDENSER PRESSURE (2) IN. HGa

          '-* RCIC FLOW                          (3)    GPM D. FEED FLOW                           (4)    MLB/HR ECCS A. RHR/LPCI FLOW-A**                  (5)    GPM RHR/LPCI FLOW-C                    (5)    GPM RHR/LPCI FLOW-Bu                  (5)    GPM RHR/LPCI FLOW-D                    (6)    GPM B. HPCI PUMP FLOW                     (7)    GPM C. CORE SPRAY FLOW-A                  (8)    GPM CORE SPRAY FLOW-B                 (9)    GPM D. SRV (OPEN) STATUS                ( 10)    #   OPEN RX COOLANT SYSTEM A. POWER                         (11-15)     % OR CPS

3. WATER LEVEL ( 17,20,21 ,22) IN. ---*

           -*  PRESSURE                     ( 18, 19)   PSIG D. TEMPERATURE                        (23)   DEGREES F E. RECIRC FLOW -   A LOOP             (24)   X 10   3 GPM RECIRC FLOW -  B LOOP             (24)   X   10 3 GPM F. JET PUMP FLOW (TOTAL)              (25)   MLB/HR CONTAINMENT A. DRYWELL PRESSURE              (25,27)     PSIG TEMPERATURE         (28,29)     DEGREES F H2 CONC.            (30,31)     %

02 CONC. (30,31) % B. SUPP. CHAMBER PRESS. (26,27) PSIG AIR TEMPERATURE (28.29) DEGREES F WATER LEVEL (32) IN. WATER TEMPERATURE (33,34) DEGREES F C. RX BLDG. DELTA P (35,36) IN. H20 SSCL A. OFFSITE POWER AVALABLE? YES/NO B. 3 OR MORE DG'S AVAILABLE? YES/NO C. DID ANY ECCS ACTUATE? YES/NO D. IS PRIMARY CONT. ISOLATED? YES/NO E. IS PRIMARY CONT. CAPABLE OF ISOLATION? YES/NO LICENSED OPERATOR REVIEW INITIALS: - - - - OTHER SIGNIFICANT ITEMS

**1F NOT IN LPCI MODE FLOW RATE IS CIRCLED (i.e. S/D COOLING, CONT. SPRAY, ETC.)

HCGS Rev. 00

ECG

                                                                          ..\TT. 8 Pg. 7 of 8 ST ATION ST ATUS CHECKLIST SSCL                           (Pg. l of 2)

Operational Information* HOPE CREEK GE.\ER.ATl~G STATIO~ \fessage Date.____ Time._ __ Transmitted Bv: .\ame- - - - - - - - - - - Position: (CR/TSC/EOFl

1. Date and Time Event Declared: Date _ __ Time (2-1 hr clock)

Event Classification: u Cnusual Event _J Site Area Emeraencv <:> *

                                  ~ Alert              ; General Emergency

]. Cause of Event: Primary Initiating Condition used for declaration EAL #( s)

Description of the event - - - - - - - - - - - - - - - - - - - - -

-1. Status of Reactor:     ~   Scrammed/Time                  =.At  Power
      =Startup      Hot Shutdown :; Cold Shutdown            ~ Refuel

i. Rx Pressure psig Rx Temp ° F Rx Water Level 111.

6. Is offsite power available? YES \0

r. Are three or more diesel generators available? YES .\0

8. Did any Emergency Core Cooling Systems actuate? 'YES .\0

9. Containment:

A. Has the Primary Containment been isolated? * . YES .\0 B. Is the Primary Containment capable of being isolated? D YES ~ .\0

10. Other pertinent information - - - - - - - - - - - - - - - - - - -

Approved: EC or TSS or SSM HCGS Re\*. 00

EC(; HT. 8 Pg. 8 of 8 STATION STATUS CHECKLIST ( PAGE 2 OF 2 ) RADIOLOGICAL INFORMATION HOPE CREEK GENERATING STATION - CALCULATION TIME:. _ __ DATE: _ __

l. GASEOUS RELEASE> TECH SPEC (T /S) LIMITS:

(T/S LIMITS: 1.2E+04 ,,Ci/sec NG or 1.70E+01 .,Cijsec IODINE) YES: [ ] RELEASE START TIME: DATE:_ _ _ __ NO: [ ] A. RELEASE TERMINATED: YES [ ] NO [ ] N/ A [ ] B. ANTICIPATED OR KNOWN DURATION OF RELEASE:____ HOURS C. TYPE OF RELEASE: GROUND [ ] ELEVATED [ ] N/A [ ] D. ADJUSTED WIND SPEED: _ _ (mph) _ _ (m/sec) WIND DIR (deg from) _ __ E. STABILITY CLASS: (A-G) F. VENT PATH OF RELEASE: NPV [ ] SPV [ ] FRVS [ ] HTV [ ] G. NG RELEASE RATE: NPV SPV_ _ _ __ FRVS._ _ __ HTV (µCi/sec) H. 1-131 RELEASE RATE: NPV____ SPV FRVS'------ HTV DEFAULT (µ.Ci/sec) (circle if default) I. TOTAL RELEASE RATE NOBLE GAS: (µCi/sec) J. TOTAL RELEASE RATE IODINE-131: (µCi/sec)

2. PROJECTED OFFSITE DOSE RATE CALCULATIONS:

TEDE DISTANCE XU/Q TEDE DOSE THYROID- THYROID-FROM VENT RATE ( 4 DAY) CDE RATE CDE DOSE (IN MILES) ( 1 /M2) (MREM/HR) (MREM) (MREM/HR) (MREM) MEA 0.56 2.00 LP::: 5.00 EPZ 10.00

3. OTHER PERTINENT INFORMATION:

4. UPDATE TO STATES (IF VERBALLY TRANSMITTED):

NAME TIME INITIALS STATE OF NEW JERSEY: STATE OF DELAWARE AGENCY: APPROVED: EC or RAC or RSM HCGS Rev. 00

ECG ATT9 Pg. 1 of 3 ATTACHMENT 9 / ~~;~***

NON-EMERGENCY NOTIFICATIONS REfERENCE (HOPE CREEK) ; .,

i '

/'~

I I. INSTRUCTIONS I,,...,.__ \c',_) I i

                                                                            ~----..                    I This attachment is the source of the names and telephone numbers for making Non-Emergency reports as directed by the ECG Attachment in effect at this time.

NOTE The SNSS may direct a communicator to make the required notification calls. The responsibility to ensure completion of each step outlined in the ECG attachment and to ensure notification information is accurate remains with the SNSS. A. REFER to Section II of this Attachment and NOTIFY the required Individuals/ Organizations IAW the ECG Attachment in effect. B. IF required to activate an individual's pager, THEN PERFORM the following:

1. DETERMINE a non-NETS phone number for the pager holder to call back on and MAKE a note of the full call back phone number.

2. DIAL the pager number of the individual you are trying to contact listed in the Communications Log.

3. WHEN you hear "Beep, Beep, Beep,"

THEN ENTER the call back phone number.

4. HANG UP the phone.

5. CONTINUE making other notifications per Step A.

HCGS Rev. 00

ECG ATT9 Pg. 2 of 3 II. TELEPHONE NUMBER REFERENCE NOTE NOTIFY ONLY those individuals by title required by the particular ECG Attachment in effect at this time. TITLES/NAMES WORK# HOME# PAGER# CAR# OPERATIONS MGR Larry Wagner 3671 582-0067 478-5332 230-0707 Harlan Hanson 3005 302-366-1378 478-5249 230-5664 Erv Parker 1289 769-0732 478-5110 230-0530 GENERAL MANAGER Mark Bezilla 3463 478:-4614 478-5059 230-5670 Larry Wagner 3671 582-0067 478-5332 230-0707 GOVERNMENT AGENCY PRIMARY# SECONDARY# LAC DISPATCHER NETS x5404 935-7300 935-8127 (FAX) NRC OPERATIONS CENTER (ENS)301-816-5100 301-951-0550 301-816-515 l(F AX) NRC REGION ONE OFFICE 610-337-5000 TITLE SIN AMES WORK# HOME# PAGER# NRC RESIDENTS Bob Summers 1074 or 935-3850 848-9171 772-7037 Scott Morris 1019 or 935-3850 302-239-0310 NRC Office 2962 or 935-5151 HCGS Rev. 00

ECG ATT9 Pg. 3 of 3 Ii. TELEPHONE NUMBER REFERENCE (cont'd) TITLES/NAMES WORK# H0!\11E# PAGER# PUBLIC INFO MGR Trish DuBois 1186 769-2454 223-3012 Nancy Sooy 1007 795-6831 223-3393 EMERG PREP REPRESENTATIVE Craig Banner 1157 728-5043 478-5215 Jim Schaffer 1575 935-5606 478-5086 Dave Burgin 1595 582-1323 478-5062 EXTERNAL AFFAIRS Ross Bell 1239 455-7435 478-5213 Max LeFevre 1243 263-7677 478-5094 Ed Johnson 1486 678-2257 478-5040 RADIOLOGICAL SUPPORT REPRESENTATIVE John Russell 2410 451-0845 478-5082 Mark Simpson 2443 302-998-4792 478-5378 Bill Weckstein 1558 455-3237 478-5186 RADIATION PROTECTION MANAGER Terry Cellmer 3037 358-3316 478-5028 Bob Gary 3578 678-4718 478-5276 Brian Sebastian 3688 451-7571 478-5653 NUCLEAR LICENSING DUTY PAGER HOLDER ----- ---------------- 573-0893 Dave Smith 5431 302-234-1197 573-5225 Dave Powell 2002 3 02-23 9-9912 573-2358 ENVIRONMENT AL LICENSING (contact one) Jim Eggers 1339 953-9075. 573-4655 Dave Hurka 1275 299-7433 573-8278 Bob Boot 1169 302-836-8203 573-3700 Don Bowman* 3238 547-3795 573-8419 Paul Behrens

1577 691-4766 573-2496

$For Spills, Hazmat, NOT Protected Aquatic Species HCGS Rev.00 j

ECG ATT 10 Pg. 1 of 3 ATTACHMENT 10 .. -:".':,---;:::;;;:;*.ii\

                                                                                             ~-J ~        tr
                                                                                       *-*.         ' ' i I

ONE HOUR REPORT i NRC REGIONAL OFFICE I i

                                                                                                               ~
                                                                        ~-~---~--.~~~-=----*

INSTRUCTIONS (HOPE CREEK SNSS or Designee) A. REFER to Attachment 9, Non-Emergency Notifications Reference, for the current listing of individuals and phone numbers. B. INITIAL each step when completed. C. Implemented b y : - - - - - - - - - - - - - - - - - Date: _ _ __ I. NOTIFICATIONS I. COl'vlPLETE an NRC Data Sheet. ( ) OBI AIN a copy from ECG Attachment 5. ( ) OBI AIN assistance from Radiation Protection personnel, as needed. ( ) OBI AIN SNSS approval.

2. NOTIFY NRC Region I Office of the event within 1 hour.

( ) RECORD additional information provided to the NRC on the NRC Data Sheet notified at hrs

             ----------------~

name time

3. NOTIFY the NRC Resident Inspector.

notified at hrs

             ----------------~

name time

4. IF a package is received Onsite that was contaminated or exceeded external radiation limits, THEN NOTIFY the final delivering carrier.

notified at hrs

              ----------------~

name time HCGS Rev. 00

ECG AIT 10 Pg. 2 of 3

5. IF NOT done previously, THEN NOTIFY the Operations Manager (OM).

name

6. NOTIFY the Public Information Manager (PTht) - Nuclear.

name

7. NOTIFY Nuclear Licensing.

name

8. NOTIFY External Affairs.

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ notified at _ _ _ _ hrs name time

9. FAX the NRC Data Sheet to BOTH Public Information and Licensing using the programmed phone numbers on the telecopier.

HCGS Rev. 00

ECG ATT 10 Pg. 3 of 3 II. REPORTING

1. ENSURE that an Action Request (AR) is prepared._

SNSS

2. FORWARD this attachment, along with the NRC Data Sheet and any supporting SNSS documentation, to the Operations Manager (OM).

3. REVIEW this ECG attachment, the AR and any other relevant information for OM correct classification of event and corrective action taken.

4. FORWARD this attachment and any other supporting documentation to the LER OM Coordinator (LERC).

5. PREP ARE required reports.

LERC Report or LER Number _ _ _ _ _ _ _ __

6. FORWARD this attachment to the Manager - Nuclear Licensing & Regulation (MNLR)

LERC

7. ENSURE offsite (state and local) reporting requirements are met.

MNLR

8. FORWARD this Attachment/LER package to the Central Technical Document Room MNLR for microfilming.

HCGS Rev.00

ECG ATill Pg. 1 of 3 ATTACHMENT 11 _..-----------~ ONE HOUR REPORT f--*~-- '{< .... =~:JP"! tf \ (COMMON SITE) i 11 SECURITY/SAFEGUARDS: ; ~ I

                                                                     ~     ~\                              ~

(....___* _ _ _.._..~- NOTE ONLY one SNSS, Hope Creek or Salem, is required to report this event which is common to BOTH stations. I. EVENT ASSESSMENT AND DETERMINATION OF NOTIFICATION RESPONSIBILITY I. NOTIFY the Salem SNSS (NETS x5 I 2 l or DID 5200).

2. DETERMINE which Station SNSS will implement this attachment.

3. IF the Hope Creek SNSS is responsible for this notification, THEN IMMEDIATELY CONTINUE with this attachment.

4. IF the Salem SNSS will implement this attachment, THEN NO further actions are required by Hope Creek except to lend assistance as necessary in restoring the lost equipment or capabilities.

INSTRUCTIONS (HOPE CREEK SNSS or Designee) A. REFER to Attachment 9, Non-Emergency Notifications Reference, for the current listing of individuals and phone numbers. B. INITIAL each step when completed. Date: C. Implemented b y : - - - - - - - - - - - - ---- II. NOTIFICATIONS

1. COMPLETE an NRC Data Sheet.

( ) OBTAIN a copy from ECG Attachment 5. ( ) OBTAIN assistance from Security personnel, as needed. ( ) ENSURE SNSS approval. HCGS Rev.00

ECG ATf 11 Pg. 2 of 3

2. NOTIFY the NRC Operations Center of the event within 1 hour.

( ) RECORD additional information provided to the NRC on the NRC Data Sheet. name

3. NOTIFY the NRC Resident Inspector.

name

4. IF NOT done previously, THEN NOTIFY the Operations Manager (OM).

name

5. NOTIFY the Public Information Manager (PIM) - Nuclear.

notified at hrs name time

6. NOTIFY Nuclear Licensing.

notified at hrs name time

7. NOTIFY External Affairs.

notified at hrs name time

8. FAX the NRC Data Sheet to BOTH Public Information and Licensing using the programmed phone numbers on the telecopier.

9. WHEN Security provides updated information on the event, THEN NOTIFY the NRC Operations Center with appropriate updates on the event.

name HCGS Rev.00

ECG ATI 11 Pg. 3 of 3 ID. REPORTING

1. ENSURE that an Action Request (AR) is prepared.

SNSS

2. FORWARD this attachment, along with the NRC Data Sheet and any supporting SNSS documentation, to the Operations Manager (OM).

3. REVIEW this ECG attachment, the AR and any other relevant information for correct OM classification of event and corrective action taken.

4. CONTACT the Nuclear Security Support Supervisor (NSSS);

OM ( ) FORWARD this attachment and any other supporting documentation received from the SNS S. ( ) REQUEST a written report (required 30 days after the event).

5. PREP ARE the required Safeguards Event Report (30 day) IAW Security Contingency NSSS Plan Procedure, SCP-14.

6. FORWARD this attachment to the Manager - Nuclear Licensing & Regulation (MNLR).

NSSS

7. ENSURE offsite (state and local) reporting requirements are met.

MNLR

8. FORWARD this Attachment/LER package to the Central Technical Document Room MNLR for microfilming.

HCGS Rev. 00

ECG ATI12 Pg. 1 of 3 ATTACHMENT 12 ONE HOUR REPORT - NRC OPERA TIO NS INSTRUCTIONS (HOPE CREEK SNSS or Designee) A. REFER to Attachment 9, Non-Emergency Notifications Reference, for the current listing of individuals and phone numbers. B. INITIAL each step when completed. C. Implemented b y : - - - - - - - - - - - - Date:

                                                     ----- . ~--    . ~-~":.~~-\

I. NOTIFICATIONS . . ..) "-* \./.>! i I

                                                            ~.*.-*                               \
                                                     \                <;S                         \

1. C01\1PLETE an NRC Data Sheet.

( ) OBTAIN a copy from ECG Attachment 5. ( ) OBTAIN assistance from Radiation Protection personnel, as needed. ( ) OBTAIN SNSS approval.

2. NOTIFY the NRC Operations Center of the event within 1 hour.

( ) RECORD additional information provided to the NRC on the NRC Data Sheet. name time

3. NOTIFY the NRC Resident Inspector.

             ~~--~--~~-~~-~-~

notified at ~~~~ hrs name time

4. IF NOT done previously, THEN NOTIFY the Operations Manager (OM).

             ~-~~~~~~~~~~~-~~

notified at ~~-~ hrs name time

5. NOTIFY the Public Information Manager (PIM) - Nuclear.

name time HCGS Rev. 00

ECG ATI 12 Pg. 2 of 3

6. NOTIFY Nuclear Licensing.

            - - - - - - - - - - - - - - - - - - notified at _ _ _ _ hrs name                                            time

7. IF a major loss of communications capability has occurred (such as loss of ENS, NETS, DID, etc.)

THEN NOTIFY: I.T. Client Service Center: (201-430-7500 or ESSX 7500) ( ) a. ENTER [ 1 1] in response to the automated answering system prompts. ( ) b. NOTIFY the Operator that the failed system is an "Emergency Priority Circuit." notified at hrs time name

8. NOTIFY External Affairs.

            - - - - - - - - - - - - - - - - - - notified at _ _ _ _ hrs name                                           time

9. FAX the NRC Data Sheet to BOTH Public Information and Licensing using the programmed phone numbers on the telecopier.

HCGS Rev.00

ECG ATf 12 Pg. 3 of 3 II. REPORTING

1. ENSURE that an Action Request (AR) is prepared.

SNSS

2. FORWARD this attachment, along with the NRC Data Sheet and any supporting SNSS documentation, to the Operations Manager (OM).

3. REVIEW this ECG attachment, the AR and any other relevant information for correct OM classification of event and corrective action taken.

4. FORWARD this attachment and any other supporting documentation to the LER OM Coordinator (LERC).

5. PREP ARE required reports.

LERC Report or LER Number _ _ _ _ _ _ _ __

6. FORWARD this attachment to the Manager - Nuclear Licensing & Regulation (MNLR).

LERC

7. ENSURE offsite (state and local) reporting requirements are met.

MNLR

8. FORWARD this Attachment/LER package to the Central Technical Document Room MNLR for microfilming.

HCGS Rev. 00

ATTACHMENT 13 i i

                                                                 \,,~                                                      \

FOUR HOUR REPORT\ ., \ CONTAMINATION EVENTS OUTSIDE OF THE RCA,,..,....,._......... .-~-* l-_,,_. . . . ,..,.r-~,_.,.,.,.....~*" ~

INSTRUCTIONS (HOPE CREEK SNSS or Designee)

A REFER to Attachment 9, Non-Emergency Notifications Reference, for the current listing of individuals and phone numbers.

8. INITIAL each step when completed.

Date: C. Implemented b y : - - - - - - - - - - - - ---- I. NOTIFICATIONS

1. RECORD the location of the Contaminated Area(s): _ _ _ _ _ _ _ _ _ __

2. DIRECT the Shift Radiation Protection Technician (SRPT) to IMPLEMENT the Onsite Contamination Event Checklist (Pages 5 - 7) of this attachment and ASSUME responsibility as the Interim Radiological Incident Response Coordinator (RIRC).

notified at hrs

              ------------------  name time

3. IF routinely accessed areas are contaminated, THEN use the Plant PA System to warn personnel to stand clear of those areas.

4. NOTIFY a Radiological Support (RS) Representative;

( ) a. DIRECT the RS individual to REPORT to the Plant and ASSUME RIRC responsibility by relieving the SRPT. ( ) b. PROVIDE the name of the SRPT and the location of the Incident Response Control Center, if established. notified at

              ---------~~~----

name

                                                                                                   --        - - hrs time HCGS                                                                                                             Rev. 00

ECG ATf 13 Pg. 2 of 7

5. NOTIFY the Salem SNSS (NETS x5121; DID x5200)

( ) a. PROVIDE a brief description of the event. ( ) b. DIRECT a similar PA announcement be made at Hope Creek to warn personnel. ( ) C. OBTAIN any available support needed to monitor and control the spread of contamination. name time

6. NOTIFY Environmental Licensing and DIRECT that any notifications IA W the DPCC/DCR Plan be made as required.

name time

7. COMPLETE an NRC Data Sheet.

( ) OBTAIN a copy from ECG Attachment 5. ( ) OBTAIN assistance from Radiation Protection personnel, as needed. ( ) OBTAIN SNSS approval.

8. NOTIFY the LAC Dispatcher of the event.

name time

9. NOTIFY the Public Information Manager (PIM) - Nuclear.

name time

10. NOTIFY the NRC Operations Center of the event within 4 hours.

( ) RECORD additional information provided to the NRC on the NRC Data Sheet. name time

11. NOTIFY the NRC Resident Inspector.

name time HCGS Rev. 00

ECG ATI 13 Pg. 3 of 7

12. IF NOT done previously, THEN NOTIFY the Operations Manager (OM).

          - - - - - - - - - - - - - - - - - - notified at _____ hrs name                                        time

13. NOTIFY Nuclear Licensing.

          - - - - - - - - - - - - - - - - - - notified at _ _ _ _ hrs name                                        time

14. NOTIFY External Affairs.

          - - - - - - - - - - - - - - - - - - notified at _ _ _ _ hrs name                                        time

15. FAX the NRC Data Sheet to BOTH Public Information and Licensing using the programmed phone numbers on the telecopier.

( HCGS Rev. 00

ECG ATI13 Pg. 4 of 7 II. REPORTING

1. ENSURE that an Action Request (AR) is prepared.

SNSS

2. FORWARD this attachment, along with the NRC Data Sheet and any supporting SNSS documentation, to the Operations Manager (OM).

3. REVIEW this ECG attachment, the AR and any other relevant information for correct OM classification of event and corrective action taken.

4. FORWARD this attachment and any other supporting documentation to the LER OM Coordinator (LERC).

5. PREP ARE required reports.

LERC Report or LER Number _ _ _ _ _ _ __

6. FORWARD this attachment to the Manager - Nuclear Licensing & Regulation (MNLR).

LERC

7. ENSURE offsite (state and local) reporting requirements are met.

MNLR

8. FORWARD this Attachment/LER package to the Central Technical Document Room MNLR for microfilming.

HCGS Rev. 00

ECG ATf 13 Pg. 5 of 7 ONSITE CONTAMINATION EVENT CHECKLIST (Page 1 or 3) A. PURPOSE This checklist provides general guidance to the Interim and Long Term Radiological Incident Response Coordinator (RIRC) for the purpose of establishing Command and Control authority

     - and responsibility for the non-emergency coordination of Nuclear Business Unit resources in mitigating the consequences of a radiological incident outside the normal RCA.

B. RESPONSIBILITY - Checklist Implemented By; Name: Time: - - - Date: - - -

      -------------~

Interim RIRC (or SRPT) Name: Time: - - - Date: - - -

      -------------~

Long Term RIRC RIRC INSTRUCTIONS:

1. Checklist steps DO NOT need to be performed in order.

2. INITIAL or N/A each step as appropriate.

3. !E an emergency is declared, THEN CONSULT with the Emergency Coordinator (EC) to determine revised priorities of the EC based upon current circumstances.

C. INITIAL ACTIONS Initials/ Date/Time

1. PERFORM surveys to establish contaminated area boundaries. (Temporary RCA)

2. POST signs and set up barriers (ropes)

( ) RESTRICT access to the Temporary RCA until posted ( ) IF access CANNOT be adequately controlled with available RP personnel, THEN request assistance from Security.

3. DIRECT Security to prohibit vehicles from entering any affected portion of the Owner Controlled Area (OCA).

4. IF areas within the Protected Area that can be routinely accessed are contaminated, THEN PROVIDE personnel monitoring at the Security Center.

5. NOTIFY the Hope Creek RP Superintendent.

6. PROVIDE a briefing to the Salem RP Superintendent and OBTAIN resource assistance (material and personnel), as needed.

HCGS Rev.00

ECG ATf 13 Pg. 6 of 7 ONSITE CONTAMINATIO!'l EVENT CHECKLIST (Page 2 or 3) D. SUBSEQUENT ACTIONS Initials/ Date/Time

1. ESTABLISH an Incident Response Control Center in an accessible location.

(e.g., TSC, NOSF, RP Office Area) Location:

2. MAINTAIN a response log.

3. IF recovery actions will take > 24 hours, THEN DEVELOP an interim organization to handle the following aspects of the event;

Site Characterization and Decontamination
Dose Assessment
Communications
Site Access Control
Document Control
SITE CHARACTERIZATION AND DECONTAMINATION

4. DEVELOP a map of the contaminated areas.

( ) ENSURE consistent survey techniques and reporting units are used.

5. PERFORM isotopic analysis on several samples before decontamination activities begin.

6. REDUCE contamination< LLD, if reasonably achievable.

7. IF contamination CANNOT be reduced< LLD, THEN CONSIDER fixing the contamination to prevent further spreading.

DOSE ASSESS1\.1ENT

8. ESTABLISH a list of individuals who may have been contaminated.

9. IF the potential for personnel contamination is high among those who have left the Site, THEN CONSIDER having those individuals recalled.

10. IF recalled personnel are contaminated or may have carried contamination offsite, THEN CONSIDER surveying their clothing, vehicles, and homes.

11. PERFORM internal dose calculations and calculate external dose from groundshine. (both realistic and bounding case assessments)

12. PERFORM confirmatory WB Counts, as required.

13. COLLECT and PROCESS TLDs, as required.

HCGS Rev. 00

ECG ATf 13 Pg. 7 of 7 ONSITE CONTAMINATION EVENT CHECKLIST (Page 3 or 3) D. SUBSEQUENT ACTIONS (cont'd) Initials/

DOSE ASSESSrv!ENT (cont'd) Date/Time

14. IF a radiological release from a plant system has occurred, THEN CALCULATE the source term (total amount of radioactive material released).

COMMUNICATIONS

15. ENSURE ALL Site Personnel are INFORMED as to the location of contaminated areas and any additional monitoring requirements via posting in the Security Center.

( ) UPDATE postings periodically, as needed.

16. DEVELOP a communications plan to provide frequent updates to plant personnel.

DOCUMENTATION

17. OBTAIN copies of ALL surveys, sample results and other related documentation AND ENSURE they are placed in the Radiological Support files.

18. FORWARD records of residual contamination, including contamination that was fixed in place, to Nuclear Licensing for inclusion in the 10CFR50.75(g) file.

19. RETURN this checklist to the Hope Creek SNSS after all items on the checklist have been addressed.

HCGS Rev.00 L

ECG ATf 14 Pg. 1 of 3 ATTACHMENT 14 FOUR HOUR REPORT - NRC OPERA TIO NS INSTRUCTIONS (HOPE CREEK SNSS or Designee) A. REFER to Attachment 9, Non-Emergency Notifications Reference, for the current listing of individuals and phone numbers.

8. INITIAL each step when completed.

C. Implemented b y : - - - - - - - - - - - - Date: _ _ __ i.,"', ... I. NOTIFICATIONS

1. COMPLETE an NRC Data Sheet.

( ) OBTAIN a copy from ECG Attachment 5. ( ) OBTAIN assistance from Radiation Protection personnel, as needed. ( ) OBTAIN SNSS approval.

2. NOTIFY the NRC Operations Center of the event within 4 hours.

( ) RECORD additional information provided to the NRC on the NRC Data Sheet. notified at hrs name time

3. NOTIFY the NRC Resident Inspector.

notified at hrs name time

4. NOTIFY the LAC Dispatcher of the event.

notified at hrs name time

5. NOTIFY the Public Information Manager (PIM) - Nuclear.

notified at hrs name time HCGS Rev.00 L

ECG ATf 14 Pg. 2 of 3

6. IF NOT done previously, THEN NOTIFY the Operations Manager (OM).

name time

7. NOTIFY Nuclear Licensing.

name time

8. NOTIFY External Affairs.

name time

9. FAX the NRC Data Sheet to BOTH Public Information and Licensing using the programmed phone numbers on the telecopier.

HCGS Rev. 00

ECG ATT 14 Pg. 3 of 3 II. REPORTING

1. ENSURE that an Action Request (AR) is prepared.

SNSS

2. FORWARD this attachment, along with.the ~C Data Sheet and any supporting SNSS documentation, to the Operations Manager (OM). *

3. REVIEW this ECG attachment, the AR and any other relevant information for correct OM classification of event and corrective action taken.

4. FORWARD this attachment and any other supporting documentation to the LER OM Coordinator (LERC).

5. PREP ARE required reports.

LERC Report or LER Number _ _ _ _ _ _ __

6. FORWARD this attachment to the Manager - Nuclear Licensing & Regulation (MNLR).

LERC

7. ENSURE offsite (state and local) reporting requirements are met.

MNLR

8. FORWARD this Attachment/LER package to the Central Technical Document Room MNLR for microfilming.

HCGS Rev. 00 L

ECG AIT 15 Pg. 1 of 3 ATTACHMENT 15 ENVIRONMENTAL PROTECTION PLAN INSTRUCTIONS (HOPE CREEK SNSS or Designee) A. REFER to Attachment 9, Non-Emergency Notifications Reference, for the current listing of individuals and phone numbers. B. INITIAL each step when completed. C. Implemented b y : - - - - - - - - - - - - Date: I ..._ *~ ._ . _;~- ~ .. -,.!

                                                                                                     -.-~

7i I. NOTIFICATIONS i li ~ r;c: \.Q._) iL.__

1. RECORD the Event

Description:

NOTE Environmental Licensing will make the Determination of Reportability for Unusual or Important Environmental Events.

2. NOTIFY Environmental Licensing.

notified at hrs

              ---------------~                                               ---~

name time ( ) a. OBTAIN a Determination ofReportability (check below). ( ) b. RECORD "Determination Time": hrs ( ) c. CONTINUE based on the Determination, as follows; ( ) 1) 4 Hour Report to the NRC, EXIT this Attachment AND REFER to RAL # 11.8.2.a. ( ) 2) 24 Hour Report to the NRC Resident, GO TO Step 3. (next page) ( ) 3) Not reportable to the NRC, GO TO Section II, Pg. 3. HCGS Rev. 00

ECG ATf 15 Pg. 2 of 3 NOTE Required reports shall be made within the appropriate time limits from the Determination Time established in Step 2. above. Initials-3_ NOTIFY the NRC Resident Inspector within 24 hours. name time

4. IF the NRC Resident Inspector CANNOT be notified, THEN NOTIFY the NRC Operations Center within 24 hours.

name time

5. IF NOT done previously, THEN NOTIFY the Operations Manager (OM).

            ----------------~ notified at _ _ _ _ hrs name                                        time HCGS                                                                            Rev. 00

ECG ATf 15 Pg. 3 of 3 II. REPORTING

1. ENSURE that an Action Request (AR) is prepared.

SNSS AR# _ _ _ _ _ _ _ _ _ __

2. FORWARD this attachment, along with the AR and any supporting documentation, SNSS to the Operations Manager (OM).

3. REVIEW this ECG attachment, the AR and any other relevant information for correct OM classification of event and corrective action taken.

4. FORWARD this attachment and any other supporting documentation to the LER OM Coordinator (LERC).

5. PREP ARE required reports.

LERC Report or LER Number _ _ _ _ _ _ _ __

6. FORWARD this attachment to the Manager- Nuclear Licensing & Regulation (MNLR).

LERC

7. ENSURE offsite (state and local) reporting requirements are met.

MNLR

8. FORWARD this Attachment/LER package to the Central Technical Document Room MNLR for microfilming.

HCGS Rev. 00

j - z~-~--~~- l .. ,: . . G\_;~Ti 16 I

                                                                        ; ,                       Pg. 1 of f7 ATTACHMENT 16                     0:)                         .i
                                                                                                          .I SPILL/DISCHARGE REPORTING---------__!

CAUTION 15 minute notification to NJDEP may be required as identified in Step 4. INSTRUCTIONS (HOPE CREEK SNSS or Designee) A. REFER to Attachment 9, Non-Emergency Notifications Reference for the current listing of, individuals and phone numbers. B. INITIAL each indicated step when completed. C. Placekeeping Bracket ( ) and Decision/Status Box D use are optional, but recommended. Date: D. Implemented b y : - - - - - - - - - - - - ---- I. NOTIFICATIONS Initials

1. IMMEDIATELY DISPATCH Site Protection to the location of the Spill/Discharge to;

( ) a. COORDINATE clean-up and containment of the spilled material. D b. IF OIL is observed ON THE RIVER (more than just sheen), THEN DIRECT Site Protection to position oil booms around the affected water intakes to limit uptake into plant systems (fouling heat exchangers).

2. ASSESS and DETERMINE; D a. IF the Spill/Discharge was EITHER

into a secondary containment
onto an impervious surface AND the material CAN BE completely cleaned up; OR D b. IF the material was EITHER
sewage
sanitary waste AND it DID NOT enter a storm drain or water body;

( ) THEN spill is NOT reportable to NJDEP. GO TO Sect. IL, REPORTING (Pg. 6). D c. OR IF OTHERWISE (more serious Spill/Discharge situation than above), ( ) THEN IMMEDIATELY GO TO Step 3 (next page). HCGS Rev. 00

ECG ATT 16 Pg. 2 of 7 NOTE DO NOT implement notification UNTIL directed to by EITHER Step 4 OR 5. I 1 *

3. COMPLETE "SPILL/DISCHARGE NJDEP NOTIFICATION FORM" (last page) and EXPEDITIOUSLY CONTINUE at Step 4. (next)

D 4. EITHER the Spill/Discharge has; D Passed through an Engineered Fill and INTO the ground water, EAL 11.5.2.!. OR D Entered INTO a storm drain or is observed on the Delaware River from ANY source, EAL 11.5.2 ..Q. THEN 'MEDIATELY (within 15 min.), ( ) a. NOTIFY the NJDEP with the NOTIFICATION FORM information completed in Step 3. (NJDEP phone #'s are on the form) ( ) b. GO TO Step 6. D 5. IF Spill/Discharge DOES NOT meet the criteria in Step 4 AND cleanup is in progress, THEN PERFORM the following: ( ) a. CONTINUE to coordinate cleanup activities and ENSURE personnel performing the cleanup activities keep the on-duty SNSS informed of their progress. ( ) b. NOTIFY Environmental Licensing with details and OBTAIN guidance concerning reportability to NRC. name time D Environmental Licensing determines the event IS reportable to the NRC, ( ) THEN GO TO Step 7. (NRC 4 Hour Report) HCGS Rev. 00

ECG ATT16 Pg. 3 of 7 Initials

5. (cont'd)

D c. IF Spill/Discharge is cleaned up within 24 hrs, ( ) THEN NJDEP notification is NOT required. GO TO Section II., REPORTING (Pg. 6). D d. after 24 hrs the Spill/Discharge is NOT yet cleaned up, ( ) THEN CONT ACT Environmental Licensing again and OBTAIN additional guidance regarding reportability and proceed as follows: name time D 1) Environmental Licensing determines that the Spill/Discharge IS reportable to the NJDEP, ( ) THEN NOTIFY 'MEDIATELY (within 15 min) the NJDEP, with the NOTIFICATION FORM information completed in Step 3. (NJDEP phone #'s are on the form) ( ) GO TO to Step 6 (below). D 2) at the completion of cleanup, Environmental Licensing determines that the Spill/Discharge is NOT reportable, ( ) THEN GO TO Section II., REPORTING (Pg. 6).

6. NOTIFY/UPDATE Environmental Licensing with event details and COMPLETE Substeps a, b, and c below:

name time ( ) a. INFORM Environmental Licensing about status of 15 min. NJDEP call: 0 Call was made within 15 min. of discovery/ confirmation. D Call was NOT made within 15 min., but was made within - - - min. of discovery/confirmation. ( ) b. DIRECT Environmental Licensing to make any required notifications IAW the DPCC/DCR plan. HCGS Rev. 00

ECG ATT16 Pg. 4 of 7 Initials

6. (cont'd)

( ) c. OBTAIN direction from Environmental Licensing concerning NRC reportability of the Event AND PROCEED as directed below: D 1) REPORTABLE to the NRC and NOT done previously, ( ) THEN GO TO Step 7 (below). D 2) NOT REPORTABLE, OR the NRC was previously contacted, ( ) THEN GO TO Section II., REPORTING (Pg. 6). D 7. IF NOT done previously, THEN NOTIFY the Operations Manager (OM). name time PERFORM all of the following Notification Steps.

8. NOTIFY Salem SNSS and provide description of the event.

name time

9. NOTIFY LAC Dispatcher within 4 hrs.

name time

10. COMPLETE an NRC Data Sheet.

( ) OBTAIN a copy from ECG Attachment 5. ( ) OBTAIN assistance from Site Protection and Environmental Licensing personnel, as needed. ( ) ENSURE SNSS approval. _ _ 11. NOTIFY NRC Operations Center within 4 hours. ( ) Use the NRC Data Sheet to record any additional information provided to the NRC. name HCGS Rev. 00

ECG ATT 16 Pg. 5 of 7

12. Notify the NRC Resident Inspector.

name time

13. NOTIFY Public Information Manager (PIM) - Nuclear.

notified at hrs name time

14. NOTIFY Nuclear Licensing.

notified at hrs name time

15. Notify External AfTairs.

name time

16. FAX the NRC Data Sheet to BOTH Public Information and Licensing using the programmed phone numbers on the telecopier.

D 17. completion of Steps 7 thru 16 (Notifications) were DIRECTED by Step 5. b, ( ) THEN GO TO Step 5.c and CONTINUE assessment and coordination of cleanup. D 18. OTHERWISE ( ) THEN GO TO Section II., REPORTING (Pg. 6). HCGS Rev. 00

ECG ATT 16 Pg. 6 of 7 II. REPORTING

1. ENSURE that an Action Request (AR) is prepared.

SNSS

2. FORWARD this attachment, along with the NRC Data Sheet and any supporting SNSS documentation to the Operations Manager (OM).

3. REVIEW this ECG attachment, the AR and any other relevant information for correct OM classification of event and corrective action taken.

4. CONTACT the LER Coordinator (LERC) and.request that the required written reports be OM prepared. Provide this attachment and any other supporting documentation received from the SNSS.

5. PROVIDE Environmental Licensing, with a copy of this attachment including t.he LERC spill/discharge notification report received from the SNSS.

6. PREP ARE LER if required. If an LER is prepared, contact Licensing and ensure that the LERC information on the LER and on the NJDEP Confirmation Report are consistent.

Report or LER Number ~~~~~~~-

7. FORWARD this attachment to the Manager - Nuclear Licensing & Regulation (MNLR).

LERC

8. ENSURE that offsite (state and local) reporting requirements have been met.

MNLR

9. Forward this Attachment/LER package to the Central Technical Document Room MNLR

for microfilming.

HCGS Rev. 00

ECG

                                                                                    .. AJ'T 16 Pg. 7 of 7 SPILL/DISCHARGE NJDEP NOTIFICATION FORM Primary phone# to NJDEP:              292-7172 Backup phone # to NJSP:               882-2000

1. CONT ACT the NJDEP Operator using the above phone numbers.

2. WHEN PROMPTED by the voice answering machme, THEN SELECT .2. for reporting non-emergency releases and an Operator will take the report.

3. RECORD NOTIFICATION TIME: - - - 4. PROVIDE the following information:

"This is notification of a Spill/Discharge." This is (name)_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _, from Hope Creek Generating Station. My call back phone# is 609-339-3027 or 609-339- _ _ __ The Spill/Discharge location is: (provide specific location) at Hope Creek Generating Station located at the Foot of Buttonwood Road, Lower Alloways Creek Township in Salem County. The Common name for the spilled/discharged substance is._ _ _ _ _ _ _ _ _ _ _ _ _ _ __ and we estimated the quantity spilled to be_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ and the substance (HAS) or (HAS NOT) been contained. time date The spill/discharge began at: on _ _ _ __ The spill/discharge was discovered at: on - - - - - The spill/discharge ended at: on _ _ _ __ A description of the Incident is:

                                 -----------------------~

Ongoing actions to contain/clean up the spill are: _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ 33 ft. Wind Direction from: _ _ _ _degrees. Wind Speed: _ _ _ mph (use MET Computer) IF the spill is NOT PSE&G's responsibility, THEN PROVIDE the following info:, Responsible person(s): _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ Company Name, Address and Phone#: _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ May I have your Operator Number please?_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ May I have our CASE Number p l e a s e ? - - - - - - - - - - - - - - - - - - - - - - HCGS Rev. 00

ECG ATf 17 Pg. l of 4 ATTACHMENT 17 ,.. _ _ _ FOUR HOUR REPORT j .,_ . . *-*.* _\_ C:G_P_Y_i_t--j FATALITY OR MEDICAL EMERG~CY ) cos l I

                                                                          ~~~"~-------. .--.~~.                   f' INSTRUCTIONS (HOPE CREEK SNSS or Designee)                                     - ...

A. REFER to Attachment 9, Non-Emergency Notifications Reference, for the current listing of individuals and phone numbers. B. INITIAL each step when completed. C. Implemented b y : - - - - - - - - - - - - Date: _ _ ___ I. NOTIFICATIONS

1. IF NOT done previously, THEN IMPLEMENT HC.FP-EO.ZZ-0003(Z), Control Room Medical Emergency Response.

2. C01\1PLETE an NRC Data Sheet.

( ) OBTAIN a copy from ECG Attachment 5. ( ) OBTAIN assistance from Radiation Protection personnel, as needed. ( ) OBTAIN SNSS approval. name time

3. NOTIFY the LAC Dispatcher of the event.

name time

4. NOTIFY the NRC Operations Center of the event within 4 hours.

( ) RECORD additional information provided to the NRC on the NRC Data Sheet. name time HCGS Rev. 00

ECG ATI17 Pg. 1 of 4 ATTACHMENT 17 FOUR HOUR REPORT FATALITY OR MEDICAL EMERGENCY,

J f

I

                                                                         ._,_._____                      f
                                                                                           *~~~~
                           ,_, INSTRUCTIONS (HOPE CREEK SNSS or Designee)                              ...

A. REFER to Attachment 9, Non-Emergency Notifications Reference, for the current listing of individuals and phone numbers.

8. INITIAL each step when completed.

Date: C. Implemented b y : - - - - - - - - - - - - - ----- I. NOTIFICATIONS

1. IF NOT done previously, THEN IMPLEMENT HC.FP-EO.ZZ-0003(Z), Control Room Medical Emergency Response.

2. COMPLETE an NRC Data Sheet.

~* ., ( ) OBTAIN a copy from ECG Attachment 5. ) ( ) OBTAIN assistance from Radiation Protection personnel, as needed. ( ) OBTAIN SNSS approval. notified at hrs name time

3. NOTIFY the LAC Dispatcher of the event.

notified at hrs name time

4. NOTIFY the NRC Operations Center of the event within 4 houn.

( ) RECORD additional infonnation provided to the NRC on the NRC Data Sheet. notified at hrs

                     ------------------------------------~                         -------

name time HCGS Rev.00

ECG ATI 17 Pg. 2 of 4

5. NOTIFY the NRC Resident Inspector.

name time

6. IF NOT done previously, THEN NOTIFY the Operations Manager (OM).

name time

7. NOTIFY the Public Information Manager (PIM) - Nuclear.

notified at hrs name time

8. NOTIFY Nuclear Licensing.

notified at hrs name time

9. IF transportation of personnel to an Offsite Medical Facility is required, THEN;

( ) a. COMPLETE the report on Pg. 4 of this attachment. ( ) b. NOTIFY the Safety Coordinator (refer to Pg. 4) name time

10. an NBU Employee has died or been seriously injured, THEN;

( ) a. NOTIFY the employee's department manager ( ) b. DIRECT the manager to coordinate notification of the employee's family. name time

11. NOTIFY External Affairs.

notified at - - - - hrs name time HCGS Rev.00

ECG ATI 17. Pg. 2 of 4

5. NOTIFY the NRC Resident Inspector.

name

6. IF NOT done previously, THEN NOTIFY the**Operations Manager*(OM).

           - - - - - - - - - - - - - - - - - notified at _ _ _ _ hrs name                                               time

7. NOTIFY the Public Information Manager (PIM) - Nuclear.

notified at hrs name time

8. NOTIFY Nuclear Licensing.

notified at hrs name time

9. IF transportation of personnel to an Offsite Medical Facility is required, THEN;

( ) a. COMPLETE the report on Pg. 4 of this attachment. ( ) b. NOTIFY the Safety Coordinator (refer to Pg. 4)

           - - - - - - - - - - - - - - - - - - ' - - notified at _ _ _ _ hrs name                                               time

10. an NBU Employee has died or been seriously injured,.,'THEN;

( ) a. NOTIFY the employee's department manager ( ) b. DIRECT the manager to coordinate notification of the employee's family.

           -----------------'---,...,...,..--,.- notified at _ _ _ _ hrs name                                               time I I. NOTIFY External AfTairs.

notified at _ _ _ _ hrs name

time HCGS Rev. 00

ECG ATI 17 Pg. 3 of 4

12. FAX the NRC Data Sheet to BOTH Public Information and Licensing using the programmed phone numbers on the telecopier.

II. REPORTING

1. ENSURE that an Injury Report is completed.

SNSS

2. ENSURE that an Action Request (AR) is prepared.

SNSS AR# _ _ _ _ _ _ _ _ __

3. FORWARD this attachment, along with the NRC Data Sheet and any supporting SNSS documentation, to the Operations Manager (OM).

4. REVIEW this ECG attachment, the AR and any other relevant information for correct OM classification of event and corrective action taken.

5. FORWARD this attachment and any other supporting documentation to the LER OM Coordinator (LERC).

6. PREP ARE required reports.

LERC Report or LER Number _ _ _ _ _ _ _ __

7. FORWARD this attachment to the Manager - Nuclear Licensing & Regulation (MNLR).

LERC

8. ENSURE offsite (state and local) reporting requirements are met.

MNLR

9. FORWARD this Attachment/LER package to the Central Technical Document Room MNLR for microfilming.

HCGS Rev. 00

ECG ATT17 Pg. 4 of 4 REPORT OF SERIOUS INJURY/DEA TH NUCLEAR BUSINESS UNIT EMPLOYEE El\1PLOYEE INFORMATION EMPLOYEE#_-_ _ _ _ AGE HOME PHONE#

                - - - - - - - - - - - MARITAL STATUS - - - - - - - -

JOB TITLE LOCATION

           -------------~                              --------

SOCIAL SECURITY # ACCIDENT/INJURY DESCRIPTION DATE OF ACCIDENT

                   - - - - - - TIME - - - - - AM/PM DID INJURIES RESULT IN DEATH     D  YES       D  NO EXTENT OF INJURIES DESCRIPTION OF ACCIDENT
                         -------------------~

WHERE TAKEN AFTER ACCIDENT _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ SAFETY COORD. WORK# HOME# PAGER# Cliff Knaub 2812 358-3074 478-5706 John Horner 2965 678-6308 342-5866 Andrew Caplinger 2828 478-5983 HCGS Rev. 00

ECG ATT 17 Pg. 4 of 4 REPORT OF SERIOUS INJURY/DEATH NUCLEAR BUSINESS UNIT EMPLOYEE EMPLOYEE INFORMATION EMPLOYEE#.- - - - - AGE HOl'v1EPHONE# _ _ _ _ _ _ _ _ _ _ _ MARITALSTATUS _ _ _ _ _ __ JOB TITLE - - - - - - - - - - - - - - LOCATION - - - - - - - - SOCIAL SECURITY # ACCIDENT/INJURY DESCRIPTION DATE OF ACCIDENT - - - - - - TIME - - - - - ' - -; AM/PM ., .,, DID INJURIES RESULT IN DEATH D YES D NO EXTENT OF INJURIES

                      ---------------------~

DESCRIPTION OF ACCIDENT WHERE TAKEN AFTER ACCIDENT

                              -----------------~

SAFETY COORD. WORK# HOl'v1E # PAGER# Cliff Knaub 2812 358-3074 478-5706 John Homer 2965 678-6308 342-5866 Andrew Caplinger 2828 478-5983 HCGS Rev. 00

ECG ATf 18

                                                                                 '*'*-.           Pg. 1 of 4 ATTACHMENT 18 I
                                                                              /

FOUR HOUR REPORT ,: * , .~ / RADIOLOGICAL TRANSPORTATION ACC~ENT ~ I

                                                                           -* *.,.,,~,....                                           I
  .--~~~~~~~~~~~~~~~~~~~~~~~~_.;.;;'"'-:::--~~~~~---,j
                                                                                             "*n.._*..-..,°"~..,,_,,,.,,..~-..,..,

INSTRUCTIONS (HOPE CREEK SNSS or Designee) A. REFER to Attachment 9, Non-Emergency Notifications Reference, for the current listing of individuals and phone numbers. B. INITIAL each step when completed. Date: C. Implemented b y : - - - - - - - - - - - - - ---- I. NOTIFICATIONS

1. C01\1PLETE the ACCIDENT NOTIFICATION FORM (last page) with initial details received regarding the accident.

2. OBTAIN a copy of the applicable Radwaste Shipping document for reference during subsequent notifications.

3. IF PSE&G is the carrier (driver is a PSE&G employee),

THEN NOTIFY the Department of Transportation (DOT) at 1-800-424-8802. ( ) PROVIDE all information recorded on the ACCIDENT NOTIFICATION FORM. ( ) RECORD any additional information requested by DOT.

              ~~-~~~~-~~~~~~~~~

notified at ~~~- hrs name time

4. DIRECT the Radiation Protection Manager (or alternate) to contact the carrier's dispatcher and coordinate assistance in implementing PSE&G's response, as required.

               ~~-~~~~~~~-~~~~~~

notified at ~~~~ hrs name time HCGS Rev. 00

ECG ATT18 Pg. 2 of 4

5. COMPLETE an NRC Data Sheet.

( ) OBTAIN a copy from ECG Attachment 5. ( ) OBTAIN assistance from Radiation Protection personnel, as needed. ( ) OBTAIN SNSS approval.

6. NOTIFY the Public Information Manager (P') - Nuclear.

name time

7. NOTIFY the NRC Operations Center within 4 hours.

( ) RECORD additional information provided to the NRC on the NRC Data Sheet. notified at hrs name time

8. NOTIFY the NRC Resident Inspector.

notified at hrs name time

9. IF NOT done previously, THEN NOTIFY the Operations Manager (OM).

notified at hrs name time IO. NOTIFY Nuclear Licensing. notified at hrs name time

11. NOTIFY External Affairs.

notified at hrs name time

12. FAX the NRC Data Sheet to BOTH Public Information and Licensing using the programmed phone numbers on the telecopier.

HCGS Rev. 00

ECG ATT 18 Pg. 3 of 4 II. REPORTING

1. ENSURE that an Action Request (AR) is prepared.

SNSS

2. FORWARD this attachment, along with the NRC Data Sheet and any supporting SNSS documentation, to the Operations Manager (OM).

3. REVIEW this ECG attachment, the AR and any other relevant information for correct OM classification of event and corrective action taken.

4. FORWARD this attachment and any other supporting documentation to the LER OM Coordinator (LERC).

5. PREP ARE required reports.

LERC Report or LER Number - - - - - - - - -

6. FORWARD this attachment to the Manager - Nuclear Licensing & Regulation (rvfNLR).**

LERC

7. ENSURE offsite (state and local) reporting requirements are met.

MNLR

8. FORWARD this Attachment/LER package to the Central Technical Document Room MNLR for microfilming.

HCGS Rev. 00

ECG ATI 18 Pg. 4 of 4 RADIOLOGICAL TRANSPORTATION ACCIDENT NOTIFICATION FORM INSTRUCTIONS: A. RECORD the minimum information required for an effective PSE&G response.

8. RECORD any additional information provided as requested by the DOT.

Time of Call Caller's Name: Phone Number: Are you the driver? 0 YES 0 NO IF YES, Trucking Company N a m e : - - - - - - - - - - - - - - - - - - - - - - IF NO, What is the status of the driver? LO CAT10N of Accident: IRoadwa1/Mile Marker/Intersection City/Town State Number of Vehicles involved? 1-2-3-4 State or Local Police on the scene? D YES D NO Any personnel injuries? D YES D NO Any Fire involving truck contents? D YES D NO Trucking Company Dispatcher notified? D YES D NO Extent of damage to truck/trailer, container and contents: ASK THE CALLER TODO THE FOLLOWING: A. IF NOT yet done, NOTIFY the State or Local Police. B. IF possible, ENSURE assistance personnel at the accident scene do the following:

1. TAKE all practical measures to protect life and property, THEN stay back and wait for trained emergency personnel.

2. REMAIN upwind of the accident; DO NOT track thru any spills.

HCGS Rev. 00

ECG ATI 19

                                                                                '~~*.)of                    3
                                                                                       '*1-    "* :.. *~-,..--

ATTACHMENT 19 .- ~-: *.; ;** .. ii" TWENTY-FOUR HOUR REP~RT j I i FITNESS FOR DUTY (FFD) PROGRAM EVENTS INSTRUCTIONS (HOPE CREEK SNSS or Designee) A. REFER to Attachment 9, Non-Emergency Notifications Reference, for the current listing of individuals and phone numbers. B. INITIAL each step when completed. Date: C. Implemented b y : - - - - - - - - - - - - ---- CAUTION The determination of reportability of significant FFD events is the responsibility of the Medical Review Officer (MRO). In order to ensure compliance with NRC notification requirements of 10CFR26.73 and also protect the rights of the individual(s} involved, information provided to any of the below contacts SHALL be limited to that supplied by the MRO or designee. I. NOTIFICATIONS

1. COMPLETE the significant FFD Event report form (last page) with the details received from the Medical Review Officer (MRO) or designee per NC.NA-AP.ZZ-0042(Q).

2. NOTIFY the NRC Operations Center within 24 hours of the time of discovery provided by the :MRO.

name

3. NOTIFY the NRC Resident Inspector.

                 ~~~~~~~~~~~~~~~~

notified at ~--- hrs name time HCGS Rev. 00

ECG ATf 19 Pg. 2 of 3

4. IF NOT done previously, THEN NOTIFY the Operations Manager (OM).

name time II. REPORTING CAUTION ALL records of this report shall be handled as CONFIDENTIAL.

1. ENSURE that an Action Request (AR) is prepared.

SNSS

2. FORWARD this attachment, along with any supporting documentation, to the SNSS Operations Manager (OM).

3. REVIEW this ECG attachment, the AR and any other relevant information for correct OM classification of event and corrective action taken.

4. FORWARD this attachment and any other supporting documentation to the Medical OM Review Officer (MRO) at the Processing Center MC - N06.

5. RETAIN this information on file IAW Nuclear Medical Department Standard Operating MRO Procedures AND ENSURE that this event is included in the 6 month FFD Report to the NRC.

HCGS Rev. 00

ECG ATT 19

                                                                                           *Pg. 3 <>f 3 CONFIDENTIAL FITNESS FOR DUTY (FFD) PROGRAM EVENT NRC NOTIFICATION REPORT FORM INSTRUCTIONS:

A. SNSS should use this form to document the details of any FFD event determined by the Medical Review Officer (MRO) to be reportable per 10CFR26.73. B. Initial NRC report shall be completed within 24 hours from the time of discovery by the licensee, as determined by the MRO. C. IF the NRC FFD Representative requires additional or more detailed information, the NRC shall directly contact the MRO. NRC NOTIFICATION: Notification Time: SNSS (name) _ _ _ _ _ _ _ _ _ _ _ __ Facility: Salem/ Hope Creek Call back phone# 609-339-_ _ _ __ EVENT DETAILS:

1. Medical Review Officer or designee: - - - - - - - - - - - - - - - - - - - -

Call back phone# 609-339-5600 (name) MRO Beeper# 609-573-4588

2. Reporting Event

( ) Sale, use, or possession of illegal drugs within the Protected Area [10CFR26.73(a)(l)] OR ( ) Any acts, by Licensed Reactor Operators, Security Force Members, or Supervisory personnel: [ 10CFR26.73(a)(2)] ( ) Involving the sale, use, or possession of a controlled substance. (i) ( ) Resulting in a confirmed positive test on such persons. (ii) ( ) Involving use of alcohol within the Protected Area. (iii) ( ) Resulting in the determination of unfitness for scheduled work due to consumption of alcohol. (iv) ( ) False Positive Lab Results due to an administrative error. [10CFR26, APP. A, 2.8(e)(5)] ( ) Any other FFD related event determined reportable by the MRO IAW NC.NA-AP.ZZ-0042(Q).

3. Discovery Time: - - - - - - h r s o n - - - - - - - (date)

4. Work Dept. of individual(s): - - - - - - - - - - - - - - - - - - - - - - -

5. Has plant safety been affected ? DYES D NO

6. Corrective actions taken or p l a n n e d ? - - - - - - - - - - - - - - - - - - - -

7. Other pertinent information: - - - - - - - - - - - - - - - - - - - - - - -

HCGS Rev. 00

ECG ATI20 Pg. 1 of 3 ATTACHMENT 20 TWENTY-FOUR HOUR REPORT NRC REGIONAL OFFICE INSTRUCTIONS (HOPE CREEK SNSS or Designee) A. REFER to Attachment 9, Non-Emergency Notifications Reference, for the current listing of individuals and phone numbers. B. INITIAL each step when completed. C. Implemented b y : - - - - - - - - - - - - - Date: -...................-_________

                                                                                   ... ..,.~.-                                 ..
                                                                                               -*        -* ._; .      ~ I; I.       NOTIFICATIONS Initials

1. COMPLETE an NRC Data Sheet. ~ 1......,-~-.r,.>.,~ ~...,,.DCl:~.~--- ..i

( ) OBTAIN a copy from ECG Attachment 5. ( ) OBTAIN assistance from Radiation Protection personnel, as needed. ( ) OBTAIN SNSS approval.

2. NOTIFY the NRC Region 1 Office within 24 hours.

( ) RECORD additional information provided to the NRC on the NRC Data Sheet. name time

3. NOTIFY the NRC Resident Inspector.

notified at

                -------              ----------                               - - - - - hrs name                                          time

4. NOTIFY the NRC Operations Center within 24 hours.

( ) RECORD additional information provided to the NRC on the NRC Data Sheet. notified at hrs

                 ---------------~                                             ----

name time

5. IF NOT done previously, THEN NOTIFY the Operations Manager (OM).

notified at hrs

                 ---------------~                                             ----

name time HCGS Rev. 00

ECG ATI20 Pg. 2 of 3

5. NOTIFY the Public Information Manager (P') - Nuclear.

name

6. NOTIFY Nuclear Licensing.

notified at hrs name time

7. NOTIFY External Affairs.

notified at hrs name time

8. FAX the NRC Data Sheet to BOTH Public Information and Licensing using the programmed phone numbers on the telecopier.

HCGS Rev. 00

r-------- ECG ATI20 Pg. 3 of 3

         *'II. REPORTING

1. ENSURE that an Action Request (AR) is prepared.

SNSS

2. . FORWARD this attachment, along with the NRC Data Sheet and any supporting SNSS documentation, to the Operations Manager (OM).

3. REVIEW this ECG attachment, the AR and any other relevant information for correct OM classification of event and corrective action taken.

4. FORWARD this attachment and any other supporting documentation to the LER OM Coordinator (LERC).

5. PREP ARE required reports.

LERC Report or LER Number - - - - - - - - -

6. FORWARD this attachment to the Manager - Nuclear Licensing & Regulation (MNLR).

LERC

7. ENSURE offsite (state and local) reporting requirements are met.

MNLR

8. FORWARD this Attachment/LER package to the Central Technical Document Room MNLR for microfilming.

HCGS *Rev. 00

ECG ATI21 Pg!.,,.l....of-2-i ATTACHMENT 21 "::-:.~~:)**{ rr \

                                                                                                           \
                                                                                                           \
                         .              REPORTABLE EVENT            \

LAC/MEMORANDUM OF UNDERSTANDING (M.O.U.) I GS 1 ,,.--..

                                                                                                            \
                                                                                                             \

INSTRUCTIONS (HOPE CREEK SNSS or Designee) A. REFER to Attachment 9, Non-Emergency Notifications Reference, for the current listing of individuals and phone numbers. B. INITIAL each step when completed. Date: C. Implemented by: - - - - - - - - - - - - - ---- I. NOTIFICATIONS Initials

l. PROVIDE an event d e s c r i p t i o n : - - - - - - - - - - - - - - - - - -

2. NOTIFY the LAC Dispatcher within four hours of the event.

notified at hrs name time

3. IF NOT done previously, THEN NOTIFY the Operations Manager (OM).

notified at hrs name time

4. NOTIFY the Public Information Manager (PIM) - Nuclear.

notified at hrs name time HCGS Rev. 00

ECG ATf 21 Pg. 2 of 2 Initials

5. NOTIFY External Affairs.

name time II. REPORTING

1. ENSURE that an Action Request (AR) is prepared.

SNSS

2. FORWARD this attachment, along with any supporting documentation, to the SNSS Operations Manager (OM).

3. REVIEW this ECG attachment, the AR and any other relevant information for correct OM classification of event and corrective action taken.

4. FORWARD this attachment and any other supporting documentation to the LER OM Coordinator (LERC).

5. PREP ARE required reports.

LERC Report or LER Number - - - - - - - - -

6. FORWARD this attachment to the Manager- Nuclear Licensing & Regulation (MNLR).

LERC

7. ENSURE offsite (state and local) reporting requirements are met.

MNLR

8. FORWARD this Attachment/LER package to the Central Technical Document Room MNLR for microfilming.

HCGS Rev. 00

ECG

                                                                                     *----~Tf 22
                                                                 ......... -~
                                                                                       . ' ~g..:* r~oT'-2--

ATTACHMENT 22 T/S REQUIRED ENGINEERING EVALU~ TION (:~;

                                                                  - ..........._____      *~*"-* ----..... ...._ __

INSTRUCTIONS (HOPE CREEK SNSS or Designee) A. REFER to Attachment 9, Non-Emergency Notifications Reference, for the current listing of individuals and phone numbers. B. INITIAL each step when completed. Date: C. Implemented by: - - - - - - - - - - - - - ---- I. NOTIFICATIONS NOTE This attachment is for initiating an Engineering Evaluation required by Technical Specifications. No Offsite or external notifications are performed by this attachment, but should be implemented as determined by the results of the evaluation. I. PROVIDE an event d e s c r i p t i o n : - - - - - - - - - - - - - - - - - - CAUTION Refer to the ECG sections related to the Initiating Conditions of this event to determine if any NRC notifications are also required.

2. IF ANY NRC Notifications are ALSO required, THEN IMPLEMENT the other referenced attachment in parallel with this one.

HCGS Rev. 00

ECG ATI22 Pg. 2 of 2 Initials

3. NOTIFY the Technical Manager or Technical Engineer with details of the event.

name time

4. IF NOT done previously, THEN NOTIFY the Operations Manager (OM).

name time II. REPORTING

1. ENSURE that an Action Request (AR) is prepared.

SNSS

2. FORWARD this attachment, along with any supporting documentation, to the SNSS Operations Manager (OM).

3. REVIEW this ECG attachment, the AR and any other relevant information for correct OM classification of event and corrective action taken.

4. FORWARD this attachment and any other supporting documentation to the LER OM Coordinator (LERC).

5. PREP ARE required reports.

LERC Report or LER Number - - - - - - - - -

6. FORWARD this attachment to the Manager - Nuclear Licensing & Regulation (MNLR).

LERC ___ 7. ENSURE offsite (state and local) reporting requirements are met. MNLR

8. FORWARD this Attachment/LER package to the Central Technical Document Room MNLR for microfilming.

HCGS Rev. 00

~--- ECG AIT24

ATTACHMENT 24

UNUSUAL EVENT (COMMON SITE) '\

                                                                       . \                             \

NOTE ONLY one SNSS is required to declare this event and assume the responsibilities of Emergency Coordinator (EC). The other SNSS should perform the duties of the Unaffected Station SNSS during the implementation of this attachment. CAUTION IN THE EVENT OF OFFSITE TOXIC GAS RELEASE AFFECTING THE SITE, EVACUATION OF NON-ESSENTIAL PERSONNEL TAKES PRECEDENCE OVER NOTIFICATIONS. I. COMMON SITE EVENT ASSESSMENT/ EC DETERMINATION Initials HOPE CREEK SENIOR NUCLEAR SHIFT SUPERVISOR (SNSS) SHOULD: A. NOTIFICATION OF SALEM SNSS

1. CONTACT the Salem SNSS and brief him/her on the specific SNSS circumstances as follows:

( ) a. SHARE information about the externally initiated event in progress. ( ) b. OBTAIN agreement on the Unusual Event classification. ( ) c. DETERMINE which SNSS will assume EC responsibilities. Emergency C o o r d i n a t o r : - - - - - - - - - - - - - - - - - -

2. IF the Hope Creek SNSS is the EC, SNSS THEN IMMEDIATELY IMPLE:rviENT this attachment as EC.

3. IF an Offsite Toxic Gas Release is threatening Site Personnel (EAL 9.4.1.a),

EC THEN IMMEDIATELY IMPLE:rviENT appropriate Protective Actions for Site Personnel including initiation of Accountability and Evacuation per Section ill., Pg. 4, PRIOR TO notifications. HCGS Rev. 00

ECG ATI24 Pg. 2 of 10 II. EMERGENCY COORDINATOR CECl LOG SHEET A. DECLARE A COMMON SITE UNUSUAL EVENT EC AT HOPE CREEK AND SALEM EAL# _ _ _ _ _ _ __ Declared at - - - - hrs on - - - - time date B. NOTIFICATIONS ( ) 1. CALL communicators to the Control Room. ( ) 2. COMPLETE the INITIAL CONTACT MESSAGE FORM (IC:MF) (last page of this attachment). ( ) 3. PROVIDE the ICMF to the Communicator (CMl) and DIRECT the CMl to implement Attachment 6. ( ) 4. DIRECT the Secondary Communicator (CM2) to implement Attachment 8 for an Unusual Event. ( ) 5. SOUND the Radiation Alert Alarm and make the following page announcement:

              "Attention all personnel. Attention all personnel."
              "Hope Creek and Salem are both in an UNUSUAL EVENT condition due to II (Repeat)

C. SECURITY RELATED EVENT

1. IF Security Related, EC THEN DIRECT the PSE&G Security Supervisor (x2222) to implement the Security Contingency Plan.

2. IF a bomb search is required, EC THEN;

a. DIRECT the OSC Coordinator to;

( ) ACTIVATE the OSC IAW EPIP 202H, OSC Activation and Operations AND ( ) IMPLEMENT Bomb Search Operations IAW Appendix 1. ( ) b. DIRECT the NCOs to check control boards for correct equipment lineups. HCGS Rev. 00

ECG ATI24 Pg. 3 of 10 Initials D. EMERGENCY COORDINATOR DUTIES ( ) 1. NOTIFY the Salem SNSS. (NETS 5121; DID 5200) ( ) 2. required, IMPLEMENT Accountability by r~ferring to the Accountability Instructions in Section III.

3. WHEN provided by the CM2, EC THEN COMPLETE and APPROVE the NRC Data Sheet for transmittal by the CM 1 within 60 minutes.

4. WHEN provided by the CM2, EC THEN REVIEW and APPROVE the Station Status Checklist (SSCL) for transmittal.

( ) a. REPEAT this step approximately every half hour. ( ) b. PERFORM immediately for any significant change in emergency status. (operational or radiological) E. TURNOVER IF relieved prior to termination of the Unusual Event, EC THEN DOCillviENT the name of your relief below:

            ~~~~~~~~~~~~~~~~~-

assumed EC duties at -~~- Name time F. ESCALATION IF event classification escalates above Unusual Event, EC THEN EXIT this attachment and implement a new attachment as directed by the EALs. G. TERMINATION

1. TERMINATE the UE IAW Section IV., Emergency Termination/Reduction/

EC Recovery (Pg. 6).

2. ENSURE appropriate reports are made IAW Section V., Reporting, of this SNSS attachment.

HCGS Rev. 00

ECG ATI24 Pg. 4 of 10 ill. ACCOUNTABILITY INSTRUCTION FOR THE PROTECTED AREA A. 'PLEMENTATION OF ASSEMBLY AND ACCOUNTABILITY Initialsffime I 1. IF NOT already done, EC THEN DIRECT the OSC Coordinator to activate the OSC IAW EPIP 202H,. OSC Activation and Operations. I 2. IF Accountability AND Evacuation is required, EC THEN DIRECT Security (x2222) to IMPLEMENT EPIP 901, Onsite Security Response, and EPIP 902, Accountability/ Evacuation, Sections 3 .2 and 3.3. I 3. IF NO Evacuation is required, EC THEN DIRECT Security (x2222) to IMPLEMENT EPIP 901, Onsite Security Response, and EPIP 902, Accountability/ Evacuation, Sections 3.1 and 3.2 ONLY, for Assembly and Accountability. I 4. DIRECT the Salem SNSS to implement EPIP 101S, Appendix 6, EC Accountability Instructions For An Unusual Event at Hope Creek. NOTE Steps A.5 thru A.9 may be delegated by the EC to any available CR Staff member. I 5. SOUND the Radiation Alert Alarm and make the following page announcement:

                      "Attention all personnel. Attention all personnel."
                     "Hope Creek and Salem are both in an UNUSUAL EVENT condition due to "All PSE&G personnel assemble at your Accountability Stations. All contractors leave the Owner Controlled Area immediately". (Repeat)

I 6. WAIT for 5 minutes for key personnel to reach their Accountability Stations, TIIEN CONTINUE with Step 7 .. I 7. SOUND the Radiation Alert Alarm and ANNOUNCE the following; (T= O Min.)

                      "Attention, Attention. All accountability stations, IMPLEMENT Accountability." (Repeat)

HCGS Rev. 00

ECG ATI24 Pg. 5 of 10 Ill. ACCOUNTABILITY INSTRUCTION FOR THE PROTECTED AREA (Cont'd) Initialsffime I 8. WHEN 10 minutes have elapsed from Step 7, ANNOUNCE the following; (T+lO Min.)

                      "Attention, Attention. All accountability stations, COMPLETE YOUR INITIAL Accountability." (Repeat)

I 9. WHEN 20 minutes have elapsed from Step 7, ANNOUNCE the following; (T+20 Min.)

                      "Attention, Attention. All accountability stations, COMPLETE YOUR 30 minute Accountability." (Repeat)

I 10. WHEN 30 minutes have elapsed from Step 7, EC (T+30 Min.) COORDINATE with the TSC Security Liaison and OBTAIN a list of unaccounted-for personnel. B. LOCATION OF UNACCOUNTED-FOR PERSONNEL

1. LOCATE unaccounted-for personnel as follows:

EC ( ) a. PAGE individuals over the plant page. ( ) b. OBTAIN feedback from co-workers/supervisors on the last known location/job assignment. ( ) c. DIRECT Security to assist in locating unaccounted for personnel. ( ) d. CALL individual's home to verify work schedule. ( ) e. IF REQUIRED, THEN DIRECT the OSCC to INITIATE Search and Rescue Operations IAW EPIP 202H.

2. UPDATE Security as missing personnel are accounted for.

EC HCGS Rev. 00

ECG ATI24 Pg. 6 of 10 IV. TERMINATION Initials

1. WHEN EITHER of the following conditions are met, EC THEN TERMINATE the emergency by proceeding to Step 2.

( ) a. NO EALs are exceeded AND the Plant is stable. ( ) b. IF any EAL CONTINUES to be exceeded AND the Plant is stable THEN REFER to the "RECOVERY CHECKLIST" (Pg. 7) AND DETERMINE if the UE can be terminated by entering Recovery.

2. WHEN the above Step is completed, EC THEN C01\.1PLETE the "UNUSUAL EVENT TERMINATION/RECOVERY NOTIFICATION FORM," (Pg. 8), as follows:

( ) a. IF terminating WITHOUT Recovery, C01\.1PLETE Part A. ( b. IF terminating WITH Recovery, C01\.1PLETE Part B.

3. IF termination with Recovery is chosen, EC THEN DIRECT the EDO to assume the duties of the Recovery Manager including:

a. EV ALU ATE the emergency and its consequences.

b. DETERMINE measures required to return the Plant to Normal Operations (termination of Recovery Status).

c. COORDINATE contractor support, as required.

4. Make Reduction in Event Notifications (Termination) by; EC

( ) a. PROVIDE the completed "EMERGENCY TERMINATION/ RECOVERY NOTIFICATION FORM," to the CMI. ( ) b. DIRECT the CMl to make the termination notifications IAW ECG Attachment 6.

5. MAKE a PA announcement to update Plant personnel.

EC

6. NOTIFY the Salem SNSS.

EC

7. GO TO Section V., Reporting.

SNSS HCGS Rev. 00

ECG ATI24 Pg. 7 of 10 IV. TERMINATION (cont'd) RECOVERY CHECKLIST FOR A COMtvlON SITE UNUSUAL EVENT THE EMERGENCY COORDINATOR SHOULD: A. ANSWER each of the following questions which are PREREQUISITES for Terminating WITH Recovery. CHECK IF YES FOR BOTH HOPE CREEK AND SALEM ( ) 1. Is the Radiological Release terminated(< Technical Specifications)? ( ) 2. Are Radiation levels in ALL areas of the SITE EITHER stable or decreasing? ( ) 3. Is the SITE in a safe, stable condition with NO reason to expect further degradation? ( ) 4. Is the integrity of the Station power supplies and ECCS equipment required for safe shutdown intact? ( ) 5. Can full time operations of BOTH OSCs be terminated? B. IF ANY of the above are negative (unchecked), THEN termination should NOT be performed, at this time. RETURN to Section II. C. IF ALL of the above are checked as YES, THEN PROCEED with Step D. D. Salem and Hope Creek EDOs have both been briefed AND (CHECK IF YES); ( ) 1. BOTH EDOs concur that terminating the UE with an EAL still exceeded is correct under the current circumstances? ( ) 2. Hope Creek EDO is prepared to assume the duties of Recovery Manager. Time E. IF EITHER of the above are negative (unchecked), THEN termination should NOT be performed, at this time. RETURN to Section I. F. IF BOTH D. l & D.2 are checked as YES, THEN SIGN below and GO TO Sect. IV., Step 2 for Terminating WITH Recovery. Emergency Coordinator Date Time HCGS Rev. 00

ECG ATI24 Pg. 8 of 10 IV. TERMINATION (cont'd) UNUSUAL EVENT TERMINATION/RECOVERY NOTIFICATION FORM PART "A" - EMERGENCY TERMINATION WITHOUT RECOVERY: TIDS IS CO:MMUNICATOR IN THE CONTROL ROOM AT

                    --------~

THE HOPE CREEK NUCLEAR GENERATING STATION. TIDS :rvtESSAGE IS TO NOTIFY YOU THAT AS OF , ON time date THE COMMON SITE UNUSUAL EVENT AFFECTING BOTH HOPE CREEK AND SALEM HAS BEEN TERMINATED. (EC Approval to transmit) PART "B" - EMERGENCY TERMINATION WITH RECOVERY: TlilS IS _ _ _ _ _ _ _ ___, CO.l\.1MUNICATOR IN THE CONTROL ROOM AT THE HOPE CREEK NUCLEAR GENERATING STATION. TIDS :rvtESSAGE IS TO NOTIFY YOU THAT AS OF ____ __, ON ____ _.. time date THE COMMON SITE UNUSUAL EVENT AFFECTING BOTH HOPE CREEK AND SALEM HAS BEEN TERMINATED AND RECOVERY OPERATIONS IMPLE:rvtENTED. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ IS THE RECOVERY MANAGER. ( HC DUTY EDO) LOCATED AT HOPE CREEK (EC Approval to transmit) HCGS Rev. 00

ECG ATT24 Pg. 8 of 10 IV. TERMINATION (cont'd) UNUSUAL EVENT TERMINATION/RECOVERY NOTIFICATION FORM PART "A" - EMERGENCY TERMINATION WITHOUT RECOVERY: THIS IS _________, COMMUNICATOR IN THE CONTROL ROOM AT THE HOPE CREEK NUCLEAR GENERATING STATION. THIS MESSAGE IS TO NOTIFY YOU THAT AS OF , ON time date THE C01\1MON SITE UNUSUAL EVENT AFFECTING BOTH HOPE CREEK' AND SALEM HAS BEEN TERMINATED. (EC Approval to transmit) PART "B"- EMERGENCY TERMINATION WITH RECOVERY: THIS IS _ _ _ _ _ _ ____, CO:rviMUNICATOR IN THE CONTROL ROOM AT THE HOPE CREEK NUCLEAR GENERATING STATION. THIS MESSAGE IS TO NOTIFY YOU THAT AS OF _ _ __ , ON _ _ _ __, time date THE C01\1MON SITE UNUSUAL EVENT AFFECTING BOTH HOPE CREEK AND SALEM HAS BEEN TERMINATED AND RECOVERY OPERATIONS IMPLEMENTED.

      - - - - - - - - - - - ' - - - - ' - - - - ' - - - - - I S THE RECOVERY MANAGER.

( HC DUTY EDO) LOCATED AT HOPE CREEK (EC Approval to transmit) HCGS Rev.00

ECG ATI24 Pg. 9 of 10 V. REPORTING INSTRUCTIONS

1. This is a permanent document.

2. ATTACH appropriate documents to this form and EXPEDITE the package through all steps.

Initials

1. PREPARE an Action Request (AR).

SNSS

2. FORWARD this attachment and supporting documentation, to the Operations SNSS Manager (OM).

3. REVIEW this attachment, the (AR) and any other relevant information for correct OM classification of event and corrective action taken.

4. CONT ACT the LER Coordinator (LERC) and request that the required reports be OM prepared. Provide this attachment and any other supporting documentation to the LERC.

5. PREP ARE required reports.

LERC Report or LER Number-------

6. FORWARD this attachment to the Central Technical Document Room for .

LERC microfilming. HCGS Rev. 00

ECG AIT24 Pg. 10 of 10 INITIAL CONTACT MESSAGE FORM I. TIDS I S - - - - - - - - - - , COMMUNICATOR IN THE CONTROL ROOM (NAME) AT THE HOPE CREEK NUCLEAR GENERATING STATION. II. 0 THIS IS NOTIFICATION OF A CO:MMON SITE UNUSUAL EVENT AFFECTING BOTH SALEM AND HOPE CREEK WHICH WAS DECLARED AT ON - - - - - - - - - (Time - 24 HR CLOCK) (DATE) EAL# - - - - - - DESCRIPTION OF EVENT: III. NO RADIOLOGICAL RELEASE IS IN PROGRESS 33 FT. LEVEL WIND DIRECTION (From): WIND SPEED: - - - - (From MET Computer) (DEGREES) . (MPH) IV. NO PROTECTIVE ACTIONS ARE RECOMMENDED AT THIS TIME EC Initials (Approval to Transmit ICMF) HCGS Rev.00

ECG

                                                                                             ~t\Tf:25
                                                                          ., . ~-*\'i~                   t
                                                              * *"'* *         . ... ,..',:*; ; "Pg. I\ of 3 ATTACHMENT 25;                                                   1 i                                           i ONE HOUR       REPOR~     !

(.. GS l l (COMMON SITE) ; ,, ,,,,._,,,, ..-~~-- MAJOR LOSS OF EMERGENCY ASSESSMENT~.OFFSITE'RESPONSE, OR COMMUNICATIONS CAPABILITY NOTE ONLY one SNSS, Hope Creek or Salem, is required to report this event which is common to both stations. I. EVENT ASSESSMENT AND DETERMINATION OF NOTIFICATION RESPONSIBILITY I. NOTIFY the Salem SNSS (NETS x5121; DID 5200).

2. DETER..1'vHNE which Station SNSS will implement this attachment.

3. IF the Hope Creek SNSS is responsible for this notification, THEN IMMEDIATELY CONTINUE with this attachment.

4. IF the Salem SNSS will implement this attachment, THEN NO further actions are required by Hope Creek except to lend assistance as necessary in restoring the lost equipment or capabilities.

INSTRUCTIONS (HOPE CREEK SNSS or Designee) A REFER to Attachment 9, Non-Emergency Notifications Reference, for the current listing of individuals and phone numbers. B. INITIAL each step when completed. Date: C. Implemented b y : - - - - - - - - - - - - ---- II. NOTIFICATIONS .

1. COMPLETE an NRC Data Sheet.

( ) OBTAIN a copy from ECG Attachment 5. ( ) OBTAIN assistance from Radiation Protection personnel, as needed. ( ) ENSURE SNSS approval. HCGS Rev.00

ECG I. ATI25 Pg. 2 of 3 Initials I* . *

2. NOTIFY the NRC Operations Center of the event within 1 hour.

( ) RECORD additional information provided to the NRC on the NRC Data Sheet. notified at hrs name time

3. NOTIFY the NRC Resident Inspector.

notified at hrs name time

4. IF NOT done previously, THEN NOTIFY the Operations Manager (OM).

notified at hrs name *time

5. NOTIFY the Public Information Manager (PIM) - Nuclear.

notified at hrs name time

6. NOTIFY Nuclear Licensing.

notified at hrs name time

7. IF a major loss of communications capability has occurred (such as loss of ENS, NETS, DID, etc.)

THEN NOTIFY: l.T. Client Service Center: (201-430-7500 or ESSX 7500) ( ) a. ENTER [ 1 1] in response to the automated answering system prompts. ( ) b. NOTIFY the Operator that the failed system is an "Emergency Priority Circuit." notified at hrs name time

8. NOTIFY External Affairs.

notified at hrs name time

HCGS kev. 00

                                                                                              """' *l I

ECG ATI25 Pg. 3 of 3

9. FAX the NRC Data Sheet to BOTH Public Information and Licensing using the programmed phone numbers on the telecopier.

III. REPORTING

1. ENSURE that an Action Request (AR) is prepared.

SNSS *

2. FORWARD this attachment, along with the NRC Data Sheet and any supporting SNSS documentat:on, to the Operations Manager (OM).

3. REVIEW this ECG attachment, the AR and any other relevant information for correct OM classification of event and corrective action taken.

4. FORWARD this attachment and any other supporting documentation to the LER OM Coordinator (LERC).

5. PREP ARE required reports.

LERC Report or LER Number _ _ _ _ _ _ _ __ - - - 6. FORWARD this attachment to the Manager - Nuclear Licensing & Regulation (MNLR). LERC ___ 7. ENSURE offsite (state and local) reporting requirements are met. MNLR ___ 8. FORWARD this Attachment/LER package to the Central Technical Document Room MNLR for microfilming.

                                                                                            /I HCGS                                                                            Rev. oo,,/}}

ML18102A794

Text

Reference:

Reference:

2.1.4 OPCON

2.1.8 DISCUSSION

2.1.9 DEVIATION

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